ML24177A138
| ML24177A138 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/25/2024 |
| From: | Mary Johnson NRC/NRR/DNRL/NLRP |
| To: | Carr E Dominion Innsbrook Technical Center |
| References | |
| EPID L-2023-SLR-0003 | |
| Download: ML24177A138 (102) | |
Text
June 25, 2024 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 - REPORT FOR THE AGING MANAGEMENT AUDIT REGARDING THE SUBSEQUENT LICENSE RENEWAL APPLICATION REVIEW (EPID NO. L-2023-SLR-0003)
Dear Mr. Carr:
By letter dated August 17, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23233A175), as supplemented by letters dated April 2, 2024 (ML24095A207), and May 6, 2023 (ML24129A200), Dominion Energy South Carolina, Inc.
(DESC), 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). DESC submitted the application pursuant to Title 10 of the Code of Federal Regulations part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for subsequent license renewal.
The NRC staff completed its aging management audit from November 6, 2023 - March 21, 2024, in accordance with the audit plan (ML23296A109). The audit report is enclosed.
If you have any questions, please contact me by email at Marieliz.Johnson@nrc.gov.
Sincerely,
/RA/
Marieliz Johnson, Project Manager License Renewal Projects Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation Docket No. 50-395
Enclosure:
Audit Report cc w/encl: Listserv
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 - REPORT FOR THE AGING MANAGEMENT AUDIT REGARDING THE SUBSEQUENT LICENSE RENEWAL APPLICATION REVIEW (EPID NO. L-2023-SLR-0003)
DATED: JUNE 25, 2024 DISTRIBUTION:
PUBLIC RidsNrrDnrl Resource RidsACRS_MailCTR Resource RidsNrrDorlLpl2 Resource RidsRgn2MailCenter Resource RidsRgn2Dnms Resource RidsRgn2Drp Resource RidsRgn2Drs Resource RidsRgn2Ora Resource RidsRgn2Opa Resource RidsNrrOd Resource RidsNrrPMSummer Resource MJohnson, NRR/DNRL LGibson, NRR/DNRL KConway, NMSS/REFS SKoenick, NMSS/REFS BSmith, NRR/DNRL SLee, NRR/DNRL EMiller, NRR/DORL DRoth, OGC TKeene, OEDO/AO SBurnell, HQ/OPA LWilkins, OCA DGasperson, RII/OPA CRead RII/DORS TMcGowan RII/DORS DDumbacher RII/DORS PCooper, RII/DORS SDowney, RII/DORS JPelchat, RII/FCO EMAIL:
Keith.J.Miller@dominionenergy.com ADAMS Accession Nos.:
PKG: ML24085A699 LTR: ML24177A138 (NONPROP)
ML24177A137 (PROP)
NRR-106 OFFICE PM:NRLP:DNRL BC:NPHP:DNRL BC:ELTB:DEX BC:SNSB:DSS BC:NCSG:DNRL NAME MJohnson MMitchell JPaige PSahd SBloom DATE 6/6/2024 03/05/2024 04/8/2024 04/23/2024 05/07/2024 OFFICE BC:ESEB:DEX BC:NVIB:DNRL BC:SFNB:DSS BC:NRLP:DNRL PM:NRLP:DNRL NAME ITseng ABuford SKrepel:
(NAmini for)
LGibson:
(BHarris for)
MJohnson DATE 06/6/2024 06/5/2024 06/5/2024 06/24/2024 06/25/2024
Enclosure Aging Management Audit Virgil C. Summer Nuclear Station, Unit No. 1 Subsequent License Renewal Application November 6, 2023 - March 21, 2024 Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
ii U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF NEW AND RENEWED LICENSES Docket No:
50-395 License No:
NPF-12 Licensee:
Dominion Energy South Carolina, Inc. (DESC)
Facility:
Virgil C. Summer Nuclear Station, Unit No. 1 Location:
Rockville, Maryland Jenkinsville, South Carolina Dates:
November 6, 2023 - March 21, 2024 Approved By:
Steven D. Bloom, Chief Corrosion and Steam Generator Branch Division of New and Renewed Licenses Angie R. Buford, Chief Vessels and Internals Branch Division of New and Renewed Licenses Ian Tseng, Chief Structural, Civil, Geotech Engineering Branch B Division of Engineering and External Hazards Lauren K. Gibson, Chief License Renewal Projects Branch Division of New and Renewed Licenses Jason Paige, Acting Chief Long Term Operations and Modernization Branch Division of Engineering and External Hazards Philip Sahd, Chief Nuclear Systems Performance Branch Division of Safety Systems Matthew A. Mitchell, Chief Piping and Head Penetrations Branch Division of New and Renewed Licenses Brian D. Wittick, Chief Containment & Plant Systems Branch Division of Safety Systems
iii Reviewers:
Name Organization Naeem Iqbal PRA Licensing B (APLB)/Division of Risk Assessment (DRA)
Adakou Foli Long Term Operations & Modernization Branch (ELTB)/Division of Engineering and External Hazards (DEX)
Matthew McConnell ELTB/DEX Jorge Cintron ELTB/DEX Liliana Ramadan Brian Correll ELTB/DEX ELTB/DEX Andrew Prinaris Structural, Civil, Geotech Engineering (ESEB)/DEX Dan Hoang ESEB/DEX Gurjendra Bedi ESEB/DEX Shaohua Lai ESEB/DEX Andrew Johnson Corrosion and Steam Generator (NCSG)/New and Renewed Licenses (DNRL)
Brian Allik NCSG/DNRL Gregory Makar NCSG/DNRL James Gavula Joel Jenkins NCSG/DNRL NCSG/DNRL Leslie Terry NCSG/DNRL Lydiana Alvarado NCSG/DNRL Matthew Yoder NCSG/DNRL Paul Klein Reena Boruk NCSG/DNRL NCSG/DNRL Tony Gardner NCSG/DNRL Ali Rezai Piping and Head Penetrations (NPHP)/DNRL Bart Fu Eric Reichelt NPHP/DNRL NPHP/DNRL Varoujan Kalikian Seung Min NPHP/DNRL NPHP/DNRL Carolyn Fairbanks Vessels and Internals (NVIB)/DNRL David Dijamco Emma Haywood Eric Palmer NVIB/DNRL NVIB/DNRL NVIB/DNRL James Medoff NVIB/DNRL On Yee NVIB/DNRL Steven Levitus NVIB/DNRL Angelo Stubbs Containment and Plant Systems (SCPB)/Division of Safety Systems (DSS)
Rao Karipineni SPCB/DSS Raul Hernandez SCPB/DSS Thang Thawn SCPB/DSS Jo Ambrosini Santosh Bhatt Nuclear Systems Performance (SNSB)/DSS SNSB/DSS Joe Messina Nuclear Methods and Fuels (SFNB)/DSS
iv Chris Tyree Marieliz Johnson License Renewal Projects Branch (NLRP)/DNRL NLRP/DNRL
v TABLE OF CONTENTS TABLE OF CONTENTS................................................................................................................ v ACRONYMS AND ABBREVIATIONS......................................................................................... viii
- 1. Introduction............................................................................................................................. 1
- 2. Audit Activities......................................................................................................................... 2 SLRA AMP B2.1.1, American Society of Mechanical Engineers (ASME)Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.......................................................... 2 SLRAAMP B2.1.2, Water Chemistry...................................................................................... 4 SLRA AMP B2.1.3, Reactor Head Closure Stud Bolting......................................................... 5 SLRA AMP B2.1.4, Boric Acid Corrosion................................................................................ 6 SLRA AMP B2.1.5, Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components......... 7 SLRA AMP B2.1.6, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS).................................................................................................................................... 8 SLRA AMP B2.1.7, Pressurized Water Reactor (PWR) Vessel Internals............................. 10 SLRA AMP B2.1.8, Flow-Accelerated Corrosion.................................................................. 18 SLRA AMP B2.1.9, Bolting Integrity...................................................................................... 19 SLRA AMP B2.1.10, Steam Generators............................................................................... 20 SLRA AMP B2.1.11, Open-Cycle Cooling Water System..................................................... 21 SLRA AMP B2.1.12, Closed Treated Water Systems........................................................... 21 SLRA AMP B2.1.13, Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems................................................................................................. 22 SLRA AMP B2.3.14, Compressed Air Monitoring................................................................. 23 SLRA AMP B2.1.15, Fire Protection..................................................................................... 24 SLRA AMP B2.1.16, Fire Water System............................................................................... 25 SLRA AMP B2.1.17, Outdoor and Large Atmospheric Metallic Storage Tanks.................... 27 SLRA AMP B2.1.18, Fuel Oil Chemistry............................................................................... 27 SLRA AMP B2.1.19, Reactor Vessel Material Surveillance.................................................. 28 SLRA AMP B2.1.20, One-Time Inspection........................................................................... 29 SLRA AMP B2.1.21, Selective Leaching.............................................................................. 29 SLRA AMP B2.1.22, ASME Code Class 1 Small-Bore Piping Program............................... 30 SLRA AMP B2.1.23, External Surfaces Monitoring of Mechanical Components.................. 31 SLRA AMP B2.1.24, Flux Thimble Tube Inspection............................................................. 31 Thimble Tube Thinning in Westinghouse Reactors.................................................................... 32 SLRA AMP B2.1.25, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components............................................................................................................. 33 SLRA AMP B2.1.26, Lube Oil Analysis................................................................................. 34
vi SLRAAMPB2.1.27, Monitoring of Neutron-Absorbing Materials Other than Boraflex......... 35 SLRA AMP B2.1.28, Buried and Underground Piping and Tanks........................................ 35 SLRA AMP B2.1.29, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks......................................................................... 36 SLRA AMP B2.1.30, ASME Section XI, Subsection IWE..................................................... 37 SLRA AMP B2.1.31, ASME Section XI, Subsection IWL...................................................... 39 SLRA AMP B2.1.32, ASME Section XI, Subsection IWF...................................................... 40 SLRA AMP B2.1.33, 10 CFR Part 50, Appendix J................................................................ 42 SLRA AMP B2.1.34, Masonry Walls..................................................................................... 43 SLRA AMP B2.1.35, Structures Monitoring.......................................................................... 44 SLRA AMP B2.1.36, Inspection of Water-Control Structures Associated with Nuclear Power Plants......................................................................................................................... 45 SLRA AMP B2.1.37, Protective Coating Monitoring and Maintenance................................. 45 SLRA AMP B2.1.38, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements................................... 46 SLRA AMP B2.1.39, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits........................................................................................................ 47 SLRA AMP B2.1.40, Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements................ 48 SLRA AMP B2.1.41, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements................ 50 SLRA AMP B2.1.42, Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR50.49 Environmental Qualification Requirements............................. 51 SLRA AMP B2.1.43, Fuse Holders....................................................................................... 53 SLRA AMP B2.1.44, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements........................................................................... 53 SLRA AMP B2.1.45, High-Voltage Insulators....................................................................... 54 SLRA AMP B3.1, Fatigue Monitoring.................................................................................... 55 SLRA AMP B3.2, Neutron Fluence Monitoring..................................................................... 57 SLRA AMP B3.3, Environmental Qualification of Electric Equipment................................... 57 SLRA AMP B3.4, Concrete Containment Unbonded Tendon Prestress............................... 58 SLRA TLAA Section 4.1, Identification of Time-Limited Aging Analyses.............................. 59 SLRA TLAA Section 4.2.1, Neutron Fluence Projections..................................................... 60 SLRA TLAA Sections 4.2.2, Upper-Shelf Energy to 4.2.6 Low Temperature Overpressure Protection....................................................................................................... 61 SLRA TLAA Section 4.3.1, Transient Cycle Projections for 80 years................................... 63 SLRA TLAA Section 4.3.2, ASME Code,Section III, Class 1 Fatigue Analyses.................. 64 SLRA TLAA Section 4.3.3, Non-Class 1 Allowable Stress Analyses.................................... 66
vii SLRA TLAA Section 4.3.4, Environmentally-Assisted Fatigue............................................. 67 SLRA TLAA Section 4.3.5, High-Energy Line Break Analysis.............................................. 69 SLRA TLAA Section 4.4, Environmental Qualification of Electric Equipment....................... 70 SLRA TLAA Section 4.5, Concrete Containment Tendon Prestress Analysis...................... 71 SLRA TLAA Section 4.6, Containment Liner Plate............................................................... 72 SLRA TLAA Section 4.7.1, Crane Load Cycle Limits............................................................ 72 SLRA TLAA Section 4.7.2, Reactor Coolant Pump Flywheel Fatigue Crack Growth Analyses................................................................................................................................ 73 SLRA TLAA Section 4.7.3, Leak-Before-Break (Plant-Specific TLAA)................................. 73 SLRA TLAA Section 4.7.4, Steam Generator Tube Wear Evaluation................................... 74 AMR Items Not Associated with an AMP.............................................................................. 75 SLRA AMR 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation......................................................................................................................... 75 SLRA AMR 3.6.2.2.2, Reduced Insulation Resistance Due to Age Degradation of Cable Bus Arrangements Caused by Intrusion of Moisture, Dust, Industrial Pollution, Rain, Ice, Photolysis, Ohmic Heating and Loss of Strength of Support Structures and Louvers of Cable Bus Arrangements Due to General Corrosion and Exposure to Air Outdoor............ 88 SLRA AMR 3.6.2.2.3, Loss of Material Due to Wind-Induced Abrasion, Loss of Conductor Strength Due to Corrosion, and Increased Resistance of Connection Due to Oxidation or Loss of Preload for Transmission Conductors, Switchyard Bus, and Connections............. 88
- 3. Supplements to the SLRA..................................................................................................... 89
- 4. Audit Questions Provided to DESC....................................................................................... 89
- 5. Applicant Personnel Contacted During the Audit.................................................................. 90
- 6. Exit Meeting.......................................................................................................................... 91
viii ACRONYMS AND ABBREVIATIONS ABPA Approved bolting pattern analysis AMP Aging Management Program AMR Aging management review AOR Analysis of Record ART Adjusted Reference Temperature ASME American Society of Mechanical Engineers ASR alkali-silica reaction BACP Boric Acid Control Program BFB Baffle-to-Former Bolts CAP Corrective action program CASS Cast Austenitic Stainless Steel CB Core barrel CRDM Control rod drive mechanism CRGT Control Rod Guide Tube CUF cumulative usage factor CUFen environmentally adjusted cumulative usage factor DESC Dominion Energy South Carolina, Inc.
DIR Design Input Request DWG Drawing EAF Environmental assisted fatigue EPRI Electric Power Research Institute EQ Environmental Qualification FAC Flow-Accelerated Corrosion FE Further evaluation FSAR Final Safety Analysis Report GALL-SLR NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal HELB high energy line break I&E Inspection and evaluation ISI inservice inspection LBB Leak-before-break LOCA Loss-of-coolant accident LR License renewal MRP Materials Reliability Program NRC Nuclear Regulatory Commission
ix NSSS Nuclear Steam Supply System OAR Owners Activity Report OE Operating experience PLEX Plant Life Extension PSW Primary Shield Wall PTS Pressurized Thermal Shock PWR pressurized-water reactor RAI Requests for additional information RCI Requests for confirmation of information RCP Reactor Coolant Pump RCS Reactor Coolant System RIVE Radiation induced volumetric expansion RPV Reactor pressure vessel RRVCH Replacement Reactor Vessel Closure Head RSC Reactor support concrete RTNDT reference temperature nil ductility RV Reactor vessel RVI reactor vessel internals SA expansion stress SLR Subsequent license renewal SLRA Subsequent license renewal application SPEO subsequent period of extended operation TLAA Time-Limited Aging Analysis UCP Upper core plate UFSAR Updated Final Safety Analysis Report UGW Upper girth weld USE Upper-Shelf Energy UT Ultrasonic testing VCSNS Virgil C. Summer Nuclear Station, Unit No. 1 WEC Westinghouse Electric Company WF Wide-flange
1 Report for the Aging Management Audit Virgil C. Summer Nuclear Station, Unit No. 1 Subsequent License Renewal Application
- 1. Introduction The U.S. Nuclear Regulatory Commission (NRC) staff conducted an aging management audit of Dominion Energy South Carolina, Inc. (DESC, the applicant). The audit focused on (1) plant-specific operating experience (OE), (2) methodology to identify the systems, structures, and components to be included within the scope of license renewal (LR) and subject to an aging management review (AMR) (Scoping and Screening Portion), and (3) aging management programs (AMPs), AMR items, Time-Limited Aging Analyses (TLAAs), and associated bases and documentation as applicable (AMP and TLAA Portion) for the subsequent license renewal (SLR) of Renewed Facility Operating License No. NPF 12 for Virgil C. Summer Nuclear Station, Unit No. 1 (V.C. Summer or VCSNS).
The purpose of the plant-specific OE portion of the audit is to identify examples of age-related degradation, as documented in the applicants corrective action program (CAP) database and to provide a basis for the staffs conclusions on the ability of the applicants proposed AMPs to manage the effects of aging in the period of extended operation. DESC searched their OE database and provided the results for the associated AMPs and TLAAs for NRC staff review.
Additional Word searches were performed by DESC, upon NRC staffs request, and the results were provided to the NRC staff for review.
The purpose of the Scoping and Screening portion of the audit is to evaluate the scoping and screening process as documented in the LR application, implementing procedures, reports, and drawings, such that the NRC staff:
obtain an understanding of the process used to identify the systems, structures, and components within the scope of LR and identify the structures and components subject to an AMR Have sufficient docketed information to allow the staff to reach a conclusion on the adequacy of the scoping and screening methodology as documented and applied The purpose of the AMP and TLAA Portion of the audit is to:
examine DESCs AMPs, AMR items, and TLAAs for V.C. Summer verify DESCs claims of consistency with the corresponding NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, issued in July 2017, AMPs, and AMR items assess the adequacy of the TLAAs Enhancements and exceptions will be evaluated on a case-by-case basis. The NRC staffs review of enhancements and exceptions will be documented in the safety evaluation (SE).
The regulatory basis for the audit was Title 10 of the Code of Federal Regulations (10 CFR)
Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants. The NRC staff also considered the guidance contained in NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR),
dated July 2017 and NUREG-2191. The SRP-SLR allows an applicant to reference in its LR application the AMPs described in the GALL-SLR Report. By referencing the GALL-SLR Report
2 AMPs, the applicant concludes that its AMPs correspond to those AMPs reviewed and approved in the report and that no further staff review is required. If an applicant credits an AMP for being consistent with a GALL-SLR Report program, it is incumbent on the applicant to make sure that the plant program contains all the elements of the referenced GALL-SLR Report program. The applicant should document this determination in an auditable form and maintain the documentation onsite.
- 2. Audit Activities A regulatory audit is a planned, license-related activity that includes the examination and evaluation of primarily non-docketed information. A regulatory audit is conducted to gain better understanding of an application, to verify information, and, if applicable, to identify information that will require docketing to support the NRC staffs conclusions that form the basis of the licensing or regulatory decision.
The subsequent license renewal application (SLRA) states that every AMP in the SLRA is consistent with the program elements of the GALL-SLR Report. To verify this claim of consistency, the staff audited each AMP, including any enhancements or exceptions associated with an AMP.
The SLRA discusses each TLAA, the disposition of the TLAA in accordance with 10 CFR 54.21(c)(1), and the basis for that disposition. To verify that the applicant provided a basis to support its disposition of the TLAA, the staff audited each TLAA.
The staff also audited AMR items not associated with an AMP to determine if the information in the SLRA is consistent with the further evaluation information in the SRP-SLR.
Furthermore, the staff audited the Final Safety Analysis Report (FSAR) descriptions for each AMP and TLAA for consistency with the SRP-SLR. During its audit, the staff interviewed the applicants staff and reviewed documentation contained in the SLRA and provided by the applicant via the ePortal.
For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and OE by reviewing documentation contained in the SLRA and ePortal.
Licensing conclusions or staff findings are not made in audit reports since licensing and regulatory decisions cannot be made solely based on an audit. Therefore, items identified but not resolved within the scope of the audit will be followed using other NRC processes, such as requests for additional information (RAIs), requests for confirmation of information (RCIs), and public meetings. Licensing conclusions, staff findings, staff review of enhancement and exceptions, and resolution of audit items will be documented in the staffs SE.
The following sections discuss the SLRA areas reviewed by the NRC staff.
SLRA AMP B2.1.1, American Society of Mechanical Engineers (ASME)Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Summary of Information in the Application. The SLRA states that AMP B2.1.1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, is an existing program with an enhancement that will be consistent with the program elements in GALL-SLR Report AMP XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.
3 At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3304 Subsequent License Renewal Project - Aging Management Program Evaluation Report - ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, V.C. Summer Nuclear Station Rev. 0, 07/2023 ETE-SLR-2023-33333 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -Operating Experience for AMPs Rev. 0, 07/26/2023 N/A VCS Operating Experience Detail Report: ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD 06/28/2023 VCSU-1183 and VCSP-952 Pressurizer Surge Line Hot Leg Weld 4500A, UT and PT 03/31/1993 WO: 0507566 Measure Diameter of Corroded C RCP Main Flange Bolt 05/21/2005 CR-15-04026 A Relief Request is Needed in Order to Notify NRC of 6 Weld Exams From the 3rd ISI Interval Which Were Not Able to Meet >90% Coverage 08/27/2015 ER-AA-NDE-VT-601 Dominion Energy Nuclear Fleet Nondestructive Examination Procedure VT-1 Visual Examination Procedure Rev. 5, 02/26/2019 ER-AA-NDE-VT-602 Dominion Energy Nuclear Fleet Nondestructive Examination Procedure VT-2 Visual Examination Procedure Rev. 6, 02/26/2019 ER-AA-NDE-VT-603 Dominion Energy Nuclear Fleet Nondestructive Examination Procedure VT-3 Visual Examination Procedure Rev. 7, 02/26/2019 ER-AA-NDE-UT-802 Ultrasonic Examination of Austenitic Piping Welds in Accordance with ASME Section XI, Appendix VIII Rev. 8, 03/03/2021 WCAP-18790-P V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Charging Line Nozzle and Hot Leg Surge Line Nozzle Flaw Tolerance Evaluation Utilizing ASME Section XI Appendix L Methodology Rev. 3.
02/2024 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP not associated with the enhancement are consistent with the corresponding elements of the GALL-SLR Report AMP. During the audit, the staff made the following observations:
SLRA Section B2.1.1 describes the applicants enhancement related to flaw tolerance evaluations performed on the normal and alternate charging cold leg nozzles and pressurizer surge line hot leg nozzle, which are the limiting environmental assisted fatigue (EAF) locations, in accordance with ASME Code Section XI, Appendix L.
These evaluations also are addressed as EAF TLAAs in SLRA Section 4.3.4. The staff noted that the SLRA does not clearly address the following items:
4
- 1. whether the normal and alternate charging nozzle welds that are subject to flaw tolerance evaluations are inspected by using ultrasonic testing examination
- 2. whether these nozzles are included in the examination items for thermal fatigue in accordance with the risk-informed inservice inspection program based on ASME Code Case N-716-1
- 3. whether a confirmatory or baseline inspection for the flaw tolerance evaluations will be performed to confirm the absence of an unacceptable flaw per Appendix L, L-3410 during the last inspection interval prior to the subsequent period of extended operation
- 4. whether the inspection frequencies are at least once every 10-year inservice inspection interval in accordance with the ASME Code Section XI, Appendix L provisions (L-3420, Successive Examinations) and, if not, whether these nozzles will be inspected at least once during the subsequent period of extended operation The NRC staff also audited the description of the SLRA AMP B2.1.1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, provided in the FSAR supplement.
The staff found that the available information was insufficient determine whether the description provided in the FSAR supplement was an adequate description of the SLRA AMP.
The staff will consider issuing an RAI to obtain the information necessary to verify the sufficiency of the FSAR supplement program description.
SLRAAMP B2.1.2, Water Chemistry Summary of Information in the Application. The SLRA states that AMP B2.1.2, Water Chemistry, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M2, Water Chemistry, as modified by SLR-ISG-2021 MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance, dated February 2021 (ML20181A434).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date SLR-302-772 Virgil C. Summer Nuclear Station Subsequent License Renewal Nuclear Sampling System Rev. 0 ETE-SLR-2023-3305 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Water Chemistry Rev. 0 CP-613 Steam Generator Chemistry Control Rev. 23, Change F CP-618 Chemistry Specifications for Borated Systems and Tanks Rev. 10, Change I CP-619 Chemistry Specifications for Makeup Water Tanks Rev. 13, Change E CP-614 Reactor Coolant Chemistry Control Rev. 21 CP-615 Feedwater and Condensate Chemistry Control Rev. 21, Change I CY-AP-SEC-200 Secondary System Chemistry Rev. 10 N/A Secondary Chemistry Strategic Water Plan September 2020 N/A Primary Chemistry Strategic Water Plan October 2020 CY-AP-PRI-100 Primary System Chemistry Rev. 11 CR-15-00249 High Feedwater Iron Concentration due to Plant Transient 1/15/2015 CR-15-00867 Low Feedwater Hydrazine Concentration 2/24/2015 CR-15-00871 Elevated MPA Concentration in the Steam Generators 2/25/2015
5 Document Title Revision/Date CR-15-00873 Elevated Feedwater MPA Concentration 2/25/2015 CR-15-00882 Incorrect Chemical Used to Batch the Chemical Addition Tank 2/25/2015 CR-16-02704 Reactor Coolant System (RCS) Lithium Concentration Error due to Intermittent Instrument Malfunction 5/27/2016 CR-17-01593 Elevated Steam Generator Chloride Concentrations 4/6/2017 CR-17-01558 Increasing Trend in Chloride Concentration in the Steam Generators 4/5/2017 CR-18-05546 Ineffective Condensate Polisher Resin Allows Sodium Pass Through 11/26/2018 EPRI Report No.
3002000505 EPRI Proprietary. Pressurized Water Reactor Primary Water Chemistry Guidelines Rev. 7 EPRI Report No.
3002010645 EPRI Proprietary. Pressurized Water Reactor Secondary Water Chemistry Guidelines Rev. 8 During the audit, the NRC staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA AMP Water Chemistry provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.3, Reactor Head Closure Stud Bolting Summary of Information in the Application. The SLRA states that AMP XI.M3, Reactor Head Closure Stud Bolting is an existing program with enhancement and exception that is consistent with the program elements in GALL-SLR Report AMP XI.M3, Reactor Head Closure Stud Bolting.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3306 Subsequent License Renewal Project Aging Management Program Evaluation Report - Reactor Head Closure Stud Bolting V.C. Summer Nuclear Station Rev. 0 ISE-7 Fourth Interval Inservice Inspection Program Plan for V.C. Summer Nuclear Station Rev. 3D ML21173A163 Virgil C. Summer Nuclear Station, Unit 1 - Proposed Alternative Request Rr-4-25 To Eliminate Examination Of Threads In Reactor Pressure Vessel Flange (EPID L-2020-LLR-0141) 08/06/2021 RF26 Work order WO1907241014 for UT of RV head closure studs 01-58.pdf UT Bolting/Stud Examination summary sheet 10/25/21 ETE-SLR-2023-3333 V. C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Operating Experience for AMPs Rev. 0
6 tbl_OE Search Terms.pdf N/A N/A During the audit, the NRC staff verified the applicants claim that the scope of program, parameters monitored or inspected, detection of aging effects, monitoring and trending, and acceptance criteria program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also verified the applicants claim that aspects of the preventive actions and corrective actions program elements not associated with the exception identified in the SLRA are consistent with the corresponding program elements inthe GALL-SLR Report AMP.
The staff also audited the description of the SLRA AMP Reactor Head Closure Stud Bolting FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.4, Boric Acid Corrosion Summary of Information in the Application. The SLRA states that AMP B2.1.4, Boric Acid Corrosion, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M10, Boric Acid Corrosion.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3307 Aging Management Program Evaluation Report - Boric Acid Corrosion Rev. 0 ER-AP-BAC-10 Boric Acid Control Program Rev. 15 ER-AP-BAC-101 Boric Acid Control Program (BACP) Inspections Rev. 16 ER-AP-BAC-102 Boric Acid Control Program (BACP) Evaluations Rev. 17 ER-AA-101 System Engineering Walkdowns Rev. 7 QSP-216 Boric Acid Corrosion Inspections Rev. 4 CR-11-04722 Condition Report on Boric Acid Leakage, Safety Injection System Coupling 09/14/2011 CR-12-05013 Condition Report on Boric Acid Leakage, Chemical and Volume Control System, Alternate Seal Injection PD Pump 10/31/2012 CR-13-01301 Condition Report on Boric Acid Leakage, Chemical and Volume Control System, Excess Letdown Heat Exchanger 03/23/2013 CR-14-01722 Condition Report on Boric Acid Leakage, Steam Generator Manways 04/08/2014 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Boric Acid Corrosion program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
7 SLRA AMP B2.1.5, Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components Summary of Information in the Application. The SLRA states that AMP B2.1.5, Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M11B, Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-induced Corrosion in Reactor Coolant Pressure Boundary Components.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3308 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Cracking of Nickel-Alloy Components and Loss of Material due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components Rev. 0 ETE-CEP-2012-1002 Identification and Evaluation of RCS [Reactor Coolant System]
Alloy 600 Locations - Pressure Boundary Locations and Reactor Internals Rev. 1, 12/08/2021 LMT-VCS_RF26-001, RF26 ISI Final Report ISI [Inservice Inspection] Final Report, V.C. Summer 1RF26 October 2021 10/28/2021 RF26 ISI Final Report Excerpts (Summary No.
VCS-1-102113)
Reactor Vessel Upper Head Bare Metal Visual Examination 10/28/2021 RF26 Framatome UT Report (Document No.
180-9338743-000)
V.C. Summer R026 Reactor Pressure Vessel 10 Year ISI and Nozzle-to-Safe End Dissimilar Metal Welds Final Report 10/30/2021 RF24 ISI Final Report Excerpts (WDI-PJF-1317864-FSR-001)
ISI Final Report 10/2018 RF25 ISI Report Excerpts (WDI-PJF-1332732-FSR-001)
ISI Final Report This report contains examination of welds overlaid.
04/2020 RF19 ISI Report Excerpts Nineteenth Refueling Inservice Inspection (ISI) Report This document contains examination of welds overlaid.
08/26/2011 CR-13-04467 and CR-13-00796 These documents confirm that the applicant has reviewed industry guidance of Electric Power Research Institute (EPRI) Report 3002000091, Nondestructive Evaluation: Guideline for Conducting Ultrasonic Examinations of Dissimilar Metal Welds, Rev. 1, and EPRI Reports 3002000041, Nondestructive Evaluation Improvement Focus Group Extent of Condition Actions in Response to North Anna Dissimilar Metal Weld Operating Experience, Rev. 1.
12/16/2014 PIR1010500, PIR1138965 These documents confirm that the applicant has reviewed industry guidance of Materials Reliability Program (MRP)-384, Guideline for Nondestructive Examination of Reactor Vessel Upper Head Penetrations, for applicability to V.C. Summer.
05/09/2017 WCAP-16388-P PWSCC [Primary Water Stress Corrosion Cracking] Susceptibility Assessment of the Alloy 600 and Alloy 82/182 Components in V.C.
Summer Nuclear Station 03/2005 Letter MRP 2019-035 Publishing Materials Reliability Program: Guideline for Nondestructive Examination of Reactor Vessel Upper Head Penetrations, Rev. 1 (MRP-384) 12/12/2019
8 Document Title Revision/Date rpt_AMP_Report M11B VCS Operating Experience Detail Report This report confirms the applicant has reviewed NRC Regulatory Issued Summary (RIS) 2018-06, Clarification of the Requirements for Reactor Pressure Vessel Upper Head Bare Metal Visual Examinations, dated December 2018 for applicability to V.C.
Summer.
06/19/2023 ER-AA-MAT-11 Alloy 600 Management Plan Rev. 18 00-01392-073_10 01 02 SER-SC_DGW-06 2006 Safety Evaluation of Flaws Detected in V.C. Summer Nozzle-to-Pipe Welds in the Hot Legs of Loops B and C (TAC No. MB4870) dated October 1, 2002 (ML022740071) 10/01/2002 LER-2000-008-01 Licensee Event Report (LER) 2000-008-01 Reactor Coolant System Pressure Boundary Degradation 03/15/2001 00E5BC / Root Cause Report C-00-1392 V.C. Summer Nuclear Station Root Cause Investigation A Hot Leg Nozzle Weld Cracks 2/15/2001 00-01392-093_MSIP SER Request-SC_DGW-06-08-2006 Inspection / Mechanical Stress Improvement (MSIP) Plan for VCSNS [V.C. Summer Nuclear Station] Hot Leg Welds 04/17/2002 DC05600-062 Crack Growth Evaluation of Pressurizer Surge Nozzle with Weld Overlay Repair Rev. 1, 04/17/2008 During the audit, the NRC staff verified applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
The original V.C. Summer Alloy 600/82/182 reactor vessel closure head was replaced by a head made with Alloy 690/52/152 materials in the spring of 2017.
The applicant confirmed that the volumetric examination performed in the fall of 2021 refueling outage on the reactor vessel hot leg nozzle-to-pipe dissimilar metal welds B and C mitigated with the Mechanical Stress Improvement Process showed no evidence of growth of the existing cracks and no presence of new cracks.
The applicant confirmed that V.C. Summer does not have nickel-alloy branch line connections susceptible to the primary water stress corrosion cracking.
During interview of the applicant staff, the applicant confirmed that Note 10 of Table 1 of 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 ActivitiesSection XI, as mandated by 10 CFR 50.55a(g)(6)(ii)(F) is met for the V.C. Summer full structural weld overlaid dissimilar metal butt welds.
The staff also audited the description of the SLRA Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-induced Corrosion in Reactor Coolant Pressure Boundary Components provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the SRP-SLR.
SLRA AMP B2.1.6, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)
Summary of Information in the Application. The SLRA states that AMP B2.1.6, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS), is an existing program that is
9 consistent with the program elements in GALL-SLR Report AMP XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3309 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)
Rev. 0, 07/2023 CGE-PP000-TR-CF-000001-P V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Thermal Aging Assessment of Cast Austenitic Stainless Steel Piping Components in the Reactor Coolant Loop and Accumulator Nozzles Rev. 2, 03/17/2022 CGE-MP010-TR-CF-000003-P V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: ASME Code Case N-481 Reactor Coolant Pump Casing Flaw Evaluation Rev. 2, 09/2022 PWROG-17033-P Updated for Subsequent License Renewal: WCAP-13045, Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems Rev. 1, 06/2018 Safety Evaluation -
Code Case N-481 Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report PWROG-17033, Rev. 1, Update for Subsequent License Renewal: WCAP-13045, Compliance to ASME Code Case N-481 of the Primary Loop Pump Castings of Westinghouse Type Nuclear Steam Supply Systems 09/26/2019 WCAP-13045 Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems 09/1991 CR-12-01671 /
rpt_AMP_Report M12 VCS Operating Experience Detail Report This report discusses V.C. Summer review of NUREG/CR-7122, An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds, and use of its guidance for volumetric examination of CASS in V.C. Summer.
04/23/2012 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
The applicant identified a total of 18 CASS pipe fittings (i.e., five elbows and one safety injection accumulator nozzle for each loop) within the reactor coolant loops, hot leg piping, crossover piping, and cold leg piping in V.C. Summer that are made of SA351 Grade CF8A steel, statically cast, with low molybdenum content, and less than 20 percent ferrite content. Based on the criteria in GALL-SLR Report AMP XI.M12, the applicant determined that the thermal aging embrittlement of the CASS pipe fittings is not significant, and therefore no additional inspection or evaluation is required to demonstrate that the material has adequate fracture toughness.
10 The applicant performed the ultrasonic testing (UT) of the welds attaching the above CASS elbows to the reactor vessel inlet nozzle safe ends in the fall months of 2003, 2009, 2015, and 2021 refueling outages. No recordable indications were identified in the weld volume examined.
The applicant identified a total of three CASS reactor coolant pump casings that are made of SA-351 Grade CF8 steel, statically cast, with low molybdenum content, and less than 20 percent ferrite content. Based on the criteria in GALL-SLR Report AMP XI.M12, the applicant determined that the thermal aging embrittlement of the CASS pump casings is not significant, and therefore no additional inspection or evaluation is required to demonstrate that the material has adequate fracture toughness.
During the audit, the applicant confirmed that it is aware of NUREG/CR-4513, Rev. 2 with errata, Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems, dated March 15, 2021.
The NRC staff also audited the description of the SLRA Thermal Aging Embrittlement of CASS provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.7, Pressurized Water Reactor (PWR) Vessel Internals Summary of Information in the Application. The SLRA states that AMP B2.1.7, PWR Vessel Internals, is an existing program that, with enhancements, will be consistent with the program elements in GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, as updated in Appendix D of Interim Staff Guidance Document No. SLR-ISG-2021-01-PWRVI (ML20217L203). At the time of the audit, the applicant had not yet fully developed the procedural documents necessary to implement this enhanced program, and the staffs audit addressed only the applicants basis document record for the AMP, as supplemented by those industry reports (e.g., those developed and issued the Electric Power Research Institute [EPRI] Materials Reliability Program [MRP] or by Westinghouse Electric Company [WEC] as the nuclear steam supply system vendor for the V.C. Summer unit) that were used to develop the basis document record for the AMP or the AMP gap analysis.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of the AMP in the SE.
Document Title Revision/Date Dominion Energy Engineering Technical Evaluation Document No.
ETE-SLR-2023-3310 V.C. Summer Subsequent License Renewal Project -
Aging Management Evaluation Report - PWR Vessel Internals (henceforth referred to as the AMP basis document)
Rev. 0 VCSNS M16A Operating Experience (OE) Listing VCS Operating Experience List Report June 28, 2023 VCSNS ECR No. 50704 Plant Modification Package: Reactor Vessel Upflow Conversion Rev. 5, February 5, 2008 VCSNS ECR No. 50868 Plant Modification Package: Replacement Reactor Vessel Closure Head (RRVCH)
Rev. 5, June 14, 2016 VCSNS Design Engineering Specification No. SP-1006 Replacement Reactor Vessel Closure Head for V.C Summer Nuclear Station Unit 1 January 22, 2014
11 Document Title Revision/Date VCSNS Document No.
CGE-RV010-TR-CA-00001 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Vessel Internals Gap Analysis Inputs (Record includes site-specific proprietary gap analysis inputs for all V.C. Summer PWR RVI components that are within the scope of SLRA AMP B2.1.7, PWR Vesse Internals)
Rev. 1 (proprietary)
VCSNS Document 16 VC Summer SLRA Review - PWR Vessel Internals -
Revised Responses 2nd Round (pdf Record provides Dominion Energys revised responses to the staff audit breakout session topics for SLRA AMP B2.1.7, PWR Vessel Internals, as of the completion of the second round of audit breakout session discussions for the AMP)
Rev. 1 EPRI MRP Non-Proprietary Report No.
3002017168 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 1-A)
Rev. 1-A (ML20175A112)
MRP 2018-022 Transmittal of MRP-191-SLR Screening, Ranking, and Categorization Results and Interim Guidance in Support of Subsequent License Renewal at U.S. Power Plants August 31, 2018 (ML19081A061)
EPRI MRP Proprietary Topical Report No.
3002018248 Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230, Rev. 3)
July 2020 (ML20244A026, proprietary)
EPRI MRP Proprietary Topical Report No.
3002013220 Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals Component for Westinghouse and Combustion Engineering PWR Design (MRP-191, Rev. 2)
November 2018 (ML19081A063, proprietary)
MRP 2023-005 MRP 227 NEI 03-08 Needed Interim Guidance for WEC/CE Core Barrel Inspections May 19, 2023 (ML23290A020)
MRP 2023-014 Transmittal of NEI-03-08 Needed Interim Guidance Regarding PWR Core Barrel Inspections as Defined in EPRI Guidance MRP-227 (Docket No. 99902021)
October 16, 2023 (ML23290A019)
MRP 2017-009 Transmittal of NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev.1 March 15, 2017 (ML17087A106)
EPRI MRP Non-Proprietary Topical Report No. 3002020105 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 2)
September 2021 (ML22129A141)
MRP 2018-002 Transmittal of NEI-03-08 Needed Interim Guidance Regarding MRP-227-A and MRP-227, Rev. 1 Baffle-Former Bolt Expansion Inspection Requirements for PWR Plants January 17, 2018 NRC Interim Staff Guidance Document No.
SLR-ISG-2021 PWRVI Updated Aging Management Criteria for Reactor Vesse Internal Components for Pressurized Water Reactors January 2021 (ML20217L203)
WEC Proprietary Class 3 Topical Report No.
WCAP-17451-P Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections Rev. 2, November 2018 Westinghouse Non-Proprietary Class 3 Technical Bulletin Reactor Vessel Head Adapter Thermal Sleeve Wear Rev. 2, February 18, 2015
12 Document Title Revision/Date No. TB-07-2 During the audit, the NRC staff verified the applicants claim that the preventative actions, parameters monitored or inspected, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of GALL-SLR AMP XIM16A, as updated in SLR-ISG-2021-01-PWRVI or will be consistent after implementation of the identified enhancements.
In addition, for the scope of program, detection of aging effects, monitoring and trending, and acceptance criteria program elements, the staff noted that sufficient information was not available at the time of the audit to determine whether the program elements were consistent with the corresponding program elements of the GALL-SLR Report AMP. As discussed in the audit observations that follow, the staffs observations on the consistency of these program elements with the corresponding program elements in GALL-SLR AMP XI.M16 pertain to (1) the defined criteria in Enhancements 1 and 2 of the AMP and (2) the staffs assessment on whether the criteria of these enhancements will make the program elements of SLRA AMP B2.1.7, PWR Vessel Internals. consistent with the program elements in GALL-SLR AMP XI.M16A.
The staff may consider issuing an RAI to obtain the information necessary to verify whether these program elements, when subject to actions in Enhancements 1 and 2, will be consistent with the corresponding program elements of GALL-SLR Report AMP XI.M16A.
During the audit, the staff made the following observations regarding SLRA AMP B2.1.7 and the program element criteria that apply to the AMP:
Program Enhancement 1 - Identified in SLRA AMP B2.1.7 as impacting AMP detection of aging effects, monitoring and trending, and acceptance criteria program elements.
Enhancement 1 of the AMP states: Procedure(s) will be revised to include a list of components that require inspections for the Primary, Expansion, and Existing Program categories specified in MRP-227, Rev. 1-A. The staff noted that the "NUREG-2191 Consistency statement in SLRA AMP B2.1.7 and the AMP basis document provided in Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 indicate that the AMP is consistent with version of GALL-SLR AMP XI.M16A, and is based on the component-specific inspection categorizations and inspection and evaluation (I&E) criteria in EPRI Report MRP-227, Rev. 1-A, as modified by the results of the applicants reactor vessel internals (RVI) gap analysis. The staff observed that the basis in Enhancement 1 could be inconsistent with the NUREG-2191 Consistency statement in SLRA AMP B2.1.7 because the enhancement only calls for the AMP procedure(s) to be revised for inclusion of Primary, Expansion, and Existing Program components as defined in MRP-227, Rev. 1-A, but where the criteria in MRP-227, Rev. 1-A are being subject to the results of the applicants AMP RVI gap analysis. The staff also observed that AMP Enhancement 1 may be impacting the scope of program element criteria of the AMP.
The staff will evaluate the adequacy of AMP Enhancement 1 in the SE section that summarizes the staffs review of the AMP. The staff may consider an RAI relative to the applicants basis for AMP Enhancement 1.
Program Enhancement 2 - Impacting AMP parameters monitored and inspected, detection of aging effects, and acceptance criteria program element criteria for the RVI hold-down spring. Enhancement 2 of the AMP states that the implementation procedure will be revised to require additional spring height measurements prior to the SPEO to establish the core barrel hold-down spring height and determine if replacement of the core
13 barrel hold-down spring is required. The staff observed that this baseline measurement basis relates to the applicants activities for managing potential changes in dimension that may occur in the hold-down spring configuration and is consistent with EPRI MRPs criteria for implementing a second round of hold-down spring baseline inspections during the subsequent period of extended operations, as recommended in the MRP 2018-022 letter (ML19081A061). The NRC staff also observed that the need for performing the additional physical measurements of the hold-down spring is defined and established in the gap analysis summary that is included in the basis document for the AMP (i.e., in Dominion Energy Record No. ETE-SLR-2023-3310).
Industry Records Used to Develop the RVI Gap Analysis - Impacting the scope of program element of the AMP. SLRA AMP B2.1.7and Sections 2.4 and 3.1.2 of the AMP basis document (Dominion Energy Record No. ETE-SLR-2023-3310) indicate that the plant-specific RVI gap analysis of the AMP integrates and relies on the interim guidelines in MRP 2018-022 as one of the industry interim guidance documents relied on for the gap analysis of the AMP. The staff observed that the AMP program discussion and the discussion in Section 2.4 of Technical Evaluation Document No. ETE-SLR-2023-3310 also indicate that the AMP uses the functionality assessment in MRP-230, Rev. 3 (ML20244A026) and the RVI component-specific screening bases in MRP-191, Rev. 2 (ML19081A063). Thus, the staff observed that the applicant may be relying on additional EPRI MRP or industry reports for the development and results of the VCSNS RVI gap analysis from those identified for derivation and development of the gap analysis results in SLRA AMP B2.1.7.
During the audit review period, the applicant provided VCSNS Document 16, VS Summer SLRA Review - PWR Vessel Internals - Revised Responses 2nd Round, onto the audit portal for the SLRA and provided audit portal documentation that the RVI gap analysis also relies on the industry guidance in MRP 2017-009 and MRP 2018-002 (i.e., in addition to use of MRP 2018-022 and WCAP-17451-P, Rev. 2 for development of the RVI gap analysis). The staff will address this topic as part of the staffs review of the RVI gap analysis in the SE.
Industry Report Relied on for Inspection of the RVI Control Rod Guide Tube (CRGT)
Support Plates - Impacting the scope of program element of the AMP.
SLRA AMP B2.1.7 and the AMP basis document (Dominion Energy Record No. ETE-SLR-2023-3310) reference the augmented I&E criteria in WEC Class 3 Proprietary Report No.
WCAP-17451-P, Rev. 2 as the RVI gap analysis basis for performing inspections of the CRGT support plates (CRGT guide cards), where the WCAP establishes the basis for inspecting a proprietary subpopulation of CRGT guide cards during the subsequent period of extended operation. In contrast to the referenced I&E criteria in WCAP-17451-P, Rev.
2, staff observed that the I&E criteria in WCAP-17451-P, Rev. 1 criteria provide the current staff-approved condition monitoring bases for inspecting WEC-design CRGT guide cards, as evaluated and accepted for implementation in Section 3.1.3.7 of the staffs April 25, 2019, SE (ML19081A001) for MRP-227, Rev. 1. The staff will evaluate which version of WCAP-17451-P should be applied for CRGT guide card inspections as part of the staffs review of AMP Enhancement 1 and the AMPs gap analysis basis in the SE. The staff may consider an RAI relative to the applicants basis for using WCAP-17451-P, Rev. 2 to perform augmented inspections of the CRGT guide cards or CRGT subassembly components that are identified in WCAP-17451-P, Rev. 2 as Expansion category components for the guide cards.
14 Acceptance Criteria Used to Evaluate Clustered Groupings of Baffle-to-Former Bolts (BFBs) with Detected Flaw Indications - Impacting the acceptance criteria element of the AMP. The AMP basis document (Dominion Energy Record No. ETE-SLR-2023-3310) provides the applicants basis for implementing sample-expansion inspections of the units core barrel-to-former bolts if large cluster groupings of BFBs (in close proximity to one another) are detected with relevant flaw indications. The NRC staff noted that the AMP basis document defines a large cluster of degraded baffle-former bolts is defined as any group of adjacent baffle-former bolts at least 3 rows high by at least 10 columns wide, or at least 4 rows high by at least 6 columns wide where 80% or greater of the baffle-former-bolts have unacceptable UT indications or are visibly degraded, and that the basis for this definition is established in MRP-2018-002 and in Note 3 of Table 5-3 in the MRP-227, Rev. 2 report [ML22129A141]). The staff also observed that the above definition for large clustered groupings of defective BFBs is different from that defined for large clustered groupings of defective BFBs in EPRI MRP Letter No. MRP 2017-009 (ML17087A106),
which establishes a large clustered grouping as three or more adjacent defective BFBs or more than 40% defective BFBs on the same baffle plate, and which was accepted in Section 3.1.3.7 of the staffs April 25, 2019, SE (ML19081A001) for MRP-227, Rev. 1.
The staff will review the AMPs acceptance criteria element definition for large, clustered groupings of defective BFBs as part of the staffs basis for evaluating the adequacy of SRLA AMP B2.1.7 Enhancement 1 and the RVI gap analysis basis for BFBs in the SE.
The staff may consider an RAI relative to which version of the definition should be applied to the BFBs under the RVI gap analysis of the AMP.
Tier Category and operating experience (OE) for VCSNS BFBs - Impacting the scope of program, monitoring and trending, and operating experience program element criteria applying to the BFBs in the AMP. VCSNS Design Modification Record (VCSCS ECR) No.
50704, Plant Modification Package: Reactor Vessel Upflow Conversion (as posted to the audit portal) provides and supports the applicants basis for inspecting the VCSNS BFBs as Primary components for the program. The staff observed that the technical information in VCSCS ECR No. 50704 provides the applicants design change report for the Fall 2009 modification of the core barrel (CB) assembly that was implemented to establish reactor coolant upflow conditions through reactor core. The staff also observed that the AMP basis document (Dominion Energy Record No. ETE-SLR-2023-3310) establishes that the applicant performed augmented ultrasonic inspections of the BFBs in 2021 and that the inspections of the BFBs did not reveal flaw indications in more than seven of the total population of examined accessible BFBs, or any evidence of clustered groupings of BFBs or significant population of BFBs (less than 5 percent population of inspected BFBs) with detected indications, that otherwise (if exceeded) would trigger sample-expansion of the inspections to the lower support column bolts or core barrel-to-former bolts in the unit.
Based on EPRI MRPs tier-based categories for WEC-design BFBs in MRP 2017-009 and the applicants relevant site-specific inspection results for BFB inspections in 2021, the staff observes that the relevant information may support the applicants basis for inspecting the VCSNS BFBs in accordance with inspection schedule for Tier 3 WEC-design BFB inspections in MRP 2017-009. Since MRP 2017-009 addresses the industrys generic basis for dispositioning OE with WEC-design BFBs and for updating the Primary category inspection bases for WEC-design BFBs, the staff will evaluate the impacts of the VCSNS BFB inspection results as part of the staffs inspection frequency and OE review of the BFBs in the SE.
Regarding the acceptance criteria element of the PWR Vessel Internals Program.
15 Section 3.6.2 of the AMP basis document (Dominion Energy Record No. ETE-SLR-2023-3310) addresses acceptance criteria and standards for evaluating the structural integrity of baffle plates containing a subpopulation of VCSNS BFBs (Primary bolts for inspection per Item W6 in Table 4-3 of MRP-227, Rev. 1-A) with detected flaw indications. The staff observed that the AMP basis document allows the acceptance criteria for demonstrating the structural of the BFBs and the corresponding baffle plates containing the bolts to be performed using an approved bolting pattern analysis (ABPA). The staff observed that use of ABPAs as an alternate aging management acceptance criterion basis for evaluating the structural integrity bases for RVI bolted assemblies was accepted in the staffs July 4, 2023, SE (ML23205A153) for WCAP-17096-NP, Rev. 2. However, in the July 4, 2023, SE, the staff issued the following condition on use of ABPAs as an alternate acceptance criterion basis for PWR RVI bolted assemblies:
For bolted assembly analysis methods that use predictive, probabilistic ABPA models, the models and methods should appropriately account for limits on the number of allowable bolts with presumed failed conditions, geometric bolt failure considerations (i.e., bolt clustering considerations), and bolting analysis reliability considerations (i.e., addressing 95% reliability confidence limits). These models and methods should be appropriately documented and justified in a site-specific or owner defined record.
During the audit period, the applicant provided VCSNS Document 16, VS Summer SLRA Review - PWR Vessel Internals - Revised Responses 2nd Round, onto the audit portal for the SLRA and provided audit portal documentation that the applicant does not have any current plans for use of an ABPA as an alternative flaw evaluation strategy and acceptance criteria basis for evaluating the structural integrity of RVI bolted assemblies containing a subpopulation of bolts with detect flaw indications in the bolts. However, the applicant indicated that, if an ABPA is used as an alternative flaw evaluation strategy, then the application of the ABPA would conform to the staffs condition on the use of ABPAs, as identified in Condition #C-02 in the staffs July 4, 2023, SE for WCAP-17096. Because the applicant has indicated that any potential use of ABPAs will address staff Condition #C-02, the matter is closed for the audit, and the staff will not need to address this matter in the SE.
Regarding potential operating experience program element impacts on the scope of program, detection of aging effects, monitoring and trending, and acceptance criteria program elements of the PWR Vessel Internals Program. SLRA AMP B2.1.7 and basis document for the AMP in Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 establish that the upper girth weld (UGW) in the CB assembly is one of the linked Expansion category components for augmented inspections that will be performed on the Primary category CB upper flange weld during the subsequent period of extended operation. The staff observed that AMP does not account for recent OE with WEC-design CB UGW cracking that is documented in EPRI MRP Letter Nos. MRP 2023-005 and in MRP 2023-014, where the EPRI MRP may be recommending elevation of the CB UGW from designation as an Expansion category weld in the program to a designated Primary category weld of the program.
The staff observed that the information provided in VCSNS Document 16, VS Summer SLRA Review - PWR Vessel Internals - Revised Responses 2nd Round, provides additional information to address how Dominion would address the referenced generic OE with WEC-design CB UGW cracking and acknowledged the applicants awareness of the interim guidance in the MRP 2023-005 letter. The staff noted that the VCSNS Document 16 records support the applicants plan to elevate the VCSNS CB UGW to
16 a Primary category component for inspection if the MRP 2023-005 interim guidance recommendations for the weld type are incorporated into a staff-approved version of MRP-227 that covers the 80-year period, and the CB UGW is elevated to Primary category status. The staff observed that the potential amendment of the enhancement in AMP B2.1.7 Enhancement 1 may rectify which inspection category (i.e., Primary or Expansion category) should be applied to the CB UGW during the subsequent period of extended operation. Thus, the staff will review the adequacy of the EPRI MRP inspection category being assigned to the CB UGW as part of the staffs basis for evaluating the adequacy of SRLA AMP B2.1.7 Enhancement 1 in the SE and as part of the staffs OE review of the AMP in the SE.
Regarding the scope of program element of the PWR Vessel Internals Program.
The staff confirmed that Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 indicates that the PWR RVI design at VCSNS does not include RVI thermal shield assembly components (including thermal shield flexures) and that the RVI design instead relies on the RVI neutron panels as the RVI components that serve neutron irradiation mitigation function for the reactor pressure vessel (RPV) interior surfaces and materials. The staff also observed and confirmed that the neutron panels are No Additional Measures category components for the program under the provision of MRP-227, Rev. 1-A and MRP 2018-022. Thus, the staff confirmed that generic OE associated with WEC-design thermal shield flexure cracking does not need to be included within the scope of the staffs OE review of the AMP in the SE.
Regarding the scope of program element of the PWR Vessel Internals Program.
The staff observed that Dominion Energy Engineering Technical Evaluation Document No.
ETE-SLR-2023-3310 indicates the PWR RVI design at VCSNS does not include baffle corner bolts and that this supports that applicants basis for implementing Primary Item W7 in Table 4-3 of the MRP-227, Rev. 1-A report only applies to Primary VT-3 inspections of the baffle plates.
Regarding the scope of program element of the PWR Vessel Internals Program. The staff observed that SLRA AMP B2.1.7 and basis document for the AMP in Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 establish that the control rod drive mechanism (CRDM) adapter housings, penetration nozzles, and sleeve configurations are not susceptible to the aging effect and mechanism combination of loss of material due to wear. The staff also observed that the design modification package in VCSNS ECR No. 50868 (as posted to the audit portal) provides a full summary of the RRVCH design modification that was performed in 2016 and provides supporting design and technical information that the modified configuration of the CRDM adapter housings, penetration nozzles, and thermal sleeve configurations in the RRVCH are not susceptible to levels of loss of material due to wear that could potentially impact the structural integrity or intended functions of the CRDM adapter housings, penetration nozzles, and thermal sleeves, or the RRVCH. Specifically, the staff observed that the relevant design record provides sufficient demonstration that the CRDM adapter housings, penetration nozzles, and thermal sleeves in the RRVCH are designed in a manner that restricts relative surface-to-surface motions to a degree that precludes the components from developing significant amounts of wear. The staff will address this as part of the staffs AMR further evaluation assessment for the specified component types in Section 3.1.2.2.10 of the SE.
Regarding the scope of program element of the PWR Vessel Internals Program. The staff observed that SLRA AMP B2.1.7 and basis document for the AMP in Dominion
17 Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 establish that the RVI fuel alignment pins in the upper and lower internals assembly are not made of malcomized Type 304 stainless steel and are instead made of non-malcomized Type 316 stainless steel that does not require aging management for the period of subsequent period of extended operation. The staff observed that the guidelines in MRP 2018-022 would only place the upper and lower fuel alignment pins in the Existing Program of the AMP and manage the pins for loss of material due to wear if the pins are made from malcomized stainless steel materials.
Regarding the operating experience element of the PWR Vessel Internals Program.
The staff observed that the basis document for SLRA AMP B2.1.7 in Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 provides a complete summary of all VCSNS Primary RVI component inspections that are specified for implementation under the NEI 03-08 Needed requirements of Section 7.3 of the MRP-227, Rev. 1-A report and the results of those Primary component inspections, as performed during the first period of extended operation for VCSNS. The staff audited the augmented inspection results of the VCSNS Primary component inspections performed for the first period of extended operation. The staff observed that the Primary inspections performed during the first period of extended operation did not yield any age-related flaw conditions or conditions of aging severe enough to invoke sample-expansion inspections of the defined Expansion category components for the program, or that might indicate the need to alter the I&E criteria of specified Primary, Expansion category, or Existing Program components from those defined for the component types in MRP-227, Rev. 1-A or in supplemental guidelines used by the applicant for development of the RVI gap analysis. Thus, the staff observed the current I&E criteria for specified Primary, Expansion category, or Existing Program components in the program are based on the industrys generic OE compiled by EPRI MRP for its development of MRP-227, Rev. 1-A report, or on the supplemental guideline criteria in MRP 2017-009, MRP 2018-002, MRP 2018-022, or WCAP-17451-P, Rev. 2 that were used for development of the applicants RVI gap analysis. However, the staff will document its OE review as part of the staffs operating experience program element review of the AMP that will be included in the SE.
During the audit, the staff made the following observations regarding the AMR line items provided for specified VCSNS RVI components in SLRA Table 3.1.2-2:
- For the AMR items in SLRA Table 3.1.2-2 (i.e., on SLRA page 3-87) applying to the clevis insert assembly bolts and clevis wear surfaces made from a Stellite material, the staff observed that the AMR items cross reference to (1) SLRA Table 3.1.1, Item 3.1.1-053c and GALL-SLR Item IV.B2.RP-399 for management of cracking in the clevis insert bolts, and (2) SLRA Table 3.1.1, Item 3.1.1-059c and GALL-SLR Item IV.B2.RP-285 for management loss of material due to wear and changes in dimension due to void swelling or distortion in the clevis insert bolts and Stellite clevis wear surfaces (as Existing Program components of the AMP). However, the staff observed that the gap analysis in Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310 establishes that the clevis insert bolts, dowels, and Stellite wear surfaces are Primary category components. The staff will review the acceptability of AMR items applying to the clevis insert assembly components in Section 3.1.2.2.9 of the SE. The staff may consider the issuance of an RAI for this AMR topic.
- For the AMR item in SLRA Table 3.1.2-2 (i.e., on SLRA page 3-87) applying to the upper core plate (UCP) inserts, the staff observed that the item cross references to the AMR items in SLRA Table 3.1.1, Item 3.1.1-059c and GALL-SLR Item IV.B2.RP-999 for
18 management of loss of material due to wear in the UCP inserts as an Existing Program component type for the AMP. However, the staff also observed that Note 1 for No Additional Measures components in SLRA Table 3.1.2-2 identifies the UCP insert bolts, locking devices, and dowel pins are No Additional Measures components for the AMP, and the table for Existing Program components in the gap analysis of the AMP basis document (i.e., Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310) does not include UCP inserts as Existing Program components. The staff will review the acceptability of the AMR item applying to the UCP inserts in Section 3.1.2.2.9 of the SE. The staff may consider the issuance of an RAI for this AMR topic.
- For the AMR item in SLRA Table 3.1.2-2 (i.e., on SLRA page 3-90) applying to the radial keys (i.e., wear surfaces in the keys), the staff observed that the item cross references to the AMR items in SLRA Table 3.1.1, Item 3.1.1-059c and GALL-SLR Item IV.B2.RP-999 for management of loss of material due to wear radial key wear surfaces, and loss of preload due to irradiation-enhanced stress relaxation or creep and changes in dimension due to void swelling or distortion in the radial keys (as Existing Program components for the AMP).
However, the staff also observed that Note 1 for No Additional Measures components in SLRA Table 3.1.2-2 identifies the that the radial keys are No Additional Measure components for the AMP, and that the Primary category table in the gap analysis of the AMP basis document (i.e., Dominion Energy Engineering Technical Evaluation Document No. ETE-SLR-2023-3310) identifies that the Stellite radial key wear surfaces are Primary category components. The staff will review the acceptability of the AMR item applying to the radial keys in Section 3.1.2.2.9 of the SE. The staff may consider the issuance of an RAI for this AMR topic.
- For the AMR item in SLRA Table 3.1.2-2 (i.e., on SLRA page 3-90) applying to the CB UGW, the staff observed the generic OE for the CB UGW potentially could impact the details of the AMR item for the weld type in SLRA Table 3.1.2-2. The staff will address this in both SE Section 3.1.2.2.9 and the SE section that documents the staffs evaluation of SLRA AMP B2.1.7.
The staff also audited the description of Updated Final Safety Analysis Report (UFSAR) summary description of the AMP (i.e., SLRA Appendix A, Section A1.7, PWR Vessel Internals) provided in the FSAR supplement of the SLRA. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the SLRAs PWR Vessel Internals Program. The staff will consider issuing an RAI to obtain the information necessary to verify the sufficiency of the FSAR supplement program description.
SLRA AMP B2.1.8, Flow-Accelerated Corrosion Summary of Information in the Application. The SLRA states that AMP B2.1.8, Flow-Accelerated Corrosion, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M17, Flow-Accelerated Corrosion.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3311 Aging Management Program Evaluation Report - Flow-Accelerated Corrosion Rev. 0
19 Document Title Revision/Date ER-AA-FAC-102 Flow-Accelerated Corrosion (FAC) Inspection and Evaluation Activities Rev. 2 ER-AA-FAC-105 Erosion Control Program Rev. 1 OP-AA-100 Conduct of Operations Rev. 46 OE2-CR-11-02856 Component Acceptable with Tmeas less than Tmin per CC N-597 5/21/2011 OE4-CR-13-02749 FAC Program Self-Assessment 6/27/2013 OE6-CR-16-02879 Increased FAC Wear of Expander due to Entrance Effect 6/07/2016 OE7-CR-09-05381 Cavitation Noises at ILV03235-FW 12/07/2009 OE8-CR-16-00262 Through Wall Leak due to Cavitation Erosion 1/15/2016 CR-13-02750 FAC Self-Assessment 6/27/2013 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Flow-Accelerated Corrosion Program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.9, Bolting Integrity Summary of Information in the Application. The SLRA states that AMP B.2.1.9, Bolting Integrity, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M18, Bolting Integrity. The staff will document its review of the enhancements in the SE.
Audit Activities. From December 4-8, 2023, the staff participated in an onsite audit at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. While onsite, the staff engaged with the applicant staff, conducted walkdowns, and reviewed documents provided by the applicant.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3312 Aging Management Program Evaluation Report - Bolting Integrity Rev. 0 SLR-0901 Aging management program Development Guide, SLR Project Guideline, Dominion Energy Rev. 4 CR-12-03695 Seal Encapsulation Devices on Class 1 and 2 Bolted Bonnet Check Valves.
08-29-2012 CR-14-02360 Thread Degradation on A Team Generator Inspection Port 04-29-2014 CR-16-05336 Boron Accumulated at Body to Bonnet Connection 10-24-2016 CR-17-001626 Boron Discovered during walkdown 04-08-2017 CR-18-04882 Minor thread damage on MS safety valve base stud 10-30-18 CM-AA-ETE-101 Engineering Technical Evaluation, Fleet Administrative Procedure, Dominion Energy Rev. 16
20 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also audited the description of the SLRA AMP provided in the UFSAR supplement in section A.1.9, Bolting Integrity. The staff verified this description is consistent with the description provided in the GALL-SLR.
SLRA AMP B2.1.10, Steam Generators Summary of Information in the Application. The SLRA states that AMP B.2.1.10, Steam Generators, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M19, Steam Generators.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3313 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Steam Generators Rev. 0 ER-AA-SGP-10 Steam Generator Program Description Rev. 2 ER-AA-SGP-101 Steam Generator Program Rev. 1 ER-AA-SGP-102 Steam Generator Degradation Assessment Rev. 1 ER-AA-SGP-103 Steam Generator Condition Monitoring and Operational Assessments Rev. 1 SGMP-100.007 Steam Generator Maintenance Procedure. Degradation and Integrity Assessments Rev. 2 STP-550.003 Steam Generator Tube Inspection Rev. 2 SAP-0158 Steam Generator Management Program Rev. 4 CR-21-02292 Condition Report Details 11/16/2022 CR-19-02552 Condition Report Details 07/16/2019 CR-18-04726 Condition Report Details 10/24/2018 CR-17-01364 Condition Report Details 03/24/2017 CP-307 Primary-to-Secondary Leakage Rate Determination Rev. 16 (3/21/16)
CP-613 Steam Generator Chemistry Control Rev. 23, Change A CP-613 Steam Generator Chemistry Control Rev. 23, Change F CY-AP-SEC-200 Secondary System Chemistry Rev. 10 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, detection of aging effects, monitoring and trending, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP. The staff also identified an exception associated with the parameters monitored or inspected, and acceptance criteria, program elements.
During the audit, the staff made the following observation:
The exception identified by the staff was the referencing of updated versions of EPRI Steam Generator Management Program guidelines that were not yet referenced in the GALL-SLR Report AMP. The applicant indicated it would revise the SLRA to include an appropriate exception.
21 The staff also audited the description of the SLRA Steam Generators Program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.11, Open-Cycle Cooling Water System Summary of Information in the Application. The SLRA states that AMP B2.1.11, Open-Cycle Cooling Water System, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M20, Open-Cycle Cooling Water System.
At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3314 VC Summer SLR - AMP Evaluation Report - Open Cycle Cooling Water System Rev. 1 ES-560.211 Service Water System Heat Exchanger Performance Rev. 13 STP-223.022A Service Water Pump Test Rev. 10 PTP-173.001 Backflushing Service Water Side of Component Cooling Heat Exchanger A(B)
Rev. 4 SAP-1255 Service Water System Reliability Optimization Program Rev. 2 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also audited the description of the SLRA Open-Cycle Cooling Water System provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report SLRA AMP B2.1.12, Closed Treated Water Systems Summary of Information in the Application. The SLRA states that AMP B.2.1.12, Closed Treated Water Systems, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M21A, Closed Treated Water Systems. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3315 VC Summer SLR - AMP Evaluation Report -
closed Treated Water Rev. 0 ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 13 ER-AA-NDE-VT-601 VT-1 Visual Examination Procedure Rev. 5
22 CP-632 Corrosion Control of Auxiliary Systems Rev. 6 Change C CY-AA-100 Conduct of Chemistry Rev. 15 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also audited the description of the SLRA AMP for the Closed Treated Water Systems provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.13, Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems Summary of Information in the Application. The SLRA states that AMP B.2.1.13, Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M23, Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems. The staff will document its review of the enhancements in the SE.
Audit Activities. From December 4-8, 2023, the staff participated in an onsite audit at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. While onsite, the staff engaged with the applicant staff, conducted walkdowns, and reviewed documents provided by the applicant.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3316 Aging Management Program Evaluation Report -
Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems Rev. 0 CR-16-01823 Metallic foreign material at rail of spent fuel pit bridge crane 04/14/2016 CR-17-03043 Metallic foreign material at rail of polar crane 05/26/2017 NUREG-0554 Single-Failure Proof Cranes for Nuclear Power Plants 05/1979 NUREG-0612 Control of Heavy Loads at Nuclear Power Plants 07/1980 ASME NOG-1 Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) 2015 ASME B30.2 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist) 2005 ASME B30.16 Overhead Hoists (Underhung) 2016 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff performed walkdowns of the Fuel Handling Building, Auxiliary Building, Turbine Building, Service Water Intake Structure, Maintenance Shop, Service Water Pond Structures
23 and Dams, Yard, etc. The applicant provided information on the identified issue on the Reactor Building (Polar) Crane bridge and this issue was resolved via the supplement at the later date.
The staff also audited the description of the SLRA AMP provided in the UFSAR supplement in section A.1.13, Inspection of Overhead Heavy Load and Light Load (Related to Refueling Handling Systems. The staff verified this description is consistent with the description provided in the GALL-SLR.
SLRA AMP B2.3.14, Compressed Air Monitoring Summary of Information in the Application. The SLRA states that AMP B.2.3.14, Compressed Air Monitoring, is an existing program that will be consistent with the program elements in GALL-SLR Report AMP XI.M24, Compressed Air Monitoring.
The staff will document its review of the enhancement(s) in the SE. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document including the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, and confirmation process.
Audit Activities. The table below lists the documents that were reviewed by the staff and were found relevant to the Compressed Air Monitoring Program. These documents were provided by the applicant on their ePortal. The staff will document its review of relevant OE in the SE.
Document Title Revision/Date Procedure: OAP -106.1 Operating Rounds OAP - 106.1 Attachment VII Rev. 19 Procedure: ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 13 ETE - Engineering Technical Evaluation-2023-3317 V.C Summer Subsequent License Renewal AMP Evaluation Report - Compressed Air Monitoring Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, acceptance criteria, and confirmation process.
program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
During the audit, the staff verified the applicants claim that the monitoring and trending, detection of aging effects and corrective actions program element(s) of the SLRA AMP will be consistent after implementation of the identified enhancements.
For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and OE by reviewing documentation contained in the SLRA and ePortal. The staff will evaluate any identified plant-specific OE in the SE.
The staff also audited the description of the SLRA AMP Compressed Air Monitoring program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
24 SLRA AMP B2.1.15, Fire Protection Summary of Information in the Application. The SLRA states that the Aging Management Program (AMP) B2.1.15, Fire Protection, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M26, Fire Protection.
At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the description of the enhancements in the SLRA, the applicants basis document, and annotated markups of procedures related to fire damper inspections.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title/Description Revision/Date ETE-SLR-2023-3318 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Fire Protection Rev. 0 STP-128.007 CO2 System Valve Lineup Verification Rev. 7 STP-428.062 Ventilation Fire Damper Inspection Units XAH-20, XAH-9B, XAH-11B, XFN-125 and XFN-124 Rev. 5, Change A STP-428.063 Ventilation Fire Damper Inspection Units XFN-147A & B, XAH-19A, XAA-12 and XAA-40B Rev. 5, Change A STP-428.061 Ventilation Fire Damper Inspection Units XAH-13A, XAH-13B, XAH-15B, XAH-8, XAH-21 and XAA-27 Rev. 5 STP-428.060 Ventilation Fire Damper Units XAH-24A, XAH-24B, XAH-31, XAA-29A, XAA-29B and XAH-17 Rev. 9, Change B STP-428.065 Ventilation Fire Damper Inspection Units XAH-11A, XAH-33, XAH-15A, XAH-7B, XAH-6, XFR-80B, XFN-80A, XAH-18, XFN-93, and XAH-48 Rev. 7 STP-428.064 Ventilation Fire Damper Inspection Units XAA-10B, XAA-12A, XAA-47, XAA-10A, XAA-48, XAA-14B, XAH-19B, XAH-7A, XAA-40A, XAA-41 and XAH-9A Rev. 5, Change D STP-128.019 Semi-annual Fire Door Inspection Rev. 8 STP-728.051 Building Fire Barrier and Penetration Seal Inspections Rev. 0 ER-AA-101 System Engineering Walkdowns Rev. 7 STP-128.024 CO2 System Functional Refueling test Rev. 10 OAP-106.1 Operations Administrative Procedure Rev. 19 SAP-131A Fire Protection Program Surveillances and Compensatory Measures Rev. 6 SLR-3318_XI Virgil C. Summer Nuclear Station Subsequent License Renewal - AMP Component Listing Fire Protection July 6, 2023 (based on AMR data)
M26_OE Detailed Evaluation VCS Operating Experience Detail Report June 28, 2023 (printed date)
CR-14-03946 Hole in gypsum wall July 16, 2014 CR-15-00850 Degraded penetration seal February 24, 2015 CR-16-01534 Degraded Kaowool fire barrier March 29, 2016 CR-16-06389 Fire door doesnt close and latch December 12, 2016 CR-18-02798 Holes in masonry wall July 10, 2018 CR-19-00004 Penetration seal separation January 2, 2019 CR-19-03884 Albi-Duraspray separating from column October 31, 2019
25 Document Title/Description Revision/Date TRP-2 Technical Requirements Package - Fire Protection Rev. 10 STP-728.042 Control Building Elevation 463 Fire Barrier Inspection Rev. 7, Change D N/A TRP 026 draft markups - Table 3.5.2-14 N/A DWG-000-023-018 Fire shutter and fire window Rev. 36 DWG-000-105-017 Elevation view of fire shutter Rev. 9 DWG-000-01-221 Fire shutter November 25, 1981 N/A Fire Window-Shutter Discussion & Ref Matl N/A During the audit, the staff verified the applicants claim that the preventive actions, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the scope of program and parameters monitored or inspected program elements, sufficient information was not available to determine whether they are consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the information necessary to verify these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Fire Protection program in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.16, Fire Water System Summary of Information in the Application. The SLRA states that AMP B2.1.16, Fire Water System, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M27, Fire Water System. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the description of the enhancements in the SLRA, the applicants basis document, and markups of ER-AA-AMP-101 and ES-900.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title/Description Revision/Date ETE-SLR-2023-3319 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Fire Water System Rev. 0 SAP-131A Fire Protection Program Surveillances and Compensatory Measures Rev. 6 ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 12 and Rev. 13 MMP-300.006 General Strainer Cleaning and Replacement Rev. 12, Change B
STP-128.012 NFPA 805 Required Fire Hydrant Inspection Rev. 8 STP-128.022 NFPA 805 Required Fire System Visual Inspection Rev. 8 ES-900 Fire Suppression System Corrosion Monitoring Program Rev. 1
26 Document Title/Description Revision/Date MMP-150.005 Diving Support for Circulating Water System Equipment Rev. 1, Change D
STP-228.001 Fire Protection System Fire Pumps Test Rev. 7 MMP-400.002 Maintenance of Circulation Water Traveling Screens Rev. 3, Change B
STP-728.402 Fire Yard Hydrant Annual Inspection and Hose Hydrostatic Testing Rev. 6, Change A
PTP-114.007 Wet Pipe Sprinkler Systems Water Flow Test Rev. 10 STP-128.021 Fire Service Water Flow Test Rev. 11 SAP-0156 License Renewal Management Program Rev. 2 PTP-114.040 Preaction Sprinkler Systems Water Flow Test Rev. 8 PTP-114.045 Sprinkler System Strainer Flush Rev. 2 NA VCS Operating Experience List Report June 19, 2023 (printed)
CR-11-05878 Small hole in fire service piping November 21, 2011 WO1117208-001 Piping replacement August 2, 2012 (close date)
CR-13-01407 Evaluation of NRC Information Notice 13-06 March 27, 2013 CR-13-04700 Evaluation of September 2013 Rev. of Loss of Control Standards November 11, 2013 WO2012465-011 UT Examination of Fire Service Piping May 25, 2021 WO88101096995 Pre-Action sprinkler piping inspection January 20, 2021 (start date)
NA Report on Sprinklers Sampled April 21, 2021 1703370 Sprinkler photos NA WO1703370-004 Sprinkler inspections January 21, 2021 (start date)
WO88101027969 Visual inspection of fire service pump bays February 8, 2022 (start date)
WO1409868-002 Inspection and cleaning of circulating water jockey pump bays, raw water pump bays and the fire service pump bays November 1, 2015 (start date)
N/A Strainer PM summary N/A WO88201709837 Heat exchanger tube bundle replacement January 16, 2024 (Completion Date)
CR1234402 Fire pump diesel heat exchanger leak July 31, 2023 STP-428.401 Diesel Driven Fire Service Pump Engine, XPP0134B-E Inspection Rev. 4, Change B
N/A SLRA 2.3.3.13 Diesel Engine Text N/A During the audit, the staff verified the applicants claim that the preventive actions, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP were consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the scope of program and parameters monitored or inspected program elements, sufficient information was not available to determine whether they
27 are consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the information necessary to verify these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Fire Water System program in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.17, Outdoor and Large Atmospheric Metallic Storage Tanks Summary of Information in the Application. The SLRA states that AMP B.2.1.17, Outdoor and Large Atmospheric Metallic Storage Tanks, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.M29, Outdoor and Large Atmospheric Metallic Storage Tanks.
At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the program elements described in the applicants basis document including the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions.
Audit Activities. The table below lists the documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR2023-3320 V.C. Summer Subsequent License Renewal Aging Management Program Evaluation Report -Outdoor and Large Atmospheric Metallic Storage Tanks Rev. 0 ER-AA-101 System Engineering Walkdowns Station, Procedures, VCSNS, Dominion Energy Rev. 17 During the audit, the staff verified applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Outdoor and Large Atmospheric Metallic Storage Tanks Program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.18, Fuel Oil Chemistry Summary of Information in the Application. The SLRA states that AMP B.2.1.18, Fuel Oil Chemistry, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M30, Fuel Oil Chemistry. The staff will document its review of the enhancements in the SE.
The SLRA states that AMP Fuel Oil Chemistry, is an existing plant-specific aging management program. The staff audited the SLRA AMP to determine consistency with SRP-SLR Section A.1.18, Aging Management Program Elements.
28 Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the Fuel Oil Chemistry Program. These documents were provided by V.C. Summer.
The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3321 V.C. Summer Subsequent License Renewal - Aging Management Program Evaluation Report -Fuel Oil Chemistry Rev. 0 Procedure STP-409, 002A, Diesel Generator Fuel Oil Storage Tank, XTK0053A Cleaning, VCSNS Rev. 2C Procedure STP-409, 002B, Diesel Generator Fuel Oil Storage Tank, XTK0053B Cleaning, VCSNS Rev. 2C P&ID Drawing IMS-55-011, Diesel Fuel System, V.C. Summer Nuclear Station Dominion Energy Rev. 4 P&ID Drawing ISM-32-080 Sheet A Fuel Oil Tank V.C. Summer Nuclear Station Dominion Energy Rev. 2 During the audit, the staff verified the applicants claim that the scope of program, "corrective action and confirmation process program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLRReport AMP. The staff also verified applicants claim that the preventive actions, parameters monitored or inspected, monitoring and trending, detection of aging effects and acceptance criteria, program elements of the SLRA AMP will be consistent after implementation of the identified enhancements with the corresponding elements of the GALL-SLR Report AMP.
For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and OE by reviewing documentation contained in the SLRA and ePortal.
The staff also audited the description of the SLRA AMP provided in the UFSAR supplement.
The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.19, Reactor Vessel Material Surveillance Summary of Information in the Application. The SLRA states that AMP XI.M31, Reactor Vessel Material Surveillance is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M31, Reactor Vessel Material Surveillance.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date VC-ETE-000-ETE-SLR-2023-3322-A V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Reactor Vessel Material Surveillance Rev. 0 ER-AA-RVI-101 Rev. 7 Reactor Vessel Integrity Program Rev. 7 rpt AMP Report m31 VCS Operating Experience Detail Report Printed 8/15/2023 ML23250A311 VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 -
RE: REQUEST TO REVISE REACTOR VESSEL MATERIAL SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES (EPID L-2023-LLL-0002)
September 20, 2023
29 Document Title Revision/Date ETE-SLR-2023-3333 V. C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Operating Experience for AMPs Rev. 0 tbl_OE Search Terms.pdf N/A N/A During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA AMP Reactor Vessel Material Surveillance FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.20, One-Time Inspection Summary of Information in the Application. The SLRA states that AMP B2.1.20, One-Time Inspection, is a new program that is consistent with the program elements in GALL-SLR Report AMP XI.M32, One-Time Inspection. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3323 VC Summer SLR Project - AMP Evaluation Report -
One-Time Inspection Rev. 0 ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 13 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA One-Time Inspection program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report Table XI-01.
SLRA AMP B2.1.21, Selective Leaching Summary of Information in the Application. The SLRA states that AMP B2.1.21, Selective Leaching, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.M33, Selective Leaching. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
30 Document Title Revision/Date ETE-SLR-2023-3324 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Selective Leaching Rev. 0 ETE-SLR-2023-3324 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Selective Leaching Rev. 1 CR-22-00114 V.C. Summer License Renewal Selective Leaching Study 01/12/2022 NESML-659 V.C. Summer License Renewal Selective Leaching Study 3 07/15/2022 NESML-660 V.C. Summer License Renewal Selective Leaching Study 2 08/09/2022 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
During the audit, the staff reviewed Rev. 1 to ETE-SLR-2023-3324 and noted a minimum of one additional destructive examination will be performed when destructive examinations do not meet acceptance criteria.
The staff also audited the description of the SLRA Selective Leaching program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.22, ASME Code Class 1 Small-Bore Piping Program Summary of Information in the Application. The SLRA states that AMP B2.1.22, ASME Code Class 1 Small-Bore Piping is an existing condition monitoring program that is consistent with the program elements in GALL-SLR Report AMP XI.M35, ASME Code Class 1 Small-Bore Piping.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date NCN-03-3436 Boron Deposit Detected Around Seal Injection Piping Weld 10/20/2003 RCA-04-0988 Root Cause Analysis Report VC Summer Nuclear Station, Weld Failure and Pressure Boundary Leakage on C Reactor Coolant Pump Seal Injection Nozzle 05/27/2004 TR-05600-023 Allowable Level of Vibration Measurement of 11/2 Piping Attached to the Reactor Coolant Pumps 03/16/2005 BWXT Final Report 1318-001-04-17 Analysis of A Leaking Reactor Coolant Pump (RCP)
Seal Injection Nozzle Flange to Thermal Barrier Weld Joint 05/2004 ECR-50547 Engineering Change Completion Summery Checklist Rev. 4 03/07/2006 LMT-PAUT-VCS_RF26-001 Phased Array Ultrasonic Examination Report License Renewal (PLEX) Socket Welded Joints 10/15/2021 Work Order 1813562 License Renewal Inspection Phased Array UT in accordance with ES-0163, Class 1 Small Bore Piping Socket Welds 11/27/2019
31 Document Title Revision/Date WDI-PJF-1332732-FSR-001 VC Summer Nuclear Power Station RF25 In-Service Inspection Final Report 05/02/2020 LMT-VCS_RF26-001 ISI Final Report VC Summer 1RF26 October 2021 10/28/2021 NCN-04-0879 Apparent RCS Boundary Leakage Identified at the Seal Injection Line to the C RCP 03/31/2004 During the audit, the staff verified applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA AMP ASME Code Class 1 Small-Bore Piping provided in the FSAR supplement. The staff verified that this description is consistent with the description provided in the SRP-SLR Report.
SLRA AMP B2.1.23, External Surfaces Monitoring of Mechanical Components Summary of Information in the Application. The SLRA states that AMP B2.1.23, External Surfaces Monitoring of Mechanical Components, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M36, External Surfaces Monitoring of Mechanical Components. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3326 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - External Surfaces Monitoring of Mechanical Components Rev. 0 ER - AA-103 System Engineering Handbook Rev. 5 ER-AA-101 System Engineering Walkdowns Rev. 7 CR-13-04775 Walkdown performed on 11-7-13, Chilled Water 11/14/2013 CR-19-04119 Walkdown performed on 11/15/2019 - Service Water System 11/15/2019 CR-19-03947 PLEX Visual Inspection - pitting on 2 piping 11/05/2019 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA External Surfaces Monitoring of Mechanical Components provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report Table XI-01.
SLRA AMP B2.1.24, Flux Thimble Tube Inspection Summary of Information in the Application. The SLRA states that AMP XI.M37, Flux Thimble Tube Inspection is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M37, Flux Thimble Tube Inspection.
32 Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date VC-ETE-000-ETE-SLR-2023-3327-A (ETE)
V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Thimble Tube Inspection Rev. 0 VC-PROCREC-000-ES-530-AE164987 (PRO)
Engineering Services Procedure ES-530 Incore Flux Thimble Wear Assessment Rev. 3 OE Detailed Evaluations XI.M37 VCS Operating Experience Detail Report Flux Thimble Tube Inspection 8/23/2023 Incore_ECT_Inspection Report VCS-RF26 VCS-RF26 October 2021 Final Report Incore Thimble Tubes 10/2021 Response to IEB 88-09 ML20206C685.pdf V.C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Response to NRC Bulletin 88-09 11/4/1988 M37 Proposed Enhancement V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Thimble Tube Inspection Draft NRC Bulletin No. 88-09 Thimble Tube Thinning in Westinghouse Reactors 7/26/1988 ETE-SLR-2023-3333 V. C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Operating Experience for AMPs Rev. 0 tbl_OE Search Terms.pdf N/A N/A During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
In addition, the staff found that for the detection of aging effects and monitoring and trending program elements, sufficient information was not available at the time of the audit to verify whether it was consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing an RAI to obtain the information necessary to verify whether this program element is consistent with the corresponding program elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations regarding SLRA AMP XI.M37, Flux Thimble Tube Inspection:
At the time of the audit, the Flux Thimble Tube Inspection program is governed by engineering services procedure, ES-530 Rev. 3. The procedure establishes a future inspection target outage by requiring a measurement of wear during the first outage after a new flux thimble tube in installed and then extrapolating when that wear rate will result in a specific percentage of through-wall wear in the newly installed flux thimble tube.
The licensee has replaced several of their flux thimble tubes with a chrome-plated tubes that have shown no appreciable wear since replacement. Since the chrome-plated tubes showed no appreciable wear in their first inspection after replacement, the engineering services procedure does not specify a target outage for future inspections. A minimum inspection periodicity is not otherwise specified by the procedure. The guidance in program element Detection of Aging Effects in GALL-SLR Report AMP XI.M37, Flux
33 Thimble Tube Inspection notes that use of more wear-resistant thimble tube materials may require that inspections are conducted at an adequate frequency for the new materials.
The applicants staff noted during the interview that current practice at VC Summer is to inspect every flux thimble tube each outage even though not all tubes require inspection each outage by procedure. The time required to set up and remove the inspection equipment is significantly greater than the time required to do inspections, so inspecting all thimble tubes is considered prudent even though it may not be required by the internal procedure or NRC Bulletin No. 88-09.
During the audit review period, the applicant provided M37 Proposed Enhancement onto the audit portal for the SLRA which documented that the applicant may be submit a supplement to the SLRA with an enhancement to AMP XI.M37, Flux Thimble Tube Inspection that addresses the detection of aging effects and monitoring and trending program elements to increase consistency with the corresponding elements of the GALL-SLR Report AMP. The staffs review of SRLA AMP XI.M37, Flux Thimble Tube Inspection in the SE will address any addition of enhancements to the AMP if they are submitted in an ensuing SLRA supplement.
The staff also audited the description of the SLRA AMP Flux Thimble Tube Inspection FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.25, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Summary of Information in the Application. The SLRA states that AMP B2.1.25, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3328 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Rev. 0 ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 13 ER-AA-NDE-VT-601 VT-1 Visual Examination Procedure Rev. 5 CR-14-03765 Leak discovered on the Service Water Piping Downstream of the A Train Water Jet Exhauster 07/07/2014 CR-18-04206 Leak discovered on the Service Water Piping Downstream of the B Train Water Jet Exhauster 10/08/2018 ETE-SLR-2023-3314 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Open-Cycle Cooling Water System Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive
34 actions, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the parameters monitored or inspected program element, sufficient information was not available to determine whether it was consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing an RAI to obtain the information necessary to verify whether this program element is consistent with the corresponding program element of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.26, Lube Oil Analysis Summary of Information in the Application. The SLRA states that AMP B.2.1.26, Lubricating Oil Analysis, is an existing program with enhancement that will be consistent with the program elements in GALL-SLR Report AMP XI.M39, Lubricating Oil Analysis.
Audit Activities. The table below lists the documents that were reviewed by the staff and were found relevant to the Lubricating Oil Analysis. These documents were provided by the applicants ePortal. The staff will document its review of relevant OE in the SE.
Document Title Revision/Date ETE-SLR-2023-3329 V.C. Summer Subsequent License Renewal - Aging Management Program Evaluation Report - Lubricating Oil Analysis Rev. 0 EMP-295.007 RCP Motor Oil Level and Alarm Check, VCSNS Rev. 5 EMP-295.015 Charging Pump Motor Maintenance, VCSNS Rev. 9 ES-527 Control and Maintenance of the Plant Lubrication Manual Database, Engineering Services, VCSNS Rev. 1 MMP-300.014 Safety injection Pump Hight Speed Gear Drive Maintenance Rev. 10 MMP-180.038 Diesel Generator A, XEG0001AE Periodic Maintenance Checks, VCSNS Rev. 7 MMP-300.007 Fire Pump Diesel, XPP0134B-E Preventive Maintenance Rev. 14 MMP-451.002 Maintenance of HVAC Mechanical Water Chillers Rev. 20 MMP-300.015 Turbine Maintenance, Emergency Feed Water Pump TPP0008 Rev. 19 CR-18-04947 A RCP and B RCP Upper Reservoir Oil Change 11/01/2018 CR-12-00578 HVAC Chiller Oil Sample Out of Specification 02/09/2012 CR-13-02646 Emergency Feed Water Pump B Out Boar Bearing 06/20/2013 CR-15-05862 C RCP Upper and Lower Reservoir Oil Change 11/11/2015 CP-400 Administrative Guidelines for the Oil Lab, Chemistry Administrative, VCSNS Rev. 24 During the audit, the staff verified applicants claim that the scope of program, preventive actions, parameters monitored or inspected, monitoring and trending, detection of aging effects, acceptance criteria and corrective actions program element of the SLRA AMP will be consistent after implementation of the identified enhancements with the corresponding elements of the GALL-SLR Report AMP.
35 The staff also audited the description of the SLRA AMP Lubricating Oil Analysis program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRAAMPB2.1.27, Monitoring of Neutron-Absorbing Materials Other than Boraflex Summary of Information in the Application. The SLRA states that AMP B2.1.27, Monitoring of Neutron-Absorbing Materials Other than Boraflex, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.M40, Monitoring of Neutron-Absorbing Materials Other Than Boraflex.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3353 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Monitoring of Neutron-Absorbing Materials Other Than Boraflex Rev. 0 REP-108.001 Surveillance Test Program for Neutron Absorbing Material Contained in the Spent Fuel Racks Rev. 8 HI-2115011 Holtec Proprietary. Summary Report of the Examination of VC Summer Nuclear Station Boral Coupon Surveillance Coupon CS111960-1-2 Rev. 0 N/A Proprietary. Virgil C. Summer Nuclear Station Unit 1 Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools.
11/03/2016 HI-2210811 Holtec Proprietary. V.C. Summer Nuclear Station Boral Coupon Surveillance Report 08192021 Rev. 0 NEI 16-03-A Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA AMP Monitoring of Neutron-Absorbing Materials Other Than Boraflex provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.28, Buried and Underground Piping and Tanks Summary of Information in the Application. The SLRA states that AMP B2.1.28, Buried and Underground Piping and Tanks, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.M41, Buried and Underground Piping and Tanks.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
36 Document Title Revision/Date ETE-SLR-2023-3330 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Buried and Underground Piping and Tanks Rev. 0 ER-AA-BPM-101 Underground Piping and Tank Integrity Program Rev. 19 SAP-1258 Underground Piping and Tank Integrity Program (UPTIP)
Rev. 1 EMP-500.003 Yard Piping Cathodic Protection Rev. 8A N/A Soil Corrosivity Analysis Report V.C. Summer Nuclear Station 12/2020 CMP-400.001 Excavation Backfill and Earthwork Rev. 11 SP-211-044461-000 Excavation and Placement of Structural Fill Rev. 10 STP-409.002A Diesel Generator Fuel Oil Storage Tank, XTK0053A Cleaning Rev. 2C STP-409.002B Diesel Generator Fuel Oil Storage Tank, XTK0053B Cleaning Rev. 2C CR-16-02603 Cathodic Protection Reading Below the 100mV Potential Change Criterion 05/23/2016 SP-337 Pipeline Specification for Conventional Piping Rev. 17 SP-545 Pipe Line Specification for Nuclear Safety Class Piping Rev. 14 SP-124-044461 Specification Yard Fire Protection System Virgil C. Summer Nuclear Station - Unit 1 Rev. 1 CR-11-01620 Water was observed puddling up outside of the South Turbine Building wall.
04/11/2011 CR-16-03645 During excavation, the 6-inch drain line from the Auxiliary Boiler House was uncovered and found with damage to the coal tar epoxy external coating.
07/19/2016 CR-17-01949 An underground potable water line break was discovered in the laydown yard of Warehouse F.
04/18/2017 CR-21-01475 Water is leaking from an underground source between the mechanic shop and the potable water building.
05/31/2021 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
During the audit, the staff reviewed ETE-SLR-2023-3330 and noted soil sample results indicating corrosivity greater than 10 points using the carbon steel column in Table 9-4, Soil Corrosivity Index from BPWORKS, of EPRI Report 3002005294, Soil Sampling and Testing Methods to Evaluate the Corrosivity of the Environment for Buried Piping and Tanks at Nuclear Power Plants, will require evaluation of potential scope expansion or category transition.
The staff also audited the description of the SLRA Buried and Underground Piping and Tanks program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.29, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Summary of Information in the Application. The SLRA states that AMP B2.1.29, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.M42, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks.
37 Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3331 Aging Management Program Evaluation Report --Internal Coatings/Linings for In-Scope Piping, Components, Heat Exchangers, and Tanks Rev. 0 ETE-SLR-2023-3319 Aging Management Program Evaluation Report -- Fire Water System Rev. 0 ETE-SLR-2023-3320 Aging Management Program Evaluation Report - Outdoor and Large Atmospheric Metallic Storage Tanks Rev. 0 ETE-SLR-2023-3121 Integrated Plant Assessment Report - Auxiliary Systems - Open Water Systems Rev. 1 ETE-SLR-2023-3119 Integrated Plant Assessment Report - Auxiliary Systems - Fire Protection System Rev. 0 CM-AA-CTG-104 Condition Assessment of Internally Coated/Lined Tanks, Components, and Pipes Subject to Immersion Service Rev. 3 VCSNS-EDBD (EF system)
Enhanced Design Basis Document, Emergency Feedwater System (EF)
Rev. 17 EDG DFOST XTK0053B Modified API-653 Internal Inspection and Integrity Assessment 03/10/2020 CR-17-01854 Walkdown Inspection of CST 04/14/2017 CR-17-02446 Inspection of component cooling water heat exchanger A 05/4/2017 N/A VCS Operating Experience Detail Report, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks 06/29/2023 ER-AA-AMP-101 Implementation of Activities Performed by License Renewal Aging Management Coordinators Rev. 13 PI-AA-200 Corrective Action Rev. 40 DOM-QA-1 Nuclear Facility Quality Assurance Program Description Rev. 35 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.30, ASME Section XI, Subsection IWE Summary of Information in the Application. The SLRA states that AMP B.2.1.30, ASME Section XI, Subsection IWE, is an existing program that, following enhancements, will be consistent with GALL-SLR Report -NUREG-2191,Section XI.S1, ASME Section XI, Subsection IWE, as modified by SLR-ISG-2021-03-STRUCTURES, Updated Aging Management Criteria for Structures Portions of Subsequent License Renewal Guidance.
Audit Activities. During its audit, the staff interviewed the applicants staff and reviewed documentation contained in the SLRA and provided by the applicant via the ePortal. For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and operating experience by reviewing documentation contained in the SLRA and ePortal.
38 Furthermore, the staff conducted additional OE searches virtually on the applicants CAP database using the following phrases: containment liner, moisture barrier, IWE examination, liner augmented examination, liner thickness loss, liner distress, liner discontinuities, liner irregularities.
The table below lists documents that were reviewed by the staff and were found to be relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3334 V. C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report - ASME Section XI, Subsection IWE Rev. 0 N/A Detail CR Evaluation (IWE)
VCS Operating Experience Detail Report Various Dates in Document N/A VCS OE Experience List Report Various date in Document OAR RF22 VCS Inservice Inspection (ISI) Owners Activity Report (OAR) For Refueling Outage 22 03/03/2016 OAR RF23 VCS Inservice Inspection OAR For Refueling Outage 23 08/29/2017 OAR RF24 VCS Inservice Inspection OAR For Refueling Outage 24 Signed 01/24/2019 OAR RF25 VCS Inservice Inspection OAR For Refueling Outage 25 07/30/2020 OAR RF26 CORR-000-SSERIAL No.22-009 VCS Inservice Inspection OAR For Refueling Outage 26 02/08/2022 CR-17-02066 OE1-CR1102834 Floor Leak Chase Exam 04/21/2017 CR-17-02023 OE1-CR1102834 Floor Leak Chase Exam Reactor Building Liner 04/20/2017 CR-11-02834 OE1-CR1102834 Floor Leak Chase Exam Reactor Building 05/20/2011 CR-14-02439 OE3-CR1402439 Crack in Coating Reactor Building (Note: Closed by WO 1405100-003) 05/14/2015 CR-14-02713 OE4-CR1402713 Waer in Hatch 05/13/2014 CR-16-02413 OE5-CR1602413 Moisture Barrier Inspection due to RIS 2016-07 09/22/2016 date of Item closed CR-20-01440 OE7-CR2001440 Damaged Moisture Barrier Reactor Building Liner 08/23/2020 SLR-3334-XI.S1 SLR -AMP Component Listing 07/06/2023 N/A Developmental References EPRI Report 1015337 Assembling Bolted Joint.
EPRI 3002008061 Bolted Joints Update IN 2007-07 (ML14070A114) Leak Chase Channel RIS 2016-07 Liner Moist Barrier (ML16068A436)
N/A During the audit, the staff verified the applicants claim that the scope of program, and acceptance criteria program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP In addition, the staff found that for the preventive actions, detection of aging effects, and corrective actions program elements, sufficient information was not available to determine whether they were consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
39 Discussions of observations made by the staff during the audit follow.
The staff reviewed the program basis document (ETE-SLR-2023-3334, Rev. 0) and the SLRA with respect to the program enhancements and noted that these needed additional clarification.
The SLR states that prior to the subsequent period of extended operation, enhancements will be implemented in the following program element(s): Parameters Monitored/Inspected (Element 3);
Detection of Aging Effects (Element 4); and Monitoring and Trending (Element 5); and all procedure(s) will be revised as described in SLR Section B2.1.30.
These issues are indicated in the breakout questions and the staff will consider issuing RAI(s) to address the issues.
The staffs review of the operating experience (OE) description of SLRA B2.1.30 noted that the factual description of plant-specific OE appeared not to adequately support the stated conclusion, but needing a supplemental description of the observed material condition of the containment pressure-retaining boundary components (within the scope of the AMP) based on past inspections. The staff will consider issuing an RAI to address this issue.
The staff also audited the description of the SLRA AMP ASME Section XI, Subsection IWE provided in the FSAR supplement. The staff verified that this description is consistent with the description provided in the SRP-SLR. However, the staff may consider issuing RAI(s) to verify the sufficiency of the FSAR supplement program description if conforming changes related to revised wording of enhancements offered by the applicant in a voluntary SLRA Supplement are not reflected in the corresponding SLR Commitments in SLRA Table A-4.0-1.
SLRA AMP B2.1.31, ASME Section XI, Subsection IWL Summary of Information in the Application. The SLRA states that AMP B.2.1.31, ASME Section XI, Subsection IWL, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.S2, ASME Section XI, Subsection IWL.
Audit Activities. An onsite audit was performed December 5-7, 2023, at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. The NRC staff reviewed the extent of leaching from the concrete at the tendon access gallery.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3335 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - ASME Section XI, Subsection IWL Rev. 0 ER-AA-CII-101 ASME Section XI Containment Inservice Inspection (Concrete/IWL) Program Fleet Implementation Requirements Rev. 5 ER-AA-NDE-VT-606 IWL Visual Examination Procedure Rev. 2 ST-07 Containment Inservice Inspection Evaluation Criteria Rev. 0 ISE-4 Containment Inservice Inspection Program Plan 3rd Inspection Interval, January 1, 2017, to December 21, 2026 Rev. 3F SP-228 Surveillance of Reactor Building Post Tension System Rev. 16 CR-11-03718 Condition Report 07/11/2011
40 Document Title Revision/Date CR-15-04979 Condition Report 10/12/2015 CR-15-05209 Condition Report 10/20/2015 CR-15-05387 Condition Report 10/26/2015 OE5-OAR-RF 25-IWL Engr Eval Rept Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Inservice Inspection (ISI) Owners Activity Report (OAR) for Refueling Outage 25 07/20/2020 STTS-1116306 Work Order 10/01/2012 STTS-1412607 Work Order 09/28/2015 STSS-1905137 Work Order 02/19/2021 CR-12-04283 Condition Report 10/04/2012 CR-09-03667 Condition Report 09/22/2009 CR-00-00988 Condition Report 08/08/2000 STTS 1116306 Work Order - VT-1C of Tendon Access Gallery 10/01/2012 STTS 1412607 Work Order - Gen Visual of Tendon Access Gallery 09/28/2015 STTS 1610235 Work Order - Gen Visual of Tendon Access Gallery 04/27/2017 STTS 1905137 Work Order - VT-1 of Tendon Access Gallery 02/19/2021 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also reviewed SLRA Further Evaluation (FE) 3.5.2.2.1.9 and noted that it is unclear how the licensee evaluated the concrete leaching around the tendon access gallery and determined that a plant-specific AMP was not warranted to monitor the leaching impacts to the intended functions of the structures.
The staff also audited the description of the SLRA ASME Section XI, Subsection IWL program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.32, ASME Section XI, Subsection IWF Summary of Information in the Application. The SLRA states that AMP B.2.1.32, ASME Section XI, Subsection IWF, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.S3, ASME Section XI, Subsection IWF.
Audit Activities. During its audit, the staff interviewed the applicants staff and reviewed documentation contained in the SLRA and provided by the applicant via the ePortal. For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and operating experience by reviewing documentation contained in the SLRA and ePortal.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3336 V. C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - ASME Section XI, Subsection IWF Rev. 0
41 Document Title Revision/Date ER-AA-ISI-100 Dominion Inservice Inspection Program - Nuclear Fleet Administrative Procedure Rev. 13 ISE-7 V. C. Summer Nuclear Station Fourth Interval ISI Program Plan Rev. 3D ER-AA-RRM-100 ASME Section XI Repair/Replacement Program Fleet Implementation Requirements.
Rev. 8 MMP-300.001 Bolted Connection Removal and Installation - Safety Related Rev. 17 Change J SAP-0643 ASME Code,Section XI, Repair/Replacement Program Rev. 8 Change D ES-0439 Engineering Services Procedure ES-0439, Inservice Inspection Program Controls, Rev. 1.
Rev. 1 Change C SAP-1294 Station Administration Procedure - SAP-124, Inservice Inspection Program - Safety Related Rev. 1 Change A MMP-305.004 Installation, Removal, Rework and Reinstallation of Pipe Supports Rev. 11 Change H N/A V. C. Summer Nuclear Station Unit 1, Inservice Inspection Owners Activity Report (OAR) for Refueling Outage 22 03/03/2016 N/A V. C. Summer Nuclear Station Unit 1, Inservice Inspection Owners Activity Report (OAR) for Refueling Outage 23 August 29, 2017 N/A V. C. Summer Nuclear Station Unit 1, Inservice Inspection Owners Activity Report (OAR) for Refueling Outage 24 Singed dated January 24, 2019 N/A V. C. Summer Nuclear Station Unit 1, Inservice Inspection Owners Activity Report (OAR) for Refueling Outage 25 July 30, 2020 N/A VCS OE Condition Report (CR) List Report-ASME Section XI, Subsection IWF Printed May 3, 2023 N/A VCS OE Detail Report, ASME Section XI, Subsection IWF
- Evaluation Various date each CR CR-12-05734 OE1-CR1205734-CRDM Missing Bolts - Appendix B August 5, 2016 CR-13-02013 VCS - NRC Integrated Inspection Report August 5, 2013 CR-14-03806 Three pipe supports were observed damaged on the Emergency Feedwater Piping July 8, 2014 CR-15-05177 ISI Inspection-the NRC inspector was not in agreement with our [licensee] position on the Reactor Vessel Supports N/A CR-18--4470 Corrosion and Coating degradation on the base plate of pipe support SWH-190 October 16, 2018 CR-18-04811 Inspection on MK-RHH-0025 on WO 1715890-001 two jam nuts were found loose October 27, 2018 CR-20-01375 ISI Work in support in 397AB support RHH-0133 was found out of alignment 12 deg and jam nut was loose and completely backed off to the end of the strut.
April 21, 2020 IN 2009-04 Age-Related Constant Support Degradation IRIS 311381 Main Steam System Snubbers identified with worn attachments March 9, 2014 CR-13-02432 RPV supports were not included in VCS ISI Program November 4, 2015 N/A Audit 21-07: Inservice Inspection, Inservice Testing and Appendix J Program October 7, 2021 SLR-3336 XI.S3 V. C. Summer Nuclear Station SLE-AMP Component Listing-ASME Section XI, Subsection IWF (Ref. ETE-SLR-2023-3336/ [XI.S3], Subsection IWF N/A During the audit, the staff verified the applicants claim that the preventive actions, and corrective actions program elements of the LRA/SLRA AMP are consistent with the
42 corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the scope of program, parameters monitored or inspected, detection of aging effects, monitoring and trending, and acceptance criteria program elements or will be consistent after implementation of the identified enhancements. The staff will consider issuing RAIs to obtain if additional information is necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP XI.S3. In addition, the staff will evaluate the exception taken to parameters monitored or inspected, program element in the SE.
During the audit, the staff made the following observations:
The staff reviewed SLRA B.2.1.32 AMP and SLRA Table A-4.0-1, Commitment 32, and noted that procedures will be revised to support the enhanced AMP (1) to include class MC components supports; (2) to evaluate the acceptability of inaccessible areas; (3) to protect ASTM A325 and ASTM A490 bolts and associated nuts and washers from corrosion during storage; (4) to specify a one-time inspection within five years prior to entering the subsequent period of extended operation of an additional 5% of the sample populations for Class 1, 2, and 3 piping supports most susceptible to age-related degradation; (5) to modify the inspection sample to include another support representative of the remaining population of supports when it is repaired to as new condition even though the original support met the ASME Code Subsection IWF acceptance standards; (6) to include additional unacceptable conditions not specified in IWF-3410 but included in GALL-SLR AMP XI.S3 Acceptance Criteria program element.
The staff will consider issuing RAIs and/or use a voluntary SLRA supplement offered by the applicant to obtain the necessary information to address issues identified during the audit.
The staff will document its review of these issues in the SE.
The staff also audited the description of the SLRA AMP ASME Section XI, Subsection IWF provided in the FSAR supplement. The staff reviewed the FSAR supplement description in SLRA Section A.1.32 and noted that it includes language such as that the AMP recommends additional one-time inspections and that volumetric examination should be performed.
However, the staff notes that an adequate FSAR supplement summary description is expected to describe what the program will do and not phrased as optional actions of what it recommends should be done. Accordingly, the staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate of a description of the SLRA AMP ASME Section XI, Subsection IWF. The staff will consider issuing an RAI to obtain the information necessary to verify the sufficiency of the FSAR supplement program description.
SLRA AMP B2.1.33, 10 CFR Part 50, Appendix J Summary of Information in the Application. The SLRA states that AMP B2.1.33, 10 CFR Part 50, Appendix J, is an existing program that is consistent with the program elements in GALL-SLR Report AMP XI.S4, 10 CFR Part 50, Appendix J.
Audit Activities. An onsite audit was performed December 5-7, 2023, at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE.
43 The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3337 Aging Management Program Evaluation Report - 10 CFR Part 50, Appendix J Rev. 0 ER-AA-CII-101 ASME Section XI Containment Inservice Inspection (Concrete/IWL) Program Fleet Implementation Requirements Rev. 5 ISE-4 Containment Inservice Inspection Program Plan 3rd Inspection Interval, January 1, 2017, to December 21, 2026 Rev. 3F CR-11-01932 Condition Report 04/21/2011 WO 1007908 Work Order 05/08/2011 CR-11-02700 Condition Report 05/16/2011 WO 1007866 Work Order 05/16/2011 CR-12-05037 Condition Report 10/31/2012 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program element(s) of the SLRA AMP is/are consistent with the corresponding element(s) of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA 10 CFR Part 50, Appendix J Program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.34, Masonry Walls Summary of Information in the Application. The SLRA states that AMP B2.1.34, Masonry Walls, is an existing program with enhancements will be consistent with the program elements in GALL-SLR Report AMP XI.S5, Masonry Walls. The staff will document its review of the enhancements in the SE.
Audit Activities. An onsite audit was performed December 5-7, 2023, at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. During the onsite audit, the NRC staff visited the Unit 1 Relay House, Auxiliary Building, and the Water Treatment Building where mortar cracking was identified in the OE documents.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3338 Aging Management Program Evaluation Report - Masonry Walls Rev. 0 TR00010-008 Maintenance Rule Inspections -2010, Assessment of In-Service Conditions of Important to Maintenance Rule Structures Rev. 0 TR-00010-009 Maintenance Rule Inspections -2015, Assessment of In-Service Conditions of Important to Maintenance Rule Structures Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive
44 actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program element(s) of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also audited the description of the SLRA Masonry Walls program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
During the walkdown of the Unit 1 Relay House, Auxiliary Building, and the Water Treatment Building, the staff noted that the conditions of the mortar cracking were consistent with the descriptions in the OE documents provided on the ePortal.
SLRA AMP B2.1.35, Structures Monitoring Summary of Information in the Application. The SLRA states that AMP B.2.1.35, Structures Monitoring, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.S6, Structures Monitoring. The staff will document its review of the enhancements in the SE.
Audit Activities. From December 4 to December 7, 2023, the staff participated in an onsite audit at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. While onsite, the staff engaged with the applicant staff, conducted walkdowns, and reviewed documents provided by the applicant.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3339 Aging Management Program Evaluation Report - Structures Monitoring Rev. 0 TR00010-008 V.C. Summer Maintenance Rule (Structural) Inspections -
2010- 2011) 06-06-2013 TR00010-009 V.C. Summer Maintenance Rule (Structural) Inspections - 2015 11-30-2016 TR00010-010 V.C. Summer Maintenance Rule (Structural) Inspections - 2020 12-02-2020 CR1502457 Surface corrosion of anchor bolts 06-02-2015 CR1105924 VC Summer evaluation of IN2011-20, Concrete degradation by ASR 11-28-2011 CR-19-001170 Corroded section of Auxiliary Building roof 03-27-2019 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
During the onsite audit, the staff performed walkdowns of the Fuel Handling Building, Auxiliary Building, Turbine Building, Service Water Intake Structure, Containment Sump, Maintenance Shop, Service water pond structures and dams, yard, roofs, transformer foundation and supports, anchor bolting, and pipe supports including snubbers, pumps and valves, and tanks, etc.
45 The staff also audited the description of the SLRA AMP provided in the UFSAR supplement in section A.1.35, Structures Monitoring. The staff verified this description is consistent with the description provided in the GALL-SLR.
SLRA AMP B2.1.36, Inspection of Water-Control Structures Associated with Nuclear Power Plants Summary of Information in the Application. The SLRA states that AMP B.2.1.36, Inspection of Water-Control Structures Associated with Nuclear Power Plants, is an existing program with enhancements that will be consistent with the program elements in GALL-SLR Report AMP XI.S7, Inspection of Water-Control Structures Associated with Nuclear Power Plants. The staff will document its review of the enhancements in the SE.
Audit Activities. From December 4 to December 7, 2023, the staff participated in an onsite audit at VCSNS in Jenkinsville, South Carolina, to gain a general overview of current conditions of the structures compared to the provided OE. While onsite, the staff engaged with the applicant staff, conducted walkdowns, and reviewed documents provided by the applicant.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3340 Aging Management Program Evaluation Report -
Inspection of Water-Control Structures Associated with Nuclear Power Plants Rev. 0 ES-400 Service Water Pond Structure and Dam Inspections, VCSNS, Dominion Energy CR1105924 VC Summer evaluation of IN2011-20, Concrete degradation by ASR 11-28-2011 CR1301688 Soil Erosion in the natural area 04-15-2013 CR1502457 Surface corrosion of anchor bolts 06-02-2015 CR1903062 Service water intake structure crack monitoring 08-28-2019 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
During the onsite audit, the staff performed walkdowns of the Service Water Intake Structure, Service Water Pond Structures and Dams, etc.
The staff also audited the description of the SLRA AMP provided in the UFSAR supplement in section A.1.36, Inspection of Water-Control Structures Associated with Nuclear Power Plants.
The staff verified this description is consistent with the description provided in the GALL-SLR.
SLRA AMP B2.1.37, Protective Coating Monitoring and Maintenance Summary of Information in the Application. The SLRA states that AMP B2.1.37, Protective Coating Monitoring and Maintenance, is an existing program that is consistent with NUREG-2191,Section XI.S8, Protective Coating Monitoring and Maintenance as modified by SLR-ISG-
46 Structures-2021-03-STRUCTURES, Updated Aging Management Criteria for Structures Portions of the Subsequent License Renewal Guidance.
Audit Activities. The table below lists the documents that were reviewed by the staff and were found relevant to the Protective Coating Monitoring and Maintenance Program. These documents were provided by the applicant and identified in the staffs search of the applicants OE database. The staff will document its review of relevant OE in the SE.
Document Title Revision/Date ETE-SLR-2023-3341 Aging Management Program Evaluation Report -
Protective Coating Monitoring and Maintenance Rev. 0 QSP-505 Quality System Procedure QSP-505 Visual Examination Rev. 12 CR-12-04684 Coatings Walkdown for outside the bio-wall 10/20/2012 CR-12-04773 Coatings walkdown for the A loop in the Reactor Building 10/23/2012 CR-14-02623 QC identified coating defects during a RB walkdown 05/08/2014 CR-18-04549 Walkdown of the emergency feedwater piping inside the Reactor Building 10/18/2018 CR-20-01676 Track Submittal of RF-25 Inservice Examination Reports 05/05/2020 S08_OE Detailed Evaluation VCS Operating Experience Detail Report Protective Coating Monitoring and Maintenance 06/28/2023 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP. During the audit of the operating experience program element, the staff reviewed a search of results of the plant-specific database to identify any previously unknown or recurring aging effects. The staff will evaluate any identified plant-specific OE in the SE.
The staff also audited the description of the SLRA AMP Protective Coating Monitoring and Maintenance program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.38, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Summary of Information in the Application. The SLRA notes that AMPB2.1.38, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is an existing program with enhancements that will be consistent with the program elements in GALL-SLRReport AMPXI.E1, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the applicants basis document, available procedures, and referenced documents.
Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date
47 ETE-SLR-2023-3343 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR50.49 Environmental Qualification Requirements Rev. 0 ETE-SLR-2023-3343 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR50.49 Environmental Qualification Requirements Rev. 1 ER-EA-CBL-103 Evaluation and Testing of Medium Voltage Cables Rev. 5 ER-AA-CBL-10 Cable Management Program Rev. 2 EA-AA-CBL-101 Evaluation and Testing of Low Voltage Cables Rev. 4 EA-CBL-103 Evaluation and Testing of Medium Voltage Cables (Markup)
Rev. 5 EA-AA-CBL-101 Evaluation and Testing of Low Voltage Cables (Markup)
Rev. 4 NA VCS Operating Experience Detail Report 07/26/2023 CR-17-02439 7.2KV Switchgear, Bus 1DB 05/04/2017 CR-12-05273 IER Level 3 12-9 Automatic Reactor Scram on Low Reactor Coolant System Flow 11/08/2012 CR-13-01190 Results of SA13-PE-01S, Cable Aging Program Snapshot Self-Assessment 03/08/2013 CR-11-01547 Faceplate for replacement transmitter 04/07/2011 CR-14-04629 High Temperature Alarm for IEB 08/22/2014 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff also audited the description of the SLRA Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff conducted onsite walkdowns of the cable raceways and cables and connections associated with onsite AC power systems.
SLRA AMP B2.1.39, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Summary of Information in the Application. The SLRA notes that AMPB2.1.39, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits, is an existing program that is consistent with the program elements in GALL-SLRReport AMPXI.E2, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.
Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
48 Document Title Revision/Date ETE-SLR-2023-3344 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Rev. 0 ICP-360.010 Aux Bldg Ventilation Radiation Monitor RMA0011 Calibration Rev. 6 STP-360.047 Reactor Building Purge Exhaust System Atmospheric Radiation Monitor RM-A14 Calibration Rev. 8A STP-360.035 Main Plant Exhaust Atmospheric Radiation Monitor RMA0003 Calibration Rev. 11C ES-0167 Non-EQ Instrumentation Cables and Connections Aging Management Program for License Renewal Rev. 1 ICP-310.020 Electrical Pretesting of Source Range N-33 Detector Rev. 1 ICP-500.002 NI Source and Intermediate Range Alignment for Troubleshooting/Repair Rev. 0 NA VCS Operating Experience Detail Report 07/26/2023 ICP-310.012 Electrical Pretesting of Power Range Detectors Rev. 5A CR-21-03004 Detector Insulation Resistance Testing under W/O2011458-001 and ICP-310.012 10/23/2021 CR-21-03060 Electrical Insulation resistance testing of Power Range Detectors (N44) 10/25/2021 NU-02SR775187 Nuclear Instrumentation Cable Testing in Support of Plant Life Extension (PLEX) at VC Summer Nuclear Power Plant 05/2020 NU-02SR771118 Final Report of AMS Tests: Testing of the NIS Power Range Circuits at V.C. Summer 11/2018 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49, Environmental Qualification Requirements Used in Instrumentation Circuits, provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff conducted onsite walkdowns of the cable raceways and cables and connections associated with onsite AC power systems.
SLRA AMP B2.1.40, Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Summary of Information in the Application. The SLRA notes that AMPB2.1.40, Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is an existing program with enhancements that will be consistent with the program elements in GALL-SLRReport AMPXI.E3A, Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, as modified by SLR-ISG-2021-04-ELECTRICAL, Updated Aging Management Criteria for Electrical Portions of the Subsequent License Renewal Guidance. At the time of the audit, the applicant had not yet fully developed the
49 documents necessary to implement this enhanced program, and the staffs audit addressed only the applicants basis document, available procedures, and referenced documents.
Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3345 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 0 ETE-SLR-2023-3345 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 2 CMP-700.013 Inspection of Electrical Manholes (Markup)
Rev. 3A CMP-700.013 Inspection of Electrical Manholes Rev. 3A ER-AA-CBL-103 Evaluation and Testing of Medium Voltage Cables Rev. 5 ER-AA-CBL-10 Cable Management Program Rev. 2 SAP-0156 License Renewal Management Program Rev. 2 SAP-1282 Electrical Cable Reliability Monitoring Program (Markup)
Rev. 0 SAP-1282 Electrical Cable Reliability Monitoring Program Rev. 0 EMP-405.029 Tan Delta Testing for Power Conductors Rev. 2 ER-AA-CBL-103 Evaluation and Testing of Medium Voltage Cables (Markup)
Rev. 5 RB42903 Engineering Evaluation of Test Results of Inaccessible Medium-Voltage Cables Not Subject to 10 CFR50.49 Environmental Qualification Requirements 02/11/2021 OAP-109.1 Guidelines for Severe Weather Rev. 6 OAP-109.1 Guidelines for Severe Weather (Markup)
Rev. 6 CR-19-02388 Water in the Electrical Manhole 01 07/03/2019 NA VCS Operating Experience Detail Report 07/26/2023 CR-11-01257 Water in the Electrical Manhole 01 03/18/2011 CR-13-01917 Water in the Electrical Manhole 08 04/30/2013 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP as modified by SLR-ISG-2021-04-ELECTRICAL, or will be consistent after implementation of the identified enhancements.
During the audit, the staff made the following observations:
The staff identified that ERA-AA-CBL-103, Evaluation and Testing of MV Cables, the fleet administrative procedure for the medium-voltage power cables in-scope of SLR only referenced North Anna Power Station. In response to a breakout question on the topic, the applicant agreed to create an enhancement in a future supplement to the AMP to include VC Summer in this procedure.
The staff identified a discrepancy between ER-AA-CBL-103, Evaluation and Testing of MV Cables, Rev. 5, the GALL-SLR Report, and the applicants fleet procedure ER-AA-CBL-10, Cable Management Program, Rev. 2, with the term long term wetting.
50 Specifically, the applicant defined the term long term wetting under the definition of the term wetted cables in Section 4.3.14, Wetted Cables, of ER-AA-CBL-10, Rev. 2, as a condition in which the cable sits in or is covered by water for a continuous period of months or longer (submerged)'. Whereas ER-AA-CBL-103, Rev. 5, and the GALL-SLR Report defines the term significant moisture as exposure to moisture that lasts more than 3 days (i.e., long term wetting or submergence over a continuous period) that if left unmanaged, could potentially lead to a loss of intended function. As indicated in the GALL-SLR Report, the terms long term wetting and submerged are equivalent to significant moisture. While the applicant indicated that ER-AA-CBL-10, Rev. 2, is a higher tier document that covers the entire fleet of cables (i.e., not just those within the scope of SLR) and will not be used as an implementing procedure for implementing GALL-SLR Report AMPs XI.E3A, B, and C for the subsequent period of extended operation, the staff informed the applicant that the existence of this alternative definition and criteria could create confusion and unintentionally allow cables within the scope of SLR to be subject to significant moisture for a continuous period of months or longer before the condition is corrected and evaluated versus the greater than the three-day threshold established in the GALL-SLR Report. The staff and applicant discussed this item during the audit in a breakout session. To minimize the probability of confusion between the necessary implementing procedures for AMP XI.E3A and the fleet cable management program, the applicant removed references to ERA-AA-CBL-10, Rev. 2, from the AMP XI.E3A basis document (i.e., ETE-SLR-2023-3345, V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, Rev. 2).
The staff also audited the description of the SLRA Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff conducted onsite walkdowns of a sample of electrical manholes around the site and examined cables and connections within the electrical manholes. The staff also examined pictures of the interior of several electrical manholes that were in the scope of SLR.
SLRA AMP B2.1.41, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Summary of Information in the Application. The SLRA notes that AMPB2.1.41, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is a new program that will be consistent with the program elements in GALL-SLRReport AMPXI.E3B, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, as modified by SLR-ISG-2021-04-ELECTRICAL, Updated Aging Management Criteria for Electrical Portions of the Subsequent License Renewal Guidance. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the applicants basis document, available procedures, and referenced documents.
Audit Activities. An onsite audit was performed December 5-7, 2023.
51 The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3350 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Instrumentation and Control Cables not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 0 ER-AA-CBL-101 Evaluation and Testing of Low Voltage Cables Rev. 4 ER-AA-CBL-101 Evaluation and Testing of Low Voltage Cables (Markup)
Rev. 4 NA VCS Operating Experience Detail Report 06/22/2023 CMP-700.013 Inspection of Electrical Manholes (Markup)
Rev. 3 CMP-700.013 Inspection of Electrical Manholes Rev. 3 CR-13-02250 Water found and removed from Pull Box PB-SG01 05/23/2013 CR-17-02277 Electrical Pretesting of Power Range Detectors (N42) 04/28/2017 CR-20-03414 Electrical Penetration exposed on end of cable 10/11/2020 CR-21-01891 Water in EMH0070 07/15/2021 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also audited the description of the SLRA Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff conducted onsite walkdowns of a sample of electrical manholes around the site and examined instrumentation cables and connections within the electrical manholes. The staff also examined pictures of the interior of several electrical manholes that were in the scope of SLR.
SLRA AMP B2.1.42, Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR50.49 Environmental Qualification Requirements Summary of Information in the Application. The SLRA notes that AMP B2.1.42, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.E3C, Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, as modified by SLR-ISG-2021-04-ELECTRICAL, Updated Aging Management Criteria for Electrical Portions of the Subsequent License Renewal Guidance. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit only addressed only the applicants basis document, available procedures, and referenced documents.
Audit Activities. An onsite audit was performed December 5-7, 2023.
52 The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3351 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 0 ETE-SLR-2023-3351 V.C. Summer Subsequent License Renewal Project-Aging Management Program Evaluation Report-Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 1 CMP-700.013 Inspection of Electrical Manholes Rev. 3 CMP-700.013 Inspection of Electrical Manholes (Markup)
Rev. 3 ER-AA-CBL-101 Evaluation and Testing of Low-Voltage Cables Rev. 4 ER-AA-CBL-101 Evaluation and Testing of Low-Voltage Cables (Markup)
Rev. 4 NA VCS Operating Experience Detail Report 06/22/2023 CR-13-03491 Circulating Water Traveling Screen E Motor XRS0005E-M C-W 08/20/2013 ER-AA-CBL-10 Cable Management Program Rev. 2 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP as modified by SLR-ISG-2021-04-ELECTRICAL.
During the audit, the staff identified a procedure conflict between ERA-AA-CBL-103, Evaluation and Testing of MV Cables and the Dominion fleet procedure ERA-AA-CBL-10, Cable Management Program, Rev. 2. with the term long term wetting definition as it relates to submergence condition. The ERA-AA-CBL-10 Rev. 2 long term wetting definition and criteria could potentially allow cables within the scope of subsequent license renewal to be subject to significant moisture for a continuous period of months or longer versus the greater than the three-day threshold established in GALL-SLR. The staffs position is that the terms long term wetting and submerged are equivalent to significant moisture and should be aligned in all the implementing procedures to ensure an effective implementation of the applicable programs.
In response to the audit question on the topic, the applicant agreed to create a condition report to remove ERA-AA-CBL-10 Rev. 2 as an implementing procedure for implementation of GALL-SLR XI.E3C AMP.
The staff also audited the description of the SLRA Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff conducted onsite walkdowns of a sample of electrical manholes around the site and examined low-voltage power cables and connections within the electrical manholes. The staff also examined pictures of the interior of several electrical manholes that were in the scope of SLR.
53 SLRA AMP B2.1.43, Fuse Holders Summary of Information in the Application. The SLRA notes that AMP B2.1.43, Fuse Holders, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.E5, Fuse Holders. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed the applicants basis document and referenced documents.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3347 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Fuse Holders Rev. 0 ETE-SLR-2023-3347 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Fuse Holders Rev. 1 XI.E5 AMP OE Report VCS Operating Experience Detail Report - Fuse Holders 07/26/2023 XI.E5 AMP OE Report VCS Operating Experience List Report - Fuse Holders 09/10/2023 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations. During its review of the Fuse Holders AMP basis document (i.e., ETE-SLR-2023-3347, Rev. 0), the staff identified that (1) the Detection of Aging Effects program element did not include the types of testing that may be used to provide an indication of the condition of the metallic clamp of the fuse holders and the time for the first visual inspections and tests for SLR and (2) the Acceptance Criteria program element did not provide the definition of an unacceptable indication and the acceptance criteria for a resistance measurement test. The staff and applicant discussed this item during the audit in a breakout session. As a result, the applicant revised the Detection of Aging Effects and Acceptance Criteria program elements in ETE-SLR-2023-3347, Rev. 1, to address the identified discrepancies. The staff reviewed the revised basis document and confirmed that it is consistent with GALL-SLR AMP XI.E5.
The staff also audited the description of the SLRA Fuse Holders provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B2.1.44, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Summary of Information in the Application. The SLRA notes that AMP B2.1.44, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.E6, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the applicants basis documents and referenced documents.
54 Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3348 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 0 ETE-SLR-2023-3106 V.C. Summer Subsequent License Renewal Project -
Integrated Plant Assessment Report - Electrical Components Rev. 0 CR-12-02338 Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 06/07/2012 SLR-3348-XI.E6 Subsequent License Renewal -AMP Component Listing -
Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 07/06/2023 NA VCS Operating Experience Detail Report - Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 08/17/2023 ER-AA-CBL-103 Evaluation and Testing of Medium Voltage Cables Rev. 5 ER-AA-CBL-103 Evaluation and Testing of Medium Voltage Cables (Markup)
Rev. 5 ETE-SLR-2023-3307 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Boric Acid Corrosion Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements provided in the FSAR supplement.
The staff verified this description is consistent with the description provided in the GALL-SLR Report.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff discussed OE related to electrical cable connections that are in the scope of SLR.
SLRA AMP B2.1.45, High-Voltage Insulators Summary of Information in the Application. The SLRA notes that AMP B2.1.45, High-Voltage Insulators, is a new program that will be consistent with the program elements in GALL-SLR Report AMP XI.E7, High-Voltage Insulators, as modified by SLR-ISG-2021 ELECTRICAL, Updated Aging Management Criteria for Electrical Portions of the Subsequent License Renewal Guidance. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this new program, and the staffs audit addressed only the applicants basis documents and referenced documents.
Audit Activities. An onsite audit was performed December 5-7, 2023.
55 The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3349 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - High-Voltage Insulators Rev. 0 NA VCS Operating Experience Detail Report - High Voltage Insulators 07/26/2023 CR-11-03190 NRC inspector noted several areas of concern while walking down the Switchyard.
06/06/2011 SLR-3349 XI.E7 Subsequent License Renewal - AMP Component Listing - High-Voltage Insulators 07/06/2023 ETE-SLR-2023-3106 V.C. Summer Subsequent License Renewal Project - Integrated Plant Assessment Report - Electrical Components Rev. 0 ETE-SLR-2023-3005 V.C. Summer Subsequent License Renewal Project - Regulated Event Report - 10 CFR 54.4(a)(3) Loss of All Alternating Current Power Rev. 0 ETE-SLR-2023-3606 V.C. Summer Subsequent License Renewal Project - Further Evaluation Recommended Report - Electrical and Instrumentation and Controls Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also audited the description of the SLRA High-Voltage Insulators provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report as modified by SLR-ISG-2021-04-ELECTRICAL.
The staff also conducted an onsite audit during which the staff interviewed the applicants staff and performed walkdowns of the facility. Specifically, the staff discussed OE related to electrical cable connections and examined high-voltage insulators that are in the scope of SLR.
SLRA AMP B3.1, Fatigue Monitoring Summary of Information in the Application. The SLRA states that AMP B3.1, Fatigue Monitoring, is an existing preventive program with enhancements that is consistent with the program elements in GALL-SLR AMP X.M1, Fatigue Monitoring. The staff will document its review of the enhancements in the SE.
At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3333 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Operating Experience for AMPs Rev. 0, 07/26/2023
56 During the audit, the staff audited the applicants claim that the scope of program, preventive actions, detection of aging effects, parameters monitored or inspected, monitoring and trending and acceptance criteria, program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
The staff found that, for the corrective actions program element, sufficient information was not available to verify whether it was consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the information necessary to verify whether this program element is consistent with the corresponding program element of the GALL-SLR Report AMP.
During the audit, the staff made the following observation. The corrective actions program element of GALL-SLR AMP X.M1 states that a flaw tolerance analysis with appropriate (e.g., inclusion of environmental effects) crack growth rate curves and associated inspections performed in accordance with Appendix L of ASME Code Section XI, is an acceptable correction action. ASME Code Section XI, Appendix L, L-3420, Successive Examinations specifies that, if the allowable operating period per the flaw tolerance evaluation is equal to or greater than 10 years, the locations subject to the flaw tolerance analysis be inspected at the end of each inservice inspection interval (i.e., 10-year inspection interval). In comparison, SLRA Section 4.3.4. indicates that the normal and alternate charging cold leg nozzles and pressurizer surge line hot leg nozzle have an 80-year projected environmentally adjusted cumulative usage N/A VCS Operating Experience Detail Report: Fatigue Monitoring 08/16/2023 N/A Operating Experience Search Terms 08/16/2023 CR-13-02110 Unanticipated RCS Cold Leg SI Piping Transients 05/14/2013 CR-21-02543 WESTEM Year-end Report 09/29/2023 CR-19-00729 SG FW Nozzle Thermal Cycling Anomaly 02/25/2019 ETE-SLR-2023-3301 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Fatigue Monitoring Program Rev. 0 08/3/2023 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1, 08/11/2023 N/A NRC Letter from G. Wunder to J. Skolds, Docket No. 50-395, Action 3, Supplement 3, Section 4, Part C, Acceptable Actions 11/27/1991 PWROG-17031-NP-A Update for Subsequent License Renewal: WCAP-15338-A, A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants Rev. 1 05/2020 N/A Virgil C. Summer Nuclear Station, Unit No. 1 - Issuance of Amendment RE: Reactor Coolant System Operational Leakage Technical Specification Surveillance Requirement 4.4.6.2.2 (EPID L-2018-LLA-0267) (ML19023A420) 03/07/2019 RC-98-0177 SCE&G Letter, NRC Bulletin 88-08: Thermal Stresses in Piping Connected to Reactor Coolant Systems 11/09/1998 N/A SCE&G Letter, NRC Bulletin 88-08, Response to NRC Questions 12/19/1988 WCAP-16168-NP-A Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval Rev. 3, 11/2011 CGE-MC000-TR-LG-000007 80-Year Fatigue and Fatigue Exemptions Summary Rev. 3, 07/24/2023
57 factor (CUFen) that is greater than the design limit of 1.0. Therefore, the applicant has performed flaw tolerance evaluations on these nozzles in accordance with the Appendix L provisions. However, SLRA Section B3.1 or 4.3.4 does not clearly discuss whether these nozzles will be inspected at the end of each 10-year inservice inspection interval for the subsequent period of extended operation in accordance with the successive inspection provisions of ASME Code Section XI, Appendix L, L-3420.
The staff also audited the description of the SLRA Fatigue Monitoring AMP provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B3.2, Neutron Fluence Monitoring Summary of Information in the Application. The SLRA states that AMP B3.2, Neutron Fluence Monitoring, is an existing program with exemptions that is consistent with the program elements in GALL-SLR AMP X.M2, Neutron Fluence Monitoring.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date WCAP-18728-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time-Limited Aging Analyses Rev. 5, 06/2023 PWROG-21037-NP Determination of Unirradiated RTNDT and Upper-Shelf Energy Values of the V.C. Summer Nuclear Station Unit 1 Reactor Vessel Materials Rev. 2, 03/2023 WCAP-16298-NP Analysis of Capsule Z from the South Carolina Electric &
Gas Company V. C. Summer Reactor Vessel Radiation Surveillance Program Rev. 0, 10/2004 WCAP-14040-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves Rev. 4, 05/2004 WCAP-16305-NP V. C. Summer Heatup and Cooldown Limit Curves for Normal Operation Rev. 0, 082004 WCAP-18729-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Heatup and Cooldown Limit Curves for Normal Operation Rev. 2, 06/2023 WCAP-18709-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Pressure Vessel Extended Beltline Neutron Exposure Evaluation Rev. 1, 02/2022 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
SLRA AMP B3.3, Environmental Qualification of Electric Equipment Summary of Information in the Application. The SLRA notes that AMP B3.3, Environmental Qualification of Electric Equipment, is an existing program that is consistent with the program elements in GALL-SLR Report AMP X.E1, Environmental Qualification of Electric Equipment.
58 Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3303 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Environmental Qualification of Electric Equipment Rev. 0 GMP-100.019 General Maintenance Procedure - EQ Area Temperature Monitoring Procedure Rev. 2 CM-AA-EQ-10 Fleet EQ Program Description Rev. 8 ES-0528 Environmental Qualification Process and Control Rev. 1 NA VCS Operating Experience List Report - Environmental Qualification of Electrical Equipment 09/10/2023 NA VCS Operating Experience Detail Report - Environmental Qualification of Electrical Equipment 07/26/2023 CR-14-02656 Air Supply Solenoid for XVT02843B-MS 05/10/2014 CR-17-04353 AD Administrative Equipment 08/10/2027 CR-18-04679 HR Hydrogen Removal System - Post Accident 10/23/2018 CR-18-04719 Closed Limit Switch for XVG08702B 10/24/2018 CR-21-00051 EQ Program Review Required for EPRI NP-1558 Revision 01/07/2021 DC0011A-159 Thermal Life of EGS Quick Disconnect Connectors (QDC)
(Qual Keys: 375 & 385)
Rev. 5 DC0011A-178 Thermal Life of PYCO 122 Series RTDs (Qual Key: 284)
Rev. 1 DC0011A-027 Qualified Life Calculation for Valcor V526 Series Solenoid Valves - Qual Key: 335 Rev. 3 DC0011A-125 Qualified Life NAMCO EA180 Limit Switches (Qual Key 273)
Rev. 5 DC0011A-153 Reliance Fan Motors - Class H, Type RN Thermal Life Analysis - Qual. Key: 298 Rev. 0 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptancecriteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLRReport AMP.
The staff also audited the description of the SLRA Environmental Qualification of Electric Equipment program provided in the FSAR supplement. The staff verified this descriptionisconsistent with the description provided in the GALL-SLR Report.
SLRA AMP B3.4, Concrete Containment Unbonded Tendon Prestress Summary of Information in the Application. The SLRA states that AMP B3.4, Concrete Containment Unbonded Tendon Prestress, is an existing program with enhancements, and is consistent with the program elements in GALL-SLR Report AMP X.S1, Concrete Containment Unbonded Tendon Prestress. The staff will document its review of the enhancements in the SE.
Audit Activities. For the OE review, the staff conducted its review of the applicants documentation contained in the SLRA and ePortal. For the AMR, the NRC staff evaluated the AMR items requiring further evaluation addresses the recommendations provided in the SLR SRP.
59 The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3352 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Concrete Containment Unbonded Tendon Prestress Rev. 0 ES-0438 Containment Inservice Inspection Program Rev. 1 ISE-4 Containment Inservice Inspection Program Plan 3rd Inspection Interval Rev. 3F SP-228 Surveillance of Reactor Building Post Tensioning System Rev. 16 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program element(s) of the SLRA AMP are consistent with the corresponding element(s) of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
During the audit, the staff made the following observations:
The staff reviewed SLRA B3.4, and noted that enhancements to program elements 1, 2, 4, and 6 will be revised to specify that the trend analyses for tendon post tensioning loss will include trends projected through the end of the subsequent period of extended operation, and elements 5 and 6 predicted trendlines for each surveillance interval will be developed for the subsequent period of extended operation. The SLRA enhancement description is vague and is unclear to the staff how each of the elements are to be enhanced specifically.
SLRA UFSAR supplement section A2.4 states that the Concrete Containment Unbonded Tendon Prestress program is based on ASME XI, IWL and supplemented by requirements of 10 CFR 50.55a(b)(2)(viii). 10 CFR 50.55a(b)(2)(viii) refers to IWL-2512(a) & (b) which instructs the responsible engineer to evaluate inaccessible areas susceptible to degradation. It is unclear which AMP and inspection procedures covers the requirement of 10 CFR 50.55a(b)(2)(viii).
The staff also audited the description of the SLRA Concrete Containment Unbonded Tendon Prestress program provided in the UFSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA TLAA Section 4.1, Identification of Time-Limited Aging Analyses Summary of Information in the Application. SLRA section 4.1, Identification of Time-Limited Aging Analyses, discusses the methodology for the identifying of TLAAs in the SLRA. To verify the implementation of the applicants methodology to identify TLAAs in accordance with 10 CFR Part 54, the staff performed an audit. Issues identified but not resolved in this report will be addressed in the SE.
Audit Activities. The table below lists a document that was reviewed by the staff and was found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date
60 ETE-SLR-2023-3701 V.C. Summer Subsequent License Renewal Project -
Identification of Time-Limited Aging Analysis (TLAA) - Key Word Searches Rev. 0 WCAP-18772 Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 0 During the audit, the staff verified that the applicant has provided its basis and methodology for identifying TLAAs in accordance with disposition of 10 CFR 54.21(c).
During the audit, the staff made the following observations:
The staff reviewed ETE-SLR-2023-3701 and confirmed that current licensing basis and design basis documentation were searched to identify potential TLAAs. The staff noted that specific key words were used during this search that would identify potential TLAAs.
The staff reviewed ETE-SLR-2023-3701 and confirmed that each potential TLAA was reviewed against the six criteria of 10 CFR 54.3(a) and that those that met all six criteria were identified as TLAAs, which require evaluation for the subsequent period of extended operation.
The staff reviewed ETE-SLR-2023-3701, and confirmed that a search of docketed licensing correspondence, the operating license, and the UFSAR identified the active exemptions currently in effect pursuant to 10 CFR 50.12.
The staff reviewed ETE-SLR-2023-3701 and confirmed that these exemptions were then reviewed to determine whether the exemption was based on a TLAA and that there were no 10 CFR 50.12 exemptions identified that are currently in effect that are based upon a TLAA.
SLRA TLAA Section 4.2.1, Neutron Fluence Projections Summary of Information in the Application. VC Summer SLRA Section 4.2.1, Neutron Fluence Projections, discusses the fluence projections that quantify the number of neutrons that contact reactor vessel surfaces. These projections have been used as input to the neutron embrittlement analyses that evaluate the reduction of fracture toughness aging effect resulting from neutron irradiation.
The VC Summer Reactor Pressure Vessel fluence evaluation for SLR was generated with the methods in compliance with RG 1.190. Estimated RV beltline and extended beltline fast neutron (E >1.0 MeV) fluences at the end of 80 years of operation were calculated in WCAP-18709-NP.
These methodologies have been approved by the NRC for the beltline region; that is, materials directly surrounding the core and adjacent materials per 10 CFR 50, Appendix G, Fracture Toughness Requirements.
Audit Activities. In Section 4.2.1 of the SLRA the licensee states use of RAPTOR-M3G and the BUGLE-96 cross section library in accordance with the methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET and WCAP-18124-NP-A Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials. The NRC approved RAPTOR-M3G and FERRET for determination of RPV fluence with two limitations and conditions included in the staff SE.
As part of the audit, the NRC staff requested the licensee to provide a discussion on how the limitations and conditions in WCAP-18124-NP-A are met or are dispositioned.
61 The licensee provided its response to information requested through the Electronic Reading Room. No breakout session was held. The NRC reviewed the response and generated a RCI to get docketed confirmation on the information provided.
The following additional documents were reviewed as part of the audit.
Document Title Revision/Date WCAP-18728-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time-Limited Aging Analyses Rev. 5, 06/2023 PWROG-21037-NP Determination of Unirradiated RTNDT and Upper-Shelf Energy Values of the V.C. Summer Nuclear Station Unit 1 Reactor Vessel Materials Rev. 2, 03/2023 WCAP-16298-NP Analysis of Capsule Z from the South Carolina Electric &
Gas Company V. C. Summer Reactor Vessel Radiation Surveillance Program Rev. 0, 10/2004 WCAP-14040-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves Rev. 4, 05/2004 WCAP-16305-NP V. C. Summer Heatup and Cooldown Limit Curves for Normal Operation Rev. 0, 08/2004 WCAP-18729-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Heatup and Cooldown Limit Curves for Normal Operation Rev. 2, 06/2023 WCAP-18709-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Pressure Vessel Extended Beltline Neutron Exposure Evaluation Rev. 1, 02/2022 SLRA TLAA Sections 4.2.2, Upper-Shelf Energy to 4.2.6 Low Temperature Overpressure Protection Summary of Information in the Application. The staffs audit addresses the following TLAAs collectively due to their relationship in analyzing the reactor vessel neutron embrittlement:
SLRA Section 4.2.2, Upper-Shelf Energy (USE) discusses the analysis for assessing the projected reduction in the USE properties of RPV beltline base metal and weld materials.
The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(ii).
SLRA Section 4.2.3, Pressurized Thermal Shock (PTS) discusses the analysis for assessing PTS of the RPV beltline materials as required by 10 CFR 50.61. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(ii).
SLRA Section 4.2.4, Adjusted Reference Temperature (ART) discusses the analysis for assessing the use of ART projections to adjust the pressure-temperature (P-T) limit curves to account for neutron irradiation effects. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(ii).
SLRA Section 4.2.5, Pressure-Temperature Limits, discusses the assessment of P-T limits for the RPV. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(iii).
SLRA Section 4.2.6, Low Temperature Overpressure Protection, discusses the assessment for the system enabling temperature and relief valve analyses that demonstrate the capability of the relief valves to protect the structural integrity of the RPV
62 and from exceeding the P-T limit curves. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(i).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
Document Title Revision/Date PWROG-21037-NP Determination of Unirradiated RTndt and Upper-Shelf Energy Values of the V.C. Summer Unit 1 Reactor Vessel Materials Rev. 2 WCAP-18728-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time Limited Aging Analyses Rev. 5 Certified Material Test Reports
- Nozzle Shell Plate, Heat# C9955-2
- Nozzle Shell Plate, Heat# C0123-2
- Outlet Nozzle 437B-1, 437B-2, and 437B-3 NA During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(ii) for SLRA Section 4.2.2, 4.2.3, and 4.2.4, its disposition of 10 CFR 54.21(c)(1)(iii) for SLRA Section 4.2.5, and its disposition of 10 CFR 54.21(c)(1)(i) for SLRA Section 4.2.6.
During the audit, the staff made the following observations:
The staff verified that the material information, except as noted below for the inlet and outlet nozzles, (e.g., initial RTNDT, %Cu, %Ni, initial USE, margin values) for the RPV materials contained in SLRA Tables 4.2.2-1, 4.2.2-2, 4.2.3-1, 4.2.4-4, 4.2.4-5, and 4.2.4-6 are:
o consistent with the applicants current licensing basis (e.g., FSAR and license amendments associated with power uprates) o based on information from certified material test reports or fabrication records for the specific material or weld type The staff verified that %Cu for the Inlet Nozzles 436B-1, 436B-2 and 436B-3, and Outlet Nozzles 437B-1, 437B-2, and 437B-3 were not available in the certified material test reports or fabrication records. The staff will document its review of this situation in the SE.
Additionally, based on these verifications above, the staff observed that the appropriate margin value consistent with Regulatory Guide 1.99, Rev. 2, were applied for each reactor vessel beltline material for the purposes of addressing PTS and ART.
The staff noted that WCAP-18728-NP, Rev. 5, provides the applicants assessment of surveillance data applicable to the reactor pressure vessels. The staff reviewed Section 3, Material Property Input, and Appendix A, VCSNS Unit 1 Credibility Evaluation, and verified the following:
o The applicants use of applicable surveillance data is credible in accordance with 10 CFR 50.61 and Regulatory Guide 1.99, Rev. 2, for the following RPV beltline materials:
Heat# 4P4784
63 Intermediate Shell Longitudinal Weld Seams BC & BD Intermediate to Lower Shell Circumferential Weld Seam AB Lower Shell Longitudinal Weld Seams BA & BB Nozzle to Intermediate Shell Circumferential Weld Seam AC o The surveillance data is non-credible in accordance with 10 CFR 50.61 and Regulatory Guide 1.99, Rev. 2, for the following RPV material:
Intermediate Shell 11-1 (Heat# A9154-1)
The staff also audited the description of the TLAAs in SLRA Sections A3.2.2, A3.2.3, A3.2.4, A3.2.5 and A3.2.6 provided in the FSAR supplement. The staff noted a discrepancy in SLRA Section A3.2.3 related to PTS, and the staff will consider issuing an RAI to obtain the necessary information. The staff verified these descriptions in SLRA Sections A3.2.2, A3.2.4, A3.2.5 and A3.2.6 is consistent with the description provided in the SRP-SLR.
SLRA TLAA Section 4.3.1, Transient Cycle Projections for 80 years Summary of Information in the Application. SLRA Section 4.3.1, Transient Cycle Projections for 80 Years, discusses the 80-year transient cycle projections for fatigue analyses. The applicant did not identify the transient cycle projections as a time-limited aging analysis (TLAA) because (1) the transient cycle projections are used only as input to fatigue TLAAs and (2) the specific dispositions of fatigue TLAAs are described in SLRA Sections 4.3.2 through 4.3.5. To verify the adequacy of the transient cycle projections, the staff audited the 80-year cycle projections.
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the 80-year cycle projections for fatigue analyses. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3333 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Operating Experience for AMPs Rev. 0 07/26/2023 N/A VCS Operating Experience Detail Report: Fatigue Monitoring 08/16/2023 CR-13-02110 Unanticipated RCS Cold Leg SI Piping Transients 05/14/2013 CR-21-02543 WESTEM Year-end Report 09/29/2021 CR-19-00729 SG FW Nozzle Thermal Cycling Anomaly 02/25/2019 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 08/11/2023 CGE-MC000-TR-LG-000004 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Transient Set and Cycles Considered in the TLAA Evaluations Rev. 2 08/5/2022 CGE-MC000-TR-LG-000005 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fatigue Usage Identification Rev. 1 08/11/2022 CGE-MC000-TR-LG-000009 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: 80-year Transient Cycle Projection Results Rev. 3 07/24/2023 CGE-MC000-TR-LG-000010 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Coolant System and Auxiliary Line Class 2/3 80-year Transient Cycle Projection Results Rev. 3 07/25/2023 During the audit of the 80-year cycle projections, the staff verified that the applicant has
64 provided its basis that supports the adequacy of the cycle projections. However, the staff found that sufficient information was not available to complete its review of the cycle projections. To obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations.
SLRA Section 4.3.1 indicates that linear-rate cycle extrapolation of the total past operating period (i.e., cycle accumulation since the start of the operation through December 31, 2019) was used to project the number of future cycles for 80 years of operation. However, the applicant did not clearly discuss whether the recent cycle data (e.g., most recent 10-year cycle data up to December 31, 2019) suggest an accelerated cycle accumulation rate for certain transients in comparison with the cycle accumulation rate based on the entire past operating period (i.e., cycle accumulation rate since the start of operation).
Table 3-2 of CGE-MC000-TR-LG-000010, Rev. 3 lists the 80-year projected cycles of the transients for the RCS and the auxiliary piping systems connected to the RCS. However, the following RCS transients listed in the table are not listed as a design transient in SLRA Table 4.3.1-1:
- 1.
feedwater cycling at hot shutdown transient
- 2.
reduce temperature return to power transient
- 3.
unit loading between 0 and 15 percent of full power transient
- 4.
unit unloading between 0 and 15 percent of full power transient
- 5.
loop out of service - normal loop shutdown transient
- 6.
loop out of service - normal loop startup transient
- 7.
boron concentration equalization transient
- 8.
refueling transient
- 9.
inadvertent reactor coolant system depressurization transient
- 10. inadvertent startup of an inactive loop transient
- 11. control rod drop transient
- 12. inadvertent safety injection actuation transient
- 13. secondary side leakage test transient
- 14. steam generator tube leakage test transient In addition, Table 3-2 of CGE-MC000-TR-LG-000010, Rev. 3 does not include the information on the inadvertent auxiliary spray transient listed in SLRA Table 4.3.1-1.
The staff also audited the description of the 80-year cycle projections provided in the FSAR supplement. The staff verified this description is consistent with the guidance in SRP-SLR Section 4.3.2.2.
SLRA TLAA Section 4.3.2, ASME Code,Section III, Class 1 Fatigue Analyses Summary of Information in the Application. SLRA Section 4.3.2, ASME Code,Section III, Class 1 Fatigue Analyses, discusses the fatigue TLAAs for ASME Code Section III, Class 1 piping and components. The applicant dispositioned the fatigue TLAAs for the Class 1 piping including the pressurizer surge line in accordance with 10 CFR54.21(c)(1)(i). In addition, the
65 applicant dispositioned the fatigue TLAAs for the Class 1 components, including the fatigue waiver evaluation, in accordance with 10 CFR54.21(c)(1)(iii).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the fatigue TLAAs for the ASME Code Section III, Class 1 piping and components.
The staff will document its review of this information in the SE.
Document Title Revision/Date CR-13-02110 Unanticipated RCS Cold Leg SI Piping Transients 05/14/2013 CR-21-02543 WESTEM Year-end Report 09/29/2023 ETE-SLR-2023-3301 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Fatigue Monitoring Program Rev. 0 08/03/2023 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 08/11/2023 CGE-MC000-TR-LG-000004 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Transient Set and Cycles Considered in the TLAA Evaluations Rev. 2 08/05/2022 CGE-MC000-TR-LG-000005 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fatigue Usage Identification Rev. 1 08/11/2022 CGE-MC000-TR-LG-000009 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: 80-year Transient Cycle Projection Results Rev. 3 07/24/2023 CGE-MC000-TR-LG-000010 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Coolant System and Auxiliary Line Class 2/3 80-year Transient Cycle Projection Results Rev. 3 07/25/2023 WCAP-9401 Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly Rev. 0, 03/1979 WCAP-7332-L-AR Topical Report Indian Point Unit No. 2 Reactor Internals Mechanical Analysis for Blowdown Excitation 11/1973 WNEP-8541 Fatigue of Steam Generator Primary Manway Closure Bolts Test Report 09/1985 Work Order 1808890-011 SG Studs ASME Section XI VT-1 Results 2018 11/08/2018 WCAP-12785 Structural Evaluation of the Vigil C. Summer Pressurizer Surge Line, Considering the Effects of Thermal Stratification Rev. 0, 12/1990 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(i) for the Class 1 piping including the pressurizer surge line and its disposition of 10 CFR 54.21(c)(1)(iii) for the Class 1 components including the fatigue waiver evaluation. However, the staff found that sufficient information was not available to complete its review of the applicants basis for its TLAA disposition. To obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations. Table 2-1 of WCAP-12785, Rev. 0 lists the surge line design transients involving thermal stratification. These transients are used in the existing CUF analysis for the surge line but are not tied to the heatup/cooldown transients.
However, some of these transients (e.g., feedwater cycling and loop out of service inactive loop transients) are not described in SLRA Table 4.3.1-1. In addition, the SLRA does not clearly discuss how the applicant determined that the projected cycles for 80 years of operation
66 do not exceed the surge line design transient cycles that are not associated with the heatup/cooldown transients discussed above.
The staff also audited the description of the fatigue TLAAs for the Class 1 piping and components provided in the FSAR supplement. The staff verified this description is consistent with the guidance in SRP-SLR Section 4.3.2.2.
SLRA TLAA Section 4.3.3, Non-Class 1 Allowable Stress Analyses Summary of Information in the Application. SLRA Section 4.3.3, Non-Class 1 Allowable Stress Analyses, discusses the TLAAs on non-Class 1 allowable stresses. The applicant dispositioned the non-Class 1 allowable stress TLAAs in accordance with 10 CFR54.21(c)(1)(i).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the TLAA on non-Class 1 allowable stresses. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3333 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Operating Experience for AMPs Rev. 0 07/26/2023 N/A VCS Operating Experience Detail Report: Fatigue Monitoring 08/16/2023 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 08/11/2023 CGE-MC000-TR-LG-000004 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Transient Set and Cycles Considered in the TLAA Evaluations Rev. 2 08/05/2022 CGE-MC000-TR-LG-000005 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fatigue Usage Identification Rev. 1 08/11/2022 CGE-MC000-TR-LG-000010 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Reactor Coolant System and Auxiliary Line Class 2/3 80-year Transient Cycle Projection Results Rev. 3 07/25/2023 ETE-SLR-2023-3231 Non-Class 1 Allowable Stress Analyses Rev. 0 05/25/2023 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(i). However, the staff found that sufficient information was not available to complete its review of the applicants basis for its TLAA disposition. To obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations. Table 3-1 of CGE-MC000-TR-LG-000010, Rev. 3 describes the 80-year projected cycles of the following piping lines and systems to demonstrate that the projected cycles of the non-Class 1 piping systems connected to the RCS do not exceed 7000 cycles:
reactor coolant line residual heat removal piping safety injection accumulator piping cold leg safety injection piping normal/alternate charging branch line normal letdown/excess letdown branch line
67 pressurizer safety and relief line pressurizer spray and auxiliary spray line The number of the projected cycles for the reactor coolant line is used to estimate the RCS cycles that contribute to the total projected cycles of the non-Class 1 piping systems connect to the RCS. However, SLRA Table 4.3.3-1 does not clearly describe the 80-year projected cycles for these piping systems and how the applicant determined these projected cycles.
The staff also audited the description of the non-Class 1 allowable stress TLAAs provided in the FSAR supplement. The staff verified this description is consistent with the guidance in SRP-SLR Section 4.3.2.2.
SLRA TLAA Section 4.3.4, Environmentally-Assisted Fatigue Summary of Information in the Application. SLRA Section 4.3.4, Environmentally-Assisted Fatigue, discusses the EAF TLAA for ASME Code Section III, Class 1 piping systems and components that contact reactor coolant. The applicant dispositioned the TLAA in accordance with 10 CFR54.21(c)(1)(iii).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the EAF TLAA for the ASME Code,Section III, Class 1 piping and components. The staff will document its review of this information in the SE.
Document Title Revision/Date N/A VCS Operating Experience Detail Report: Fatigue Monitoring 08/16/2023 CR-13-02110 Unanticipated RCS Cold Leg SI Piping Transients 05/14/2013 CR-21-02543 WESTEM Year-end Report 09/29/2023 ETE-SLR-2023-3301 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Fatigue Monitoring Program Rev. 0 08/03/2023 PWROG-17031-NP-A Update for Subsequent License Renewal: WCAP-15338-A, A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants Rev. 1, 05/2020 WCAP-16168-NP-A Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval Rev. 3, 10/2011 CGE-MC000-TR-LG-000005 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Equipment and Piping Environmentally-Assisted Fatigue Screening Evaluation Results Rev. 1 08/11/2022 CGE-MC000-TR-LG-000013 V.C Summer Nuclear Station Unit 1 Subsequent License Renewal: Sentinel Location Reduction Rev. 0 01/25/2023 CGE-MC000-TR-LG-000011 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Environmentally Assisted Fatigue Evaluations Summary Report Rev. 3 06/19/2023 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 08/11/2023 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(iii). However, the staff found that sufficient information was not available to complete its review of the applicants basis for its TLAA
68 disposition. To obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations:
SLRA Section 4.3.4 does not clearly describe how the applicant calculated the most conservative Fen values in the EAF screening evaluation. The SLRA does not clearly discuss how the applicant determined that the temperature, strain rate, and sulfur content (as applicable) in the Fen (environmental fatigue correction factor) calculations used for the screening process. In addition, the SLRA does not discuss how the applicant refined the screening CUFen values to remove the associated conservatism in the screening CUFen calculations.
SLRA Section 4.3.4 explains that among the limiting EAF locations listed in SLRA Table 4.3.4-1, flaw tolerance evaluations were performed on the normal and alternate charging cold leg nozzles and pressurizer surge line hot leg nozzle in accordance with ASME Code Section XI, Appendix L. The staff noted that the corrective actions program element of GALL-SLR AMP X.M1 states that a flaw tolerance analysis with appropriate (e.g., inclusion of environmental effects) crack growth rate curves and associated inspections performed in accordance with Appendix L of ASME Code Section XI is an acceptable correction action.
ASME Code Section XI, Appendix L, L-3420, Successive Examinations specifies that if the allowable operating period per the flaw tolerance analysis is equal to or greater than 10 years, the locations subject to the flaw tolerance analysis be inspected at the end of each inservice inspection interval (i.e., 10-year inspection interval). However, the SLRA does not clearly address the following items:
- 1. whether the limiting locations subject to flaw tolerance evaluations are inspected by using ultrasonic testing examination
- 2. whether these nozzles are included in the examination item for thermal fatigue in accordance with the risk-informed inservice inspection program based on ASME Code Case N-716-1 (SLRA Section B2.1.1)
- 3. when the baseline inspection for the flaw tolerance evaluations will be performed to confirm the absence of an unacceptable flaw per Appendix L, L-3410 prior to the subsequent period of extended operation
- 4. whether the inspections will be performed at the end of each 10-year inservice inspection interval in accordance with the ASME Code Section XI, Appendix L provisions (L-3420, Successive Examinations) and, if not, whether these nozzles will be inspected at least once during the subsequent period of extended operation.
The following reference indicates that the pressurizer lower head at the penetration (low alloy steel lower head location) has an 80-year projected CUFen is 0.994 (Ref.:CGE-MC000-TR-LG-000011, Rev. 3, Table 4-1). However, SLRA Table 4.3.4-1 does not identify the pressurizer lower head at the heater penetration (low alloy steel location) as a limiting EAF location. Even though the 80-year CUFen for the pressurizer lower head location is almost identical to 1.0, the SLRA does not clearly discuss the following information:
- 1. whether the CUFen value for the pressurizer lower head location is based on the 80-year cycle projections or design cycles
- 2. why the pressurizer lower head location is not identified as a limiting EAF location in SLR Section 4.3.4
- 3. whether the location will be monitored by the Fatigue Monitoring AMP and, if not, how the applicant will ensure the CUFen of the location will not exceed the design limit of 1.0
69
- 4. if the CUFen value approaches the design limit, whether a corrective action (e.g.,
refinement of CUFen or repair/replacement activities) will be performed to ensure that the CUFen does not exceed the design limit WCAP-18772-P, Table 4-26 and SLRA Table 4.3.4-1 address the CUFen values for the following EAF locations: (a) safety injection 6-inch reactor coolant line cold-leg nozzle and (b) residual heat removal 6-inch hot leg nozzle. However, the applicant did not credit the Fatigue Monitoring AMP to manage the effects of EAF for these nozzle locations. In addition, SLRA Section 4.3.4 does not clearly discuss the following items related to these locations:
- 1. whether the design cycles or 80-year projected cycles are used in the CUFen calculations
- 2. why the Fatigue Monitoring AMP is not credited to manage the aging effects of EAF for these components
- 3. how the applicant will ensure the CUFen of these locations does not exceed the design limit of 1.0
- 4. whether a corrective action will be taken (e.g., refinement of CUFen or repair/replacement activities) if the CUFen of locations approaches the design limit of 1.0 The staff also audited the description of the EAF TLAA for the Class 1 piping and components provided in the FSAR supplement. The staff verified this description is consistent with the guidance in SRP-SLR Section 4.3.2.2.
SLRA TLAA Section 4.3.5, High-Energy Line Break Analysis Summary of Information in the Application. SLRA Section 4.3.5, High-Energy Line Break Analysis, discusses the high-energy line break (TLAA) TLAA for ASME Code Section III, Class 1 piping systems. The applicant dispositioned the TLAA in accordance with 10 CFR54.21(c)(1)(i).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the TLAA on high energy line break (HELB) location postulation. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3333 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Operating Experience for AMPs Rev. 0 07/26/2023 N/A VCS Operating Experience Detail Report: Fatigue Monitoring 08/16/2023 WCAP-18772-NP/P Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 08/11/2023 CGE-MC000-TR-LG-000004 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Transient Set and Cycles Considered in the TLAA Evaluations Rev. 2 08/5/2022 CGE-MC000-TR-LG-000005 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fatigue Usage Identification Rev. 1 08/11/2022 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(i). However, the staff found that sufficient information was not available to complete its review of the applicants basis for its TLAA disposition. To obtain the necessary information, the staff will consider issuing RAIs.
70 During the audit, the staff made the following observations:
SLRA Section 4.3.5 explains that the HELB analysis for the ASME Code Class 1 piping is based on a CUF threshold of 0.1 for postulating the intermediate-piping break locations in the HELB analysis. In comparison, FSAR Section 3.6.2.1.2 also indicates that the postulation of HELB locations for ASME Code non-Class 1 piping is, in part, based on the allowable stress range for expansion stress (SA), consistent with Branch Technical Position MEB 3-1 (ML19137A335).
Expansion stress may need to be adjusted by a stress range reduction factor that is determined by the number of thermal cycles, as addressed in the implicit fatigue analysis in SLRA Section 4.3.3. However, SLRA Sections 4.3.5 and A3.3.5 (FSAR supplement) do not clearly discuss whether the HELB location postulation for the non-Class 1 piping, which involves SA, is a basis for identifying the HELB analysis as a TLAA.
The staff also audited the description of the TLAA on HELB location postulation provided in the FSAR supplement. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the TLAA on HELB location postulation. The staff will consider issuing an RAI to obtain the information necessary to verify the sufficiency of the FSAR supplement description in relation to the HELB analysis discussed above.
SLRA TLAA Section 4.4, Environmental Qualification of Electric Equipment Summary of Information in the Application. SLRA Section 4.4, Environmental Qualification of Electric Equipment, discusses the analyses for the thermal, radiation, and cyclical aging of plant electrical components. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(iii).
Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3303 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report -
Environmental Qualification of Electric Equipment Rev. 0 GMP-100.019 General Maintenance Procedure - EQ Area Temperature Monitoring Procedure Rev. 2 CM-AA-EQ-10 Fleet EQ Program Description Rev. 8 ES-0528 Environmental Qualification Process and Control Rev. 1 NA VCS Operating Experience List Report - Environmental Qualification of Electrical Equipment 09/10/2023 NA VCS Operating Experience Detail Report - Environmental Qualification of Electrical Equipment 07/26/2023 CR-14-02656 Air Supply Solenoid for XVT02843B-MS 05/10/2014 CR-17-04353 AD Administrative Equipment 08/10.2027 CR-18-04679 HR Hydrogen Removal System - Post Accident 10/23/2018 CR-18-04719 Closed Limit Switch for XVG08702B 10/24/2018 CR-21-00051 EQ Program Review Required for EPRI NP-1558 Revision 01/07/2021 DC0011A-159 Thermal Life of EGS Quick Disconnect Connectors (QDC)
(Qual Keys: 375 & 385)
Rev. 5 DC0011A-178 Thermal Life of PYCO 122 Series RTDs (Qual Key: 284)
Rev. 1
71 Document Title Revision/Date DC0011A-027 Qualified Life Calculation for Valcor V526 Series Solenoid Valves - Qual Key: 335 Rev. 3 DC0011A-125 Qualified Life NAMCO EA180 Limit Switches (Qual Key 273)
Rev. 5 DC0011A-153 Reliance Fan Motors - Class H, Type RN Thermal Life Analysis - Qual. Key: 298 Rev. 0 During the audit of the TLAA, the staff verified that the applicant has provided itsbasis that supports its disposition of 10CFR54.21(c)(1)(iii).
The staff also audited the description of the SLRA Environmental Qualification of Electric Equipment provided in the FSAR supplement. The staff verified this descriptionisconsistent with the description provided in the SRP-SLR.
SLRA TLAA Section 4.5, Concrete Containment Tendon Prestress Analysis Summary of Information in the Application. SLRA Section 4.5, Concrete Containment Tendon Prestress Analysis, discusses the regression analysis on the containment unbonded tendon force trend lines in order the forecast the tendon forces for the subsequent period of extended operation. The applicant dispositioned the TLAAs in accordance with 10 CFR54.21(c)(1)(iii).
Audit Activities For the OE review, the staff reviewed the applicants documentation contained in the SLRA and ePortal, including documents associated with GALL XI.S1 AMP.
The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3352 V.C. Summer Subsequent License Renewal Project -
Aging Management Program Evaluation Report - Concrete Containment Unbonded Tendon Prestress Rev. 0 SP-228 Surveillance of Reactor Building Post Tension System Rev. 16 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(iii).
During the audit, the staff made the following observation:
The staff reviewed SLRA Section 4.5 and Figures 4.5-3 and noted that the predicted lower confidence limit of the prestress tendon force falls below the minimum required value at T = 77.8, which is prior to the subsequent period of extended operation. The SLRA also notes corrective actions will be completed if the measured tendon force trend lines predict the prestressing force will fall below the minimum required value before the next scheduled examination.
The staff also audited the description of the SLRA TLAA for Concrete Containment Tendon Prestress Analysis provided in the UFSAR supplement in SLRA Appendix A3.5. The staff verified this description is consistent with the description provided in the SRP-SLR for TLAAs.
72 SLRA TLAA Section 4.6, Containment Liner Plate Summary of Information in the Application. SLRA Section 4.6, Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis, discusses that the 80-year fatigue waiver used 200 startup and shutdown cycles, exceeding the two startup/ shutdown cycles-per-year assumption used previously. The 80-year transient cycle projection predicts 148 startup/shutdown cycles through 80 years, based upon trending of actual transient occurrences since the beginning of plant operation. The applicant dispositioned the TLAA in accordance with 10 CFR54.21(c)(1)(ii).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3204 Identification and Evaluation of Potential TLAA for Reactor Building Containment Liner and Other Class MC Components Rev. 2 ETE-SLR-2023-3205 Identification and Evaluation of Potential TLAA for Reactor Building Containment Penetrations Rev. 2 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(ii).
The staff also audited the description of the Containment Liner Plate Evaluation TLAA provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the SRP-SLR.
SLRA TLAA Section 4.7.1, Crane Load Cycle Limits Summary of Information in the Application. SLRA Section 4.7.1, Crane Load Cycle Limits, discusses the 80-year load cycle limits for each evaluated crane, the total number of projected load cycles is less than the maximum design allowable 100,000 cycles for Service Class A cranes. Therefore, the load cycle analyses remain valid for the subsequent period of extended operation. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(i).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3316 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) handling Systems Rev. 0 CMAA Specification # 70 Specifications for Electric Overhead Traveling Cranes Revised 1975 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(i).
The staff also audited the description of the Crane Load Cycle Limits TLAA provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the SRP-SLR.
73 SLRA TLAA Section 4.7.2, Reactor Coolant Pump Flywheel Fatigue Crack Growth Analyses Summary of Information in the Application. SLRA Section 4.7.2, Reactor Coolant Pump Flywheel Fatigue Crack Growth Analyses, discusses the potential failure due to the growth of an undetected fabrication flaw, which could grow to the critical flaw size during normal or accident conditions for the Reactor Coolant Pump (RCP) Flywheel. The applicant dispositioned the TLAA in accordance with 10 CFR 54.21(c)(1)(i).
Audit Activities. In addition to the OE contained in the SLRA and ePortal, the staff observed the applicants search of its operating experience database using the keyword: (crack and fatigue).
The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3238 CGE-MP010-TR-CF-000002 - V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Assessment of Reactor Coolant Pump Flywheel Fatigue Crack Growth Rev. 1 DOM-SLR-21-012 Subject - Design Input Request (DIR) Flywheel (WS08a) 11/04/2021 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(i). During the audit, the staff observed that the assumed 10 RCP start/stop cycles per heatup/cooldown cycle were supported by operator interviews, as transmitted as a design input to Westinghouse.
The staff also audited the description of the RCP Flywheel Fatigue Crack Growth TLAA provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the SRP-SLR Section 4.7.
SLRA TLAA Section 4.7.3, Leak-Before-Break (Plant-Specific TLAA)
Summary of Information in the Application: SLRA Section 4.7.3 Leak-Before-Break describes the V.C. Summers TLAA on the leak-before-break (LBB) evaluation for the reactor coolant system (RCS) piping. V.C. Summer Nuclear Station (VCSNS) dispositioned the TLAA with 10 CFR 54.21(c)(1)(ii) for the reactor coolant system loop piping.
To verify that V.C. Summer provided a basis to support its disposition of the TLAA, the staff audited the TLAA. The staff will address the issues identified but not resolved in this report in the SE.
Audit Activities: For the OE review, the staff conducted its review of the applicants methodology and operating experience by reviewing documentation contained in the SLRA and ePortal. The staff conducted its review of the VCSNS OE database using the following keywords: leak-before-break, rupture, crack. leak rate, leaks in RCS primary piping, and through wall leaks in RCS primary piping.
The table below lists the documents that were reviewed by the staff and were found to be relevant to TLAA Section 4.7.3, Leak-Before-Break. These documents were identified in the staffs search of VCSNSs OE database. During the audit of the operating experience program element, the staffs independent database search did not identify any OE that would indicate the LBB analysis may not be adequate to manage the aging effects.
74 During the audit breakout session, the staff interviewed VCSNS and Westinghouse staff and reviewed documentation provided by VCSNS. The staff reviewed relevant documents listed above. The staff also verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(ii). The staff found sufficient information was available to complete its review of the applicants basis for its TLAA disposition. If further information is needed to complete the staffs SE input, the staff will consider issuing RAIs.
The staff also audited the description of the SLRA TLAA LBB provided in the UFSAR supplement. The staff verified that this description is consistent with the information provided in the FSAR supplement of the SLRA TLAA LBB. If further information is needed to complete the staffs SE input, the staff will consider issuing RAIs.
SLRA TLAA Section 4.7.4, Steam Generator Tube Wear Evaluation Summary of Information in the Application. SLRA Section 4.7.4, Steam Generator Tube Wear Evaluation, discusses the wear analysis for the steam generator tubes. The applicant dispositioned the TLAA in accordance with 10 CFR54.21(c)(1)(iii).
Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
Document Title Revision/Date WNEP-9314 Delta 75 Steam Generator - Flow Induced Vibration and Tube Wear/Corrosion Evaluation Rev. 0 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(iii).
The staff also audited the description of the Steam Generator Tube Wear Evaluation TLAA Document Title Revision/Date NUREG-0800, Section 3.6.3 Leak-Before-Break Evaluation Procedures Rev. 1, 03/2007 NUREG/CR-4513 Estimation of Fracture Toughness of Cast Stainless Steel During Thermal Aging in LWR Systems Rev. 1, 05/1994 NUREG/CR-4513 Estimation of Fracture Toughness of Cast Stainless Steel During Thermal Aging in LWR Systems Rev. 2, 05/2016 WCAP-13206 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Virgil C. Summer Nuclear Plant Rev. 1. 06/2000 WCAP-13206 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Virgil C. Summer Nuclear Plant Rev. 2. 07/2001 WCAP-13206 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Virgil C. Summer Nuclear Plant Rev. 3, 01/2018 SLR-AMP-XI.M01 ASME Section XI Inservice Inspection SLR-AMP-XI.M02 Water Chemistry SLR-AMP-XI.M32 One-Time Inspection WCAP-13206 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Virgil C. Summer Nuclear Plant Rev. 4, 04/2022
75 provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the SRP-SLR.
AMR Items Not Associated with an AMP SLRA AMR 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation Summary of Information in the Application. During the audit, the staff reviewed plant documentation associated with the following:
SLRA Section 3.5.2.2.2.6, which describes the applicants Further Evaluation or FE of effects of aging on the Primary Shield Wall (PSW) and Secondary Shield Wall concretes, grout, the RV steel supports, structural bolting and the embedded wide-flange (WF) steel sections and welded plates within the PSW exposed to neutron and gamma radiation and radiation induced heating in air-indoor uncontrolled environment. The SLRA further evaluation is associated with Table 3.5.1 Items 3.5.1-097 for concrete, 3.5.1-055 for grout, 3.5.1-027, 3.5.1-057, 3.5.1-077 for steel, and 3.5.1-087, 3.5.1-089 for structural bolting.
Associated Table 2 AMR items associated with this further evaluation are:
SLRA Table 3.5.2-1 Containment Structure - Aging Management Evaluation, which cites the PSW biological shield wall concrete (reinforced) exposed to air-indoor uncontrolled environment associated with Table 1 item 3.5.1-097.
Relevant Table 2 AMR items but not determined yet whether can be associated with this further evaluation are:
SLRA Table 3.5.2-15 Structures and Component Supports -NSSS Supports - Aging Management Evaluation, which cites grout for reduction in concrete anchor capacity exposed to air-indoor uncontrolled environment associated with Table 1 item 3.5.1-055 SLRA Table 3.5.2-1 Containment Structure - Aging Management Evaluation, which cites metal accessories (liner, metal plate, etc.) exposed to air-indoor uncontrolled environment associated with Table 1 item 3.5.1-027 SLRA Table 3.5.2-15 Structures and Component Supports - NSSS Supports - Aging Management Evaluation, which cites guides exposed to air-indoor uncontrolled environment associated with Table 1 item 3.5.1-57 SLRA Table 3.5.2-1 Containment Structure - Aging Management Evaluation, which cites steel elements (steel components: all structural steel) exposed to air-indoor uncontrolled environment or embedded associated with Table 1 item 3.5.1-077 SLRA Table 3.5.2-15 Structures and Component Supports - NSSS Supports - Aging Management Evaluation, which cites high strength structural bolting exposed to air-indoor uncontrolled environment associated with Table 1 item 3.5.1-87.
SLRA Table 3.5.2-1 Containment Structure - Aging Management Evaluation, which cites steel elements (support members; welds; bolted connections; support anchorage to building structure) exposed to air with borated water leakage environment or embedded associated with Table 1 item 3.5.1-89
76 Audit Activities. The table below lists documents that were reviewed by the staff and were found relevant to SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation. The staff will document its review of this information in the SE.
Document Title Revision/Date GILBERT 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.
Rev. 0 (11/15/1976)
DWG S-423-046 Structural - Anchor Bolt List. Reactor Vessel Support Anchor Rev. 2 DWG E-511-231 Reactor Building. Reactor Vessel Special Steel. Erection Plan at El. 422 -8 Rev. 7 DWG E-511-232 Reactor Building. Reactor Vessel Special Steel. Erection Plan at El. 426 1/16 Rev. 19 DWG E-511-233 Reactor Building. Reactor Vessel Special Steel. Erection Plan at El. 430 -9 Rev. 9 DWG E-511-234 Reactor Building. Reactor Vessel Special Steel. Erection Plan at El. 437 -21/2 Rev. 4 DWG E-511-235 Reactor Building. Reactor Vessel Special Steel. Liner Plate Erection Stretch-Out Rev. 4 DWG E-511-236 Reactor Building. Reactor Vessel Special Steel. Erection -
Sections and Details Rev. 7 DWG E-511-237 Reactor Building. Reactor Vessel Special Steel. Erection -
Sections and Details Rev. 2 DWG E-511-216 Reactor Building Reactor Vessel Support Anchors Sections and Details Rev. 10 DWG E-511-214 Reactor Building Reactor Vessel Support Anchors Sections and Details Rev. 9 DWG E-511-215 Reactor Building Reactor Vessel Support Anchors Sections and Details Rev. 9 DWG E-511-213 Reactor Building Reactor Vessel Support Anchor Plan El.
426-6 Rev. 2 DWG E-511-223 Reactor Building Liner Plate Sections and Details Rev. 4 DWG E-511-224 Reactor Building Reactor Vessel Liner Plate Sections and Details Rev. 6 DWG E-511-219 Pipes Sleeves and Liner Plate Sectional Plans Rev. 9 DWG E-511-217 Reactor Building Cover Plates Rev. 9 DWG E-511-251 Reactor Building Interior Walls Primary Shield Wall and Fuel Transfer Canal Walls Rev. 8 DWG E-511-222 Reactor Building Baffle Location Rev. 15 DWG E-411-283 Reactor Building Primary Shield Wall Sections Rev. 3 DWG E-411-280 Reactor Building Primary Shield Wall Sections Rev. 4 DWG E-411-275 Reactor Building Primary Shield Wall Sectional Plan at Elev.
417-11 Rev. 1 WESTINGHOUSE 1099E34 Standard 3 Loop Plant RV Support Hardware Details and Assembly Rev. 4 1095E19 Primary Loop Layout Standard 3 Loop Plant Without Stop Valves Rev. 0 583F788/(see also DOM DWG 1MS 057 - Sh. 1)
RCS Equipment Supports. Reactor Vessel Supports (General Notes 1through 14)
Rev. 13
77 Document Title Revision/Date CGE-GENW-TR-LG-000003 VC Summer Nuclear Station Unit 1 Subsequent License Renewal: Input to Section 3.5.2.2.2 6 Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation Rev. 0 CGE-CA120-CN-SA-000001 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment Rev. 1 CGE-CA120-TM-SA-000001 VC Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment Rev. 1 WCAP-18772-P Proprietary Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Rev. 1 WCAP-18785-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports Rev. 2 WCAP-18785-P Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports Revs 1 and 2 WCAP-18785-P Proprietary DRAFT FEA Markup Rev. 0 CGE-REAC-TM-AA-000005 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Bioshield Concrete Neutron Exposure Evaluation Rev. 1 CGE-MC000-TR-LG-000005 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Fatigue Usage Identification Rev. 1 CGE-MC000-TR-CF-000005 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Equipment and Piping Environmentally Assisted Fatigue Screening Evaluation Results Rev. 1 CGE-MC000-TR-LG-000004 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Transient Set end Cycles Considered in TLAA Evaluations Rev. 2 CGE-MC000-TR-LG-0000011 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Environmentally Assisted Fatigue Evaluations Summary Report Rev. 3 CGE-MC000-TR-LG-0000013 Proprietary V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Sentinel Location Reduction Rev. 0 Response from WEC Supplemental Information for Question #8 Rev. 0 Response from WEC Supplemental Information for Question #6 Rev. 0 DOMINION TR00170-003 Structures Aging Management Review for License Renewal Rev. 1 SP-201 Engineering Services Specification: Structural Concrete Rev. 7B VCSNS-DBD Class 1 Piping (CP)
Rev. 5 DC 07020-002 RB Reactor Compartment Cooling System Pressure Loss Rev. 3 DC 00020-209 Thermal Analysis Reactor Cavity Rev. 1 DC 00020-209 Thermal Analysis Reactor Cavity Rev. 0 PG-15923 Pittsburgh Testing Laboratory Report of Test 713 Master Builders Grout for South Carolina Electric and Gas 02/20/1976 GAI/RE 1275 Memorandum Gilbert/Commonwealth Grouting Westinghouse of Reactor Vessel Supports 09/26/1977 CR-00-01324 Boric acid leak and accumulation in A loop RB412 from HL Air Boot area 10/07/2000 CR-13-00737 Administrative Asset. RV Supports not Included in the Licensees ASME Code Section XI, ISI Program (failure to perform examinations of reactor pressure vessels supports).
06/07/2013
78 Document Title Revision/Date CR-15-05177 Reactor Vessel Shoe & Shims Package (includes RV steel support ISI in 2015 per Section XI per IWF-3410 acceptance criteria) 10/19/2015 SP-625-044461-000 Specification for Special Steel for Reactor Building, Virgil C.
Summer Nuclear Station Unit 1 05/12/1975 G678888 Proprietary Reactor Coolant System Component Supports Equipment Specification Rev. 0 1MS-07-057 - Sh. 2 RCS Equipment Supports. Reactor Vessel Supports (General Notes 15 through 21 - Shimming, Ref DWG 583F788)
Rev. 1 ECR 50874L New Chiller Mod Reactor BLDG:
Primary Shield Wall Design Review 4461 VCS PSW Summary Reinforcement from PSW AOR 12/03/1975 076 VCSNS FE Irradiated Concrete Steel Breakout Questions (proprietary)
Responses to breakout Audit questions Rev. 4 EPRI 3002018400 2020 Update to Irradiation of Concrete Guidance. Basis for Evaluation of Concrete Biological Shield Wall for Aging Management.
Rev. 1 3002008129 Long-Term Operations: Impact of Radiation Heating on PWR Biological Shield Concrete Rev. 0 3002013084 Long-Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Component (Structural Tools)
Rev. 0 3002011710 Irradiation damage of the concrete biological shield. Basis for evaluation of the concrete biological shield wall for aging management Rev. 0 During the audit, the staff made several observations on the RV support system.
The staff will conduct a limited scope audit (ML24109A115) to discuss the first set of observations made below (parts a through r). The staff will document its review of this information and resolution of the issues in the SE.
GAI-AOR = Structural Engineering Calculation No. 1.53. Reactor Building Interior Concrete Area. Reactor Vessel Support Anchor Assembly/Anchor Assembly under RV Support - Primary Shield Wall W-AOR = CGE-CA120-CN-SA-000001, Rev. 1, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment a) It appears from Gilbert and Associates Inc. analysis of record (GAI-AOR) that the noted entries in SLRA Table 3.5.2.2.2.6-2, Summary of SLR Demand to Capacity Ratios for PSW Concrete, come from the single loading combination of accident, temperature, and pressure loads. Were there any other loading combinations that contained in addition to large loss-of-coolant accident (LOCA) loads, jet and seismic forces? Was the W-AOR evaluation considered the recalculated reinforcing steel at the same locations to match those of the GAI-AOR?
79 b) It is not clear what is the air gap of the RV annulus so that there is unrestricted air flow to the PSW. Design Basis Document, Rev. 9, states an air gap of 3 inches, the site reports (reference: DC07020-002, Rev. 3) 1.75 inches maximum and 1.5 inches minimum. Please clarify what is the current unrestricted air gap between the PSW and the RV with and without insulation.
c) Regarding gamma heating, discuss the concrete compressive strength for 10 inches or more portions of the PSW. Reference(s) indicate that potentially, it could vary from 0.9 multiplied by 5,381 to 0.9 multiplied by 6,083 psi or 4,843 to 5,475 psi (reference: 4 to Calcs DC0020-209, Rev. 0, SP-201, ACI 318, and EPRI documentation).
d) It appears that the W-AOR calculations are based on a 2.5-inch concrete clear cover. The clear cover in construction drawings is shown to be 4.5 inches (vertical bars). What was the clear cover used in the legacy and W-AOR calculations? Is there enough margin in structural concrete calculations at Effective Full-Power Years (EFPY) 72 to obviate the concern of its more accurate representation?
e) The W-AOR states that given the conservatisms in the analyses quantification other considerations are not necessary. There is no discussion that describes the other considerations that can be omitted from the analyses. Please discuss/describe these other considerations.
f) Confirmation needed that the PSW strength calculations are based on a 3,000 psi compressive strength concrete. Concrete construction drawings state that the compressive strength of concrete is 5,000 psi at 90 days. Was the 3,000 psi concrete compressive strength used for the corbel, the PSW, and other relevant concrete construct designs to loading conditions and combinations (including static and dynamic loads resulting from the mass/weight of the RV)?
g) The GAI-AOR states that the concrete carries vertical loads and the embedded steel caries lateral loads. The W-AOR states that concrete carries vertical loads, lateral loads, and torsional loads. Are both correct? Please clarify.
h) The grout rests on approximately 4 inches of zero compressive strength of PSW concrete. In addition, it interconnects the Westinghouse short columns to the Gilbert PSW/corbel embedded steel. The audit referenced material (e.g., CR-00-01234, CGE-CA120-TM-SA-000001) indicates that a portion of the grout that has been exposed to boric acid, may also be exposed to effects of radiation at 72 EFPY as well. Response to Audit Breakout Question Number 6 states that the GAI-AOR indicates the grout allowable compressive strength to be 6,453 psi at 72 EFPY versus that of 6,333 psi required at 14 days. A further review of the GAI-AOR, however, shows the demand for the grout compressive strength to be as high as 7,104 psi at 14 days. In addition, literature (NUREG/CR 7280) indicates that creep could also potentially affect the grout deformation and hence stability of the RV placement. Given that it is not clear what would be the compressive strength/load bearing capacity and state of the grout at 72 EFPY, monitoring and management of applicable aging effects (e.g., loss of mechanical properties, reduction of strength, deformation, cracking) for all applicable environments (including radiation) consistent with the GALL-SLR and SRP-SLR Reports review principles is considered essential. Clarify steps that either are or will be taken to make sure that the grout fulfills its intended function and that stability of the RV under all loading combinations, conditions and environments remain consistent with the current licensing basis to the end of the subsequent period of extended operation.
80 i)
Locally the PSW concrete was exposed to boric acid. It is not clear whether the reinforced concrete structural steel and the embedded structural shapes and plates are/were affected by the boric acid. Hence monitoring of the PSW reinforced concrete is considered essential. Clarify steps are/to be taken to make sure that the effects of aging (e.g., loss of material, cracking, loss of strength) to all applicable environments (including that of radiation) for PSW reinforced concrete and embedded structural shapes and plates are managed consistent with the GALL-SLR and SRP-SLR Reports review principles to the end of the subsequent period of extended operation.
j)
Potential loss of grout support function/compressive strength at approximate elevation 426 feet 8 inches also may result to liner overload and reduction to its barrier/retaining wall capacity to further contain portions of the grout and concrete that are considered ineffective to support the Westinghouse short columns in their entirety at 72 EFPY.
Potential liner bulging or deformation could affect the RV short column leveling (as each of these is tangent to the liner), RV balance, and potentially the Nuclear Steam Supply System (NSSS) configuration at 72 EFPY. Hence monitoring of the liner for potential deformation is considered essential. Clarify steps are/to be taken to make sure the effects of aging (e.g., liner deformation, bulging) for the PSW liner, grout, and reinforced concrete are managed consistent with the GALL-SLR and SRP-SLR Reports review principles to the end of the subsequent period of extended operation.
k) Attachment 14 to Calcs DC0020-209, Rev. 0, states that the PSW concrete coarse aggregates are siliceous. Several publications (e.g., see references in Section 3.5.2.2.2.6 of Point Beach SLRA SE - ML22140A127) show that such aggregates when irradiated could produce as much as a 20 percent radiation induced volumetric expansion (RIVE). As a result, at the fuel mid-plane elevation potentially the expanding concrete/aggregates and irradiated strained/cracked liner, at 72 EFPY could affect the stability of the RV. Hence monitoring of the liner for deformation and cracking is considered essential. Clarify steps are/to be taken to make sure that the effects of aging (e.g., loss of material, cracking, RIVE) to all applicable environments (including that of radiation) for the PSW liner and reinforced concrete are managed consistent with the GALL-SLR and SRP-SLR Reports review principles during the subsequent period of extended operation.
l)
The applicant defined radiation screening rules in proprietary W-AOR, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment, Rev. 1, appear to reduce the effects of radiation/fluence on the RV support components (steel, cementitious materials) at 72 EFPY. State where these screening rules were applied and confirm that their use did not materially affect results of calculations described in SLRA Section 3.5.2.2.2.6 including those in demand-to-capacity ratio tables.
m) GAI-AOR states that the embedded steel assembly design did not consider any composite action of the surrounding concrete, especially in calculating stresses in the WF sections. This is conservative, hence the overall adequacy of the anchor assembly to transfer the Westinghouse RV loads to the main concrete of the PSW is satisfactory.
However, it does say that the headed concrete anchors engage the concrete. Clarify the two statements.
n) GAI-AOR also states the following loads for each WF anchor support: Vertical Seismic Design-Basis Event (DBE) = 700 kips, Vertical Accident = 2,950 kips, Horizontal Shear Seismic DBE = 500 kips, Horizontal Shear Accident = 3,000 kips. Discuss what accident load (e.g., large break LOCA) was analyzed in the GAI-AOR. (Can be discussed under the GAI-AOR/W-AOR clarification question.)
81 o) SLRA 3.5.2.2.2.6 states: the primary vertical, hoop, and shear reinforcement which is considered in the analysis of record is located outside the zones of neutron and gamma threshold exceedances. Therefore, there is no reduction in the ability of the primary reinforcement to carry load, and no reduction in PSW capacity as a result of rebar capacity reduction. Details of the capacity reduction calculations are in the proprietary W-AOR.
Clarify whether reinforcement in the capacity reduction calculations in the W-AOR includes only the ((
)) (This one can also be discussed under the GAI-AOR/W-AOR clarification question.)
p) W-AOR states: ((
.)) It is not clear in the response to question 17a of the breakout audit (ML24085A701) whether the 3/4 x 73/16 inch headed concrete anchors are the same as the
(( )) described in the W-AOR. The response to question 17a further states that the headed concrete anchors engage the concrete, but composite action is not credited in the design. This response seems to be inconsistent with the W-AOR. Clarify/discuss this inconsistency.
q) SLRA Table 4.3.4-1 reports cumulative usage factors due to cyclic fatigue loads for the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. The RV outlet and inlet nozzles are supported by the support shoes of the RV steel supports.
Therefore, it is not clear whether the support shoes experience cyclic fatigue loads.
Proprietary document CGE-MC000-TR-CF-000005 provides some details of the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. CR-15-01577 shows a block of metal referred as Weld Build-Up between the RV nozzle and the support shoe and states that the Weld Build-Up is part of the nozzle and is therefore considered part of the pressure boundary. Clarify whether this Weld Build-Up is the support pad.
r) The bases for the 20 percent and 25 percent uncertainties applied to the fluence estimates to the PSW and RV support structure, respectively, are not documented in the application or in a document that can be cited in the NRC staffs safety evaluation.
General Observations:
a) Using design drawings, SLRA Figure 3.5.2.2.2.6-1, figures in the VCSNS UFSAR, the applicant clarified the load path sequence from the RV to the PSW concrete as follows:
reactor vessel nozzle, support shoe, support box plates, shear bars underneath the support box plates, grout, corresponding shear bars of the embedded steel, WF embedded steel and welded plates, PSW concrete.
b) With regard to inspection of components under observation (a), the applicant verified the CR-15-05177 condition report and it is contained within EIR 81987, Rev. C, owner-defined jurisdictional inservice inspection boundaries during the subsequent period of extended operation to remain as follows:
o The SLRA described RV support structure (short-columns) is inspected by ASME Section XI Subsection IWF.
o The SLRA described PSW, grout, and embedded steel including rebars within the PSW are inspected by the structures monitoring program.
82 However, the applicant did not assign any Table 2 AMR line items for the effects of aging for potential reduction of strength and mechanical properties of concrete including grout (cementitious materials) and loss of fracture toughness for steel components exposed to irradiation.
c) Dominion supplemental information for breakout question 8 includes CGE-REAC-TM-AA-000005, Rev. 1, Figure 4, which is the E >0.1 MeV fluence map of the PSW concrete at core midplane elevation. It is derived based on WCAP-18124-NP-A and WCAP-18124-1NP-A Rev. 0 supplement 1-P A methodology. It states and graphically demonstrates that all PSW concrete near the anchor bolts will experience fast neutron fluence (E >0.1 MeV) exposure that are less than the SRP-SLR limit of 1.0E+19at 72 FPY. It also states that these conclusions would not change if the neutron exposure used to generate Figure 4 were adjusted to account for an estimated analytical uncertainty of 20 percent.
d) The Westinghouse AOR (W-AOR), CGE-CA120-CN-SA-000001, states: ((
.)) It is not clear in the response to question 17a of the breakout audit whether the 3/4 x 73/16 inch headed concrete anchors are the same as the (( )) described in W-AOR. The response to question 17a further states that the headed concrete anchors engage the concrete, but composite action is not credited in the design.
e) SLRA 3.5.2.2.2.6 states: the primary vertical, hoop, and shear reinforcement which is considered in the analysis of record is located outside the zones of neutron and gamma threshold exceedances, and that therefore, there is no reduction in the ability of the primary reinforcement to carry load, and no reduction in PSW capacity as a result of rebar capacity reduction. Details of the capacity reduction calculations are in the proprietary W-AOR.
However, it is not clear from the description of the capacity reduction calculations whether reinforcement includes only the ((
.))
f) The GAI-AOR, Structural Engineering Calculation No. 1.53, states that the embedded steel assembly design did not consider any composite action of the surrounding concrete, especially in calculating stresses in the WF sections. This is conservative, hence the overall adequacy of the anchor assembly to transfer the Westinghouse RV loads to the main concrete of the PSW is satisfactory. This calculation also shows the following loads for each WF anchor support: Vertical Seismic DBE = 700 kips, Vertical Accident = 2,950 kips, Horizontal Shear Seismic DBE = 500 kips, Horizontal Shear Accident = 3,000 kips. The calculation did not state specifically what accident load, e.g., large break LOCA, was analyzed.
Fluence Observations:
a) The actual plant-specific geometry of VC Summer Unit 1 was modeled in the Westinghouse-developed RAPTOR-M3G (WCAP-18124-P-A) method to calculate the fluence reported to the various components. The plant-specific model is as shown in Figures 1-3 of proprietary document CGE-REAC-TM-AA-000005, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Bioshield Concrete Neutron Exposure Evaluation.
83 b) Westinghouse explained how the 20 percent and 25 percent uncertainties were determined for the fluence to the PSW and the RV support structure.
c) The 20% uncertainty for the PSW concrete is based on the fluence estimation at 30 centimeters, or approximately 12 inches, above the top of the active fuel height on the RV using WCAP-18124 Supplement 1-P-A methods.
d) The 25 percent uncertainty number for the RV steel support structure (i.e., support box plate, support box, support shoe, and support box plate bolts) is based on fluence estimation at 90 centimeters or approximately 36 inches above the top of fuel of the RV using WCAP-18124 Supplement 1-P-A methods even though all parts of the support steel are within 2 feet of the top of the active fuel height.
e) The applicant stated that the grout sees a maximum E >0.1 MeV fluence of 2.18E19 neutrons per square centimeter (n/cm2) (includes a 20% uncertainty), which is greater than the 1E19 n/cm2 SRP-SLR defined threshold. The NRC staff asked about the use of the 20% uncertainty rather than the 25% uncertainty that was used for the support shoe since the support shoe rests on the grout and the support box structure. The applicant responded that the bottom of the grout (the part that would have the peak fluence) is within ~1 foot (30 cm) from the top of the active fuel height.
Cementitious Support Structure Materials (concrete, grout) Observations:
a) VCSNS SP-201, Structural Concrete - Safety Related states that structural concrete shall have a minimum ultimate compressive strength of 3,000 psi at 28 days or 5,000 PSI at 90 days. It also states that the concrete for the reactor building shell, including its foundation mat, cylindrical walls, dome ring girder, and interior concrete, shall have a minimum ultimate compressive strength of 5,000 psi at 90 days except were a minimum ultimate compressive strength of 5,000 psi at 28 days is specifically designated on the drawings.
b) The W-AOR indicates the original design code of record is ACI 318-71. It discusses elimination of large LOCA from the design basis which reduces demands on PSW when compared with the original GAI PSW AOR. It also states that the PSW is analyzed using 3,000 psi concrete and grade 60 steel reinforcement. Design loading of PSW results to bending moment, axial compression, and torsion. Of these, its states that bending with compression at the inner surface of the PSW would be expected to be most severely impacted by discrediting of concrete at the inner surface, due to high dependence of the moment capacity on the moment arm and the compression block. It also states that use of bending moment capacity is the appropriate metric for bounding to quantify the reduction in capacity because the region of interest is the same as the location where the RV support loads are applied. It then states that consideration of a bending moment is representative of the actual dominant loading condition of the PSW in that region.
Compressive load is present but does not significantly affect behavior in the region.
c) The SLRA Table 3.5.2.2.2.6-2 summary of demand-to-capacity loads are not revised to include the reduction in LBB loads. The table is a modification of the VCSNS PSW Summary of Reinforcement from GAI PSW AOR for steel reinforcement provided versus that was calculated for three finite elements from elevation 417 feet 3 inches to elevation 426 feet 2 inches as representatives of the loading and structural capacity of the PSW.
d) The grout supports the Westinghouse-designed and SLRA-described RV support structure (short-columns). It links the short columns to the PSW and the underlying embedded steel structure (see DWGs in NUREG 7280). Hence the irradiated grout provides structural support. It transfers the current licensing basis defined (with the exception those induced
84 by large break LOCAs) vertical, radial, and tangential loads to the PSW concrete and the underlying embedded steel structure.
e) GAI-AOR for grout bearing calculations states the:
loads from Westinghouse Load Table at the support shoe are as follows: vertical load (Fy): 4400 kips, radial load (Fx): 500 kips, tangential load (Fz): 2320 kips required minimum design grout bearing strength for RV supports to be 7,104 psi at 14 days with expected [unirradiated] grout strength of 7,000 to 7,500 psi at 28 days (GAI-AOR, page 71).
f) The applicant reports that for E >0.1 MeV the calculated maximum grout neutron fluence augmented with an analytical uncertainty of 20 percent is (( )) n/cm2, which exceeds the SRP-SLR for E >0.1 MeV limit of 1.0E19 for cementitious materials. From Figure 4, of CGE-CA120-TM-SA-000001 and DWG E-511-224, it appears that the fluence exceeds the SRP-SLR limit of 1.0E19 n/cm2 at the top of the RV support embedded WF sections/bottom of grout potentially affecting the bearing capacity of the grout for about 2 inches inside the reactor cavity for RV support structure (short-columns). The applicant qualified the compressive strength of the Masterflow 713 grout to be 6,453 psi at EFPY 72 based on a 14-day Pittsburgh Laboratory Tests compressive strength tests of 7,870 psi, followed by an 18 percent reduction in strength due to radiation following the guidance of EPRI Report 3002018400 extracted from Maruyama et al. studies (JACT, volume 15, Figure 54).
g) GAI-AOR states that:
Vertical loads are taken by concrete bearing. Vertical loads are as follows: dead load:
350 kips, thermal load: 185 kips, seismic DBE: 700 kips, accident: 2,950 kips.
For elevated temperature effect strengths are reduced by 5 percent or by a factor of
.95.
h) The applicant defined radiation screening rules in proprietary W-AOR, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment, Rev. 1, to reduce the effects of 72 EFPY fluence on the RV components described under observation (a) above. Rule 2 was considered for neutron fluence only in demand-to-capacity ratios in SLRA Table 3.5.2.2.2.6-2 associated with the PSW concrete but not in the displacements per iron atom assessments in Table 3.5.2.2.2.6-3 for the RV steel support structure. However, the sensitivity analysis performed by the applicant indicates that the effects of using Rule 2 to fluence reduction would be limited. The increase in the radius of neutron and RIVE affected concrete is estimated to be about 10 percent.
i)
The applicant calculated neutron fluence for E >0.1 MeV at 4 inches at any discrete point inside the PSW concrete and adjusted for analytical uncertainty to be below the SRP-SLR limit of 1.0E19 n/cm2 is based on implementation of screening/smoothing rules 1 and 2. In the absence of smoothing, the W-AOR reports that the WCAP-18124-NP-A calculated fluence at any 4-inch location inside the PSW concrete to be (( )) n/cm2, which exceeds the SRP-SLR fluence limit for reduction of strength and mechanical properties of concrete due to irradiation.
j)
Section 4.17.4.6 of DBD RB, Rev. 9, states that the air gap between the insulated RV and the PSW lined concrete is 3 inches. However, calculation DC07020-002, Rev. 3, states that the gap between the PSW liner and RV insulation is 1.5 inches to 1.75 inches with the gap measured in cold condition (confirmed in 1MS-15-163) but expected to be smaller
85 when the RV is at temperature. The calculations also state that the diameter to the PSW liner is 187.5 inches per DWG 411-251 and the diameter to RV insulation is 184.5 inches per 1MS-07-159.
k) Calculation DC07020-002, Rev. 3, states that based on actual field measurements of operating fans amperage the flow rate (air flow velocity) with the seal ring removed (no blockage) is 2033 ft/min or 34 ft/sec with 45,000 cfm total PSW air cavity flow, with 41,880 cfm through the seal ring area, 240 cfm through the neutron detector wells, and 2880 cfm through the primary shield penetrations flow path.
l)
VCSNS also points out to UFSAR Sections 9.4.8.2.1, Reactor Building Cooling System, and 6.2.2.1.2, Reactor Building Cooling System Design Basis for Normal Operation, for cooling capacity based on the installed air handling units during normal operation and loss of offsite power. In 2015, VCSNS initiated replacement of the evaporative industrial cooling towers with mechanical chillers to increase the heat removal capacity of reactor building cooling units by lowering the entering water temperature of the industrial cooling supply to the cooling units.
m) Calculation DC000020-209, Revisions 0 and 1, examine the following three thermal scenarios associated with the reactor air cavity concrete:
The maximum steady state and transient temperatures in the reactor cavity concrete heating with removal of the refueling ring and subsequent change in air cavity distributions to range between 188° to 192°F, based on a 45,000-cfm cavity air flow with a range of 2,880 to 2,250 cfm air flows for penetrations and the assumption of maintaining 150° F at PSW liner concrete interface. May through August thermocouple measurements indicate that the maximum measured concrete temperature asymptotically approaches 115°F, resulting in a 77°F conservatism.
1996 loss of cavity cooling (loss of fans) event to be 243.4°F for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> based on a natural circulation of air cavity flow of 3,500 cfm and 228°F for 90 days Refueling outage 12 discovery of 0.15 gpm mass release from the A hot leg crack for 18 months leading to a total air cavity flow blockage n) Assumptions made for the above scenarios in calculation DC000020-209, Revisions 0 resulting in the noted air cavity conditions are as follows:
((
.))
((
.))
((
.))
((
.))
((
.))
86 o) Regarding the boron leakage calculation that assumed total annulus blockage. On December 9 and 19, 2000, the site stated that:
PSW annulus blockage due to boron was 80 percent to 20 percent per side, or approximately a net blockage of 50 percent extending from the PSW elevation 426 feet 8 inches to 411 feet 10 inches.
The extent of leakage below the PSW elevation of 426 feet 8 inches was unknown.
After examining the condition of the boron deposited on the horizontal surfaces in the air tubes, it was concluded that the liquid carrying the boron was subcooled and the resulting deposition was very flat (indicating that the boron was deposited by evaporation) and there were no hills of piles of boron on the horizontal surfaces (from liquid flashing to steam). There were, however, boron deposits on the vertical surfaces that appeared to be from liquid running down from inside the sandbox through joints in the steel structure.
p) Attachment 14 to Calculation DC000020-209, Revisions 0, dated January 5, 2001, examines effects of high temperatures on comparable concretes and associated conditions with those experienced at VCSNS during the 18 months of boron leakage.
The attachment examines the state of the 5,000-psi reactor support concrete (RSC) compressive strength exposed to 250°F for 90 days. It states that petrographic analysis of the VCSNS concrete contains Type II Portland cement and granite coarse aggregate of high quartz content and thus a high siliceous material content. It also references Portland Cement Association tests of concrete specimens with siliceous aggregates that maintained their original strengths when tested hot to about 800°F. It then examines the moisture retention of the RSC due to the liner on the reactor side and the sealed by coating on the steam generator side and critically reviews applicable data referenced by American Concrete Institute in SP-25 Temperature and Concrete. The report concludes that the worst-case scenario would be a 25 percent loss of concrete compressive strength heated at 250°F for 90 days. Given that the 90-day tested concrete average compressive strength was 6,759 psi and theoretically further increased based on Portland Cement Association and Bureau of Reclamation studies by 20 percent to 8,100 psi, the net strength of the current (circa 2001) RSC concrete is 6,083 psi. The report also discusses the AOR Code ACI 318 statistical approach resulting in an average useable present day (circa 2001) concrete compressive strength of 5,381 psi but selects the 6,083 psi strength to be the governing compressive strength of concrete as the 5,381 psi strength resulted from unsealed tested specimens.
Steel/Metallic Supports not Embedded in the PSW Concrete Observations:
a) SLRA 3.5.2.2.2.6 describes the RV steel supports as consisting of the support box plates, support shoe, anchor bolts, and hold down/guide pins. The proprietary design specification for the RV steel supports includes requirements for:
((.))
((.))
((
.))
b) WCAP-18785-NP/P, Rev. 2, contains details of the fracture mechanics evaluation for the RV steel supports, as follows:
87 The components of the RV steel supports that were evaluated (i.e., the support box plates, support shoe, anchor bolts, and hold down/guide pins) represent the locations of highest stresses and/or located near the active core and subjected to high neutron irradiation.
Effects of embrittlement on the fracture toughness of the components that make up the RV steel supports (i.e., support shoe, support box plates, anchor bolts, hold down/guide pins) have been conservatively included. Either a conservative lower bound fracture toughness value appropriate for the plant-specific material of the RV steel support component was used or the embrittlement shift in nil ductility transition temperature was determined based on the upper bound curve of Figure 3-1 of NUREG-1509.
Effect of strain rate on fracture toughness has been included in the evaluation.
Appropriate design load combinations have been evaluated.
Weld residual stress was included in the evaluation for welded components.
Stresses were applied in a conservative manner when computing applied stress intensity factors.
Applied stress intensity factors were based on appropriate postulated flaw models.
Tables 7-1 through 7-4 show that critical stresses are larger than actual stresses for the support box plates and support shoe.
Table 8-2 shows that critical flaw lengths are larger than ASME Code,Section XI allowable flaw lengths for the anchor bolts and hold down/guide pins.
For the support shoe, Table 7-4 summarizes the comparison of the critical stress values versus the actual stress values for 42 EFPY and 72 EFPY.
For the hold down/guide pins, Table 7-6 summarizes the critical flaw lengths determined for 42 EFPY and 72 EFPY.
c) The response to question 18 of the breakout audit states that the shim plates are fabricated from tool steel (i.e., hardened steel) and are intended to provide consistent frictional resistance and allow nozzle pad free movement during radial thermal growth. As they are hardened steel, the response states that the steel will become harder when exposed to radiation and improve resistance to galling.
d) SLRA Table 4.3.4-1 reports cumulative usage factors due to cyclic fatigue loads for the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. The RV outlet and inlet nozzles are supported by the support shoes of the RV steel supports.
Therefore, it is not clear whether the support shoes experience cyclic fatigue loads.
Proprietary document CGE-MC000-TR-CF-000005 provides some details of the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. CR-15-01577 shows a block of metal referred as Weld Build-Up between the RV nozzle and the support shoe and states that the Weld Build-Up is part of the nozzle and is therefore considered part of the pressure boundary. However, it is not clear whether this Weld Build-Up is the support pad.
e) Condition report CR-15-05177 included summary data sheets of the ASME Code,Section XI, IWF, inspections performed in 2015. The summary data sheets showed acceptable results per the IWF-3410 acceptance criteria for the RV steel supports.
88 f) Condition report CR-00-01324 documented the boric acid leak in Hot Leg Loop A in 2000.
The condition report states in part, The reactor vessel support system and the annulus between the PSW and the reactor vessel have been inspected with a borescope or other remote device and have been cleanedNo significant corrosion of the steel portions of the RV support system was noted.
SLRA AMR 3.6.2.2.2, Reduced Insulation Resistance Due to Age Degradation of Cable Bus Arrangements Caused by Intrusion of Moisture, Dust, Industrial Pollution, Rain, Ice, Photolysis, Ohmic Heating and Loss of Strength of Support Structures and Louvers of Cable Bus Arrangements Due to General Corrosion and Exposure to Air Outdoor Summary of Information in the Application. During the audit, the staff reviewed plant documentation associated with the following:
SLRA Table 3.6.1, Summary of Aging Management Programs for the Electrical Components Evaluated in Chapter VI of the GALL-SLR Report, Item Numbers 3.6.1-029, Cable bus: electrical insulation; insulators - exposed to air-indoor controlled or uncontrolled, air-outdoor, 3.6.1-030, Cable bus: external surface of enclosure assemblies composed of steel exposed to air - indoor uncontrolled or air-outdoor, and 3.6.1-031, Cable bus external surface of enclosure assemblies composed of galvanized steel; aluminum exposed to air-outdoor Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant. The staff will document its review of this information in the SE.
Document Title Revision/Date VCSNS - DBD South Carolina Electric & Gas Company - Virgil C. Summer Nuclear Station - Design Basis Document - Cable and Raceway (CB)
Rev. 2 ETE-SLR-2023-3343 V.C. Summer Subsequent License Renewal Project - Aging Management Program Evaluation Report - Electrical Cables and Connections not Subject to 10 CFR 50.49 Environmental Qualification Requirements Rev. 0 ETE-SLR-2023-3106 V.C. Summer Subsequent License Renewal Project -
Integrated Plant Assessment Report - Electrical Components Rev. 0 ETE-SLR-2023-3606 V.C. Summer Subsequent License Renewal Project -
Further Evaluation Recommended Report - Electrical and Instrumentation and Controls Rev. 0 ETE-SLR-2023-3005 V.C. Summer Subsequent License Renewal Project -
Regulated Event Report - 10 CFR 54.4(a)(3) Loss of All Alternating Current Power Rev. 0 VCSNS SLR LRA Section 2.1.3.4 Station Blackout - page 2-11 Correction 01/08/2024 The NRC staff conducted an onsite audit and discussed maintenance practices associated with cable buses within scope of the SLR. The staff observed that the installed cable bus system appeared to be well maintained.
SLRA AMR 3.6.2.2.3, Loss of Material Due to Wind-Induced Abrasion, Loss of Conductor Strength Due to Corrosion, and Increased Resistance of Connection Due to Oxidation or Loss of Preload for Transmission Conductors, Switchyard Bus, and Connections
89 Summary of Information in the Application. During the audit, the staff reviewed plant documentation associated with the following:
SLRA Table 3.6.1, Summary of Aging Management Programs for the Electrical Components Evaluated in Chapter VI of the GALL-SLR Report, Item Numbers 3.6.1-004, Transmission conductors composed of aluminum; steel exposed to air-outdoor 3.6.1-005, Transmission connectors composed of aluminum; steel exposed to air-outdoor, 3.6.1-006, Switchyard bus and connections composed of aluminum; copper; bronze; stainless steel; galvanized steel exposed to air-outdoor, and 3.6.1-007, Transmission conductors composed of aluminum; steel exposed to air-outdoor Audit Activities. An onsite audit was performed December 5-7, 2023.
The table below lists documents that were reviewed by the staff and were found relevant. The staff will document its review of this information in the SE.
Document Title Revision/Date ETE-SLR-2023-3606 V.C. Summer Subsequent License Renewal Project -
Further Evaluation Recommended Report - Electrical and Instrumentation and Controls Rev. 0 ETE-SLR-2023-3106 V.C. Summer Subsequent License Renewal Project -
Integrated Plant Assessment Report - Electrical Components Rev. 0 ETE-SLR-2023-3005 V.C. Summer Subsequent License Renewal Project -
Regulated Event Report - 10 CFR 54.4(a)(3) Loss of All Alternating Current Power Rev. 0 ETE-SLR-2023-3115 V.C. Summer Subsequent License Renewal Project -
Integrated Plant Assessment Report - Yard Structures Rev. 0 VCSNS SLR LRA Section 2.1.3.4 Station Blackout - page 2-11 Correction 01/08/2024 The staff conducted an onsite audit and performed walkdowns of the switchyard and the transmission conductors associated with the station blackout recovery path within scope of the SLR.
The staff also discussed maintenance practices related to transmission conductors and the switchyard. The staff observed that the switchyard and the transmission conductors appeared to be well maintained.
- 3. Supplements to the SLRA By letters dated April 2 (ML24095A207) and May 6, 2024 (ML24129A200), DESC voluntarily submitted Supplements 1 and 2 to the SLRA resulting from discussions held during the audit.
- 4. Audit Questions Provided to DESC Over the course of the audit, the NRC staff provided audit questions to DESC to facilitate the audit discussions (ML24085A701 and ML24085A700), for the non-proprietary and proprietary versions, respectively.
Subject Area Meeting Date Breakout Session: Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components 12/20/2023 Breakout Session: Selective Leaching 01/08/2024
90 Subject Area Meeting Date Breakout Session: Buried and Underground Piping and Tanks 01/09/2024 Breakout Session: ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD 01/09/2024 Breakout Session: PWR Vessel Internals 01/10/2024 01/24/2024 Breakout Session: Cumulative Fatigue Damage, Metal Fatigue, Transient Cycle Projections for 80 years, ASME Code Section III Class 1 Fatigue Analyses 01/10/2024 Breakout Session: Scoping and Screening Section 2.3.3.13 01/11/2024 Breakout Session: Non-Class 1 Allowable Stress Analyses, Environmentally-Assisted Fatigue, High-Energy Line Break Analysis 01/16/2024 Breakout Session: Water Chemistry 01/16/2024 Breakout Session: Steam Generators 01/17/2024 Breakout Session: ASME Code Class 1 Small-Bore Piping 01/18/2024 Breakout Session: Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components 01/18/2024 Breakout Session: High-Energy Line Break Analysis, Fatigue Monitoring 01/18/2024 Breakout Session: Reactor Head Closure Studs Bolting, Flux Thimble Tube Inspection 01/22/2024 Breakout Session: Leak-Before-Break 01/23/2024 Breakout Session: Bolting Integrity, Structures Monitoring, Inspection of Water-Control Structures Associated with Nuclear Power Plants 01/24/2024 Breakout Session: Inaccessible Medium Voltage Cables, Inaccessible Instrument and Control Cables, Inaccessible Low Voltage Cables and Fuse Holders 01/25/2024 Breakout Session: Reactor Vessel Neutron Embrittlement Analysis 02/01/2024 Breakout Session: Fire Protection 02/01/2024 Breakout Session: Fire Water System 02/05/2024 03/12/2024 Breakout Session: ASME Section XI, Subsection IWE 02/14/2024 Breakout Session: ASME Section XI, Subsection IWF 02/14/2024 Breakout Session: ASME Section XI, Subsection IWL, Concrete Containment Unbonded Tendon Prestress 02/15/2024 Breakout Session: Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation 02/20/2024 02/22/2024 02/26/2024 02/27/2024 02/28/2024 02/29/2024
- 5. Applicant Personnel Contacted During the Audit Name Affiliation Alexus Hancock Dominion Energy Allen I Hiser Dominion Energy Charles A Tomas Dominion Energy Craig H Heah Dominion Energy Damon Bryson Dominion Energy Daniel P Madden Dominion Energy David T Clohecy Dominion Energy Garrett S Haltiwanger Dominion Energy Gregory J Travers Dominion Energy Jeffrey Lackovic Dominion Energy John M Thomas Dominion Energy Mary E Kowalski Dominion Energy Michael W Guthrie Dominion Energy
91 Name Affiliation Pratt Cherry Dominion Energy Richard Eagan Dominion Energy Rod Steffy Dominion Energy Roger Rucker Dominion Energy Ronald Kiehl Dominion Energy Shawn Sweigart Dominion Energy Stephen Ariail Dominion Energy Tom Snow Dominion Energy Andrew Heffner Westinghouse Andy Hawk Westinghouse Anees Udyawar Westinghouse Brian J Hall Westinghouse Daniel Carlin Westinghouse Dulal Bhowmick Westinghouse Hall Gordon Westinghouse Hamilton Hsiao Westinghouse Maria Rizzilli Westinghouse Nick DeDominicis Westinghouse Stephen Longwell Westinghouse Tom Gerlowski Westinghouse Xi Liu Westinghouse