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Category:Letter
MONTHYEARML24038A0322024-02-0707 February 2024 (Vcsns), Unit 1, 10 CFR 21 Report - Emergency Diesel Generator Flash Contactor Relay IR 05000395/20230042024-01-31031 January 2024 Integrated Inspection Report 05000395/2023004 ML24022A1582024-01-22022 January 2024 (Vcsns), Unit 1 - Reply to a Notice of Violation: EA-23-093 ML24022A1202024-01-22022 January 2024 Notification of Target Set Inspection and Request for Information (NRC Inspection Report 05000395/2024403) ML24018A0112024-01-18018 January 2024 CFR 21 Report Emergency Diesel Generator Fuel Oil Piping IR 05000395/20240112024-01-17017 January 2024 Notification of VC Summer Station Comprehensive Engineering Team Inspection - U.S. Nuclear Regulatory Commission Inspection Report 05000395/2024011 ML23361A1042023-12-21021 December 2023 Inservice Testing Program for Pumps and Valves Fifth Interval Update and Associated Relief and Alternative Requests IR 05000395/20230912023-12-21021 December 2023 Final Significance Determination of a White Finding and Notice of Violation and Assessment Followup Letternrc Inspection Report 05000395/2023091 ML23354A1802023-12-20020 December 2023 301 Examination Report with Cover Letter ML23347A0492023-12-12012 December 2023 Technical Specification Bases Changes Updated Through November 2023 ML23334A2462023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23334A2372023-11-30030 November 2023 (Vvsns), Unit 1 - Annual Commitment Change Summary Report ML23332A1942023-11-28028 November 2023 (Vcsns), Unit 1, Correction of Typographical Error in License Amendment Request - Inverter Allowed Outage Time (AOT) Extension ML23317A2242023-11-0909 November 2023 (Vcsns), Unit 1 - License Amendment Request - Inverter Allowed Outage Time (AOT) Extension ML23289A1172023-11-0303 November 2023 Tribal Request for Scoping Comments Concerning the Environmental Review of the V.C. Summer SLRA ML23289A1152023-11-0303 November 2023 Ltr to R. Nelson - Request for Scoping Comments Concerning the Environmental Review of the V.C. Summer SLRA ML23289A1162023-11-0303 November 2023 Ltr to W. Emerson - Request for Scoping Comments Concerning the Environmental Review of the V.C. Summer SLRA IR 05000395/20234032023-11-0101 November 2023 Material Control and Accounting Program Inspection Report 05000395/2023403 Cover IR 05000395/20230032023-11-0101 November 2023 Integrated Inspection Report 05000395/2023003 ML23284A1792023-10-16016 October 2023 Subsequent License Renewal Application Online Reference Portal ML23275A0142023-10-11011 October 2023 SLRA - Acceptance and Opportunity for Hearing - Letter ML23284A1662023-10-10010 October 2023 (Vcsns), Unit 1 - Request for Regulatory Conference; (EA-23-093) IR 05000395/20230902023-10-0404 October 2023 NRC Inspection Report 05000395/2023090 and Preliminary Yellow Finding and Apparent Violation IR 05000395/20230112023-09-27027 September 2023 Focused Engineering Inspection Report 05000395/2023011 ML23250A3112023-09-20020 September 2023 Request to Revise Reactor Vessel Material Surveillance Capsule Withdrawal Schedules ML23244A2352023-09-18018 September 2023 Request to Use a Subsequent Edition of the ASME Code, Section XI ML23230A0012023-09-15015 September 2023 Request to Use a Provision of a Previous Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for Non-Destructive Examinations ML23235A0432023-09-0505 September 2023 Receipt and Availability of the Subsequent License Renewal Application IR 05000395/20234012023-08-30030 August 2023 Security Baseline Inspection Report 05000395/2023401 ML23242A1192023-08-29029 August 2023 (Vcsns), Unit 1 - Inservice Inspection (ISI) Owner'S Activity Report (OAR) for Refueling Outage 27 ML23243B0362023-08-25025 August 2023 Request for FEMA Consultation on the Dominion Energy Proposal to Consolidate the Emergency Operations Facility for the Virgil C. Summer Nuclear Station Unit 1 IR 05000395/20230052023-08-23023 August 2023 Updated Inspection Plan for Virgil C. Summer Nuclear Station (Report 05000395 2023005) ML23233A1752023-08-17017 August 2023 Application for Subsequent Renewed Operating License IR 05000395/20230022023-08-14014 August 2023 Integrated Inspection Report - Integrated Inspection Report 05000395/2023002 and Apparent Violation IR 05000395/20230102023-08-12012 August 2023 Biennial Problem Identification and Resolution Inspection Report 05000395/2023010 ML23207A1102023-07-26026 July 2023 NRC Regulatory Issues Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations IR 05000395/20234412023-07-17017 July 2023 95001 Supplemental Inspection Report 05000395/2023441 and Follow Up Assessment Letter ML23190A0012023-07-0909 July 2023 Special Report 2023-003 Inoperable Radiation Monitoring Instrumentation Channel ML23187A5092023-07-0606 July 2023 Request for Federal Emergency Management Agency Consultation on the Dominion Energy Proposal to Consolidate the Emergency Operations Facility for the Virgil C. Summer Nuclear Station Unit 1 ML23178A1682023-06-26026 June 2023 2022 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML23173A1052023-06-22022 June 2023 Special Report 2023-002 Inoperable Loose-Part Monitoring System Channel ML23163A2352023-06-12012 June 2023 Request to Use a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for the SPS Sixth Inservice Inspection (Isi) Intervals and VCSNS Fifth ISI Interval ML23159A2332023-06-0808 June 2023 License Amendment Request - Emergency Response Organization Augmentation Time Change, Emergency Operations Facility Relocation and Other Emergency Plan Changes ML23156A3092023-06-0505 June 2023 Requalification Program Inspection Virgil C. Summer Nuclear Station Unit 1 - 05000395/2023003 2024-02-07
[Table view] Category:Report
MONTHYEARML23325A2022024-01-16016 January 2024 CFR Part 52 Construction Lessons-Learned Report ML23024A1542023-01-23023 January 2023 Proposed Reactor Vessel Surveillance Capsule Withdrawal Schedule to Support Potential Subsequent License Renewal Activity ML22286A1392022-10-13013 October 2022 Special Report 2022-005, Inoperable Radiation Monitoring Instrumentation Channel ML22279A9892022-09-23023 September 2022 Restoration Project - Final Status Survey Release Record North Protected Area Yard Survey Unit 12201C - Revision 2 ML22069B1172022-03-10010 March 2022 Application for Alternative Request - Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination) ML22049B0242022-02-18018 February 2022 2021 Q4 Summary Page IR 05000395/20210052021-08-24024 August 2021 Updated Inspection Plan for Virgil C.Summer Nuclear Station, Unit 1 (Report 05000395/2021005) ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML20296A7082020-10-22022 October 2020 (VCSNS) Unit 1 - Alternative Requests RR-4-25 for Elimination of Reactor Pressure Vessel Threads in Flange Examination for the Remainder of the Fourth 10-Year ISI Interval ML20247J6162020-09-0303 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20212L5762020-07-30030 July 2020 Annual Commitment Change Summary Report ML20203M1602020-07-20020 July 2020 VA Elec. & Power Co., Dominion Energy Nuclear Co. Inc., Dominion Energy Sc Inc., Millstone Power Station 2, N. Anna & Surry Power Stations 1 & 2, Virgil C. Summer Station 1, Updated Anchor Darling Double Disc Gate Valve Information & Status ML19204A1172019-07-17017 July 2019 Vigil C. Summer, Unit 1, Proposed Alternative Request RR-4-20 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML19122A5172019-05-0202 May 2019 Annual Commitment Change Summary Report ML19056A4122019-01-31031 January 2019 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Source Water Baseline Biological Characterization Data ML19056A4132019-01-31031 January 2019 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Cooling Water System Data ML19056A4112019-01-31031 January 2019 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Cooling Water Intake Structure Data ML19056A4142019-01-31031 January 2019 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Entrainment Performance Studies ML19056A4102019-01-31031 January 2019 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Source Water Physical Data RC-18-0117, (Vcsns), Unit 1 - Fukushima Near-Term Task Force Recommendation 3.1: Seismic Probabilistic Risk Assessment2018-09-28028 September 2018 (Vcsns), Unit 1 - Fukushima Near-Term Task Force Recommendation 3.1: Seismic Probabilistic Risk Assessment ML18179A4162018-06-28028 June 2018 ECCS Evaluation Model Revisions Report RC-18-0064, (Vcsns), Unit 1 - Annual Commitment Change Summary Report2018-05-18018 May 2018 (Vcsns), Unit 1 - Annual Commitment Change Summary Report ML17206A4592017-09-26026 September 2017 Staff Assessment of Response to Information Request Pursuant to 10 CFR 50.54(F) - Recommendation 9.3 of the Near-Term Task Force, Communications Assessment RC-17-0089, Focused Evaluation for External Flooding2017-06-30030 June 2017 Focused Evaluation for External Flooding RC-17-0057, Emergency Core Cooling System Evaluation Model Revisions Annual Report2017-05-15015 May 2017 Emergency Core Cooling System Evaluation Model Revisions Annual Report ML19056A4152017-02-28028 February 2017 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Appendix B, Entrainment Study, 2016 and Revised 2017 RC-16-0170, Mitigating Strategies Assessment (MSA) Report Submittal2016-12-22022 December 2016 Mitigating Strategies Assessment (MSA) Report Submittal RC-16-0143, (VCSNS) Unit 1 - Report of Full Compliance and Final Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design...2016-10-31031 October 2016 (VCSNS) Unit 1 - Report of Full Compliance and Final Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design... RC-16-0081, Submittal of 2015 Annual Commitment Change Summary Report2016-06-22022 June 2016 Submittal of 2015 Annual Commitment Change Summary Report RC-16-0086, Special Report 2016-003 Regarding Fire Barrier Not Restored to Operability Status within 7 Days of Inoperability2016-06-0909 June 2016 Special Report 2016-003 Regarding Fire Barrier Not Restored to Operability Status within 7 Days of Inoperability RC-16-0008, Transmittal of Expedited Seismic Evaluation Process Report, Revision 12016-01-28028 January 2016 Transmittal of Expedited Seismic Evaluation Process Report, Revision 1 ML15296A3772015-11-0303 November 2015 Supplement to Staff Assessment of Response to 10 CFR 50.54(f) Information Request- Flood Causing Mechanism Reevaluation RC-15-0153, Submittal of 10 CFR 50.59 Biennial Report2015-10-0707 October 2015 Submittal of 10 CFR 50.59 Biennial Report ML15194A0552015-07-20020 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 Fukushima Dai-Ichi RC-15-0020, Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 22015-02-25025 February 2015 Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 2 ML15061A0332015-02-25025 February 2015 Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 2 ML14314A0332014-11-0505 November 2014 Submittal Special Report (Spr) 2014-006 ML14261A2762014-09-18018 September 2014 Draft September 23, 2014, Category 1 Public Meeting with V. C. Summer - Draft License Amendment Related to Approval of the Technical Support Center ML14141A4612014-06-0606 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident RC-14-0048, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-03-26026 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML14051A3632014-03-0505 March 2014 Closure Letter Concerning 2012 Annual Emergency Core Cooling System Evaluation Model Revisions Report (TAC No. Mf 2722) ML19056A4082014-02-28028 February 2014 Virgil C. Summer Nuclear Station NPDES Permit No. SC0030856 Renewal Application, Thermal Mixing Zone Evaluation, Addendum: Additional Modeling Cases for Revised Reservoir Ambient and Discharge Temperatures ML14034A3392014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14037A2282014-02-21021 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Virgil C. Summer Nuclear Station, Unit 1, TAC MF2338 ML14010A4152014-01-30030 January 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML13309B0532013-11-0404 November 2013 ECCS Evaluation Model Revisions 30-Day Report RC-13-0130, WCAP-17758-NP, Rev. 0, Technical Basis for Westinghouse Embedded Flaw Repair for V.C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles and Attachment Welds.2013-08-31031 August 2013 WCAP-17758-NP, Rev. 0, Technical Basis for Westinghouse Embedded Flaw Repair for V.C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles and Attachment Welds. RC-13-0038, Flooding Hazard Reevaluation Response to NRC Request for Information Pursuant to 10 CFR 50.54(F) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task.2013-03-12012 March 2013 Flooding Hazard Reevaluation Response to NRC Request for Information Pursuant to 10 CFR 50.54(F) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task. RC-13-0037, Permanent ILRT Interval Extension Risk Impact Assessment2013-03-0101 March 2013 Permanent ILRT Interval Extension Risk Impact Assessment ML13072A0832013-01-31031 January 2013 12-4116.IPEC.V002, Revision 1, Overall Integrated Plan for Reliable Spent Fuel Pool Instrumentation: EA-12-051 2024-01-16
[Table view] Category:Miscellaneous
MONTHYEARML22286A1392022-10-13013 October 2022 Special Report 2022-005, Inoperable Radiation Monitoring Instrumentation Channel ML22049B0242022-02-18018 February 2022 2021 Q4 Summary Page IR 05000395/20210052021-08-24024 August 2021 Updated Inspection Plan for Virgil C.Summer Nuclear Station, Unit 1 (Report 05000395/2021005) ML20247J6162020-09-0303 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20212L5762020-07-30030 July 2020 Annual Commitment Change Summary Report ML19122A5172019-05-0202 May 2019 Annual Commitment Change Summary Report ML18179A4162018-06-28028 June 2018 ECCS Evaluation Model Revisions Report RC-18-0064, (Vcsns), Unit 1 - Annual Commitment Change Summary Report2018-05-18018 May 2018 (Vcsns), Unit 1 - Annual Commitment Change Summary Report ML17206A4592017-09-26026 September 2017 Staff Assessment of Response to Information Request Pursuant to 10 CFR 50.54(F) - Recommendation 9.3 of the Near-Term Task Force, Communications Assessment RC-17-0057, Emergency Core Cooling System Evaluation Model Revisions Annual Report2017-05-15015 May 2017 Emergency Core Cooling System Evaluation Model Revisions Annual Report RC-16-0170, Mitigating Strategies Assessment (MSA) Report Submittal2016-12-22022 December 2016 Mitigating Strategies Assessment (MSA) Report Submittal RC-16-0081, Submittal of 2015 Annual Commitment Change Summary Report2016-06-22022 June 2016 Submittal of 2015 Annual Commitment Change Summary Report RC-16-0086, Special Report 2016-003 Regarding Fire Barrier Not Restored to Operability Status within 7 Days of Inoperability2016-06-0909 June 2016 Special Report 2016-003 Regarding Fire Barrier Not Restored to Operability Status within 7 Days of Inoperability RC-16-0008, Transmittal of Expedited Seismic Evaluation Process Report, Revision 12016-01-28028 January 2016 Transmittal of Expedited Seismic Evaluation Process Report, Revision 1 ML15296A3772015-11-0303 November 2015 Supplement to Staff Assessment of Response to 10 CFR 50.54(f) Information Request- Flood Causing Mechanism Reevaluation RC-15-0153, Submittal of 10 CFR 50.59 Biennial Report2015-10-0707 October 2015 Submittal of 10 CFR 50.59 Biennial Report ML15194A0552015-07-20020 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 Fukushima Dai-Ichi RC-15-0020, Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 22015-02-25025 February 2015 Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 2 ML15061A0332015-02-25025 February 2015 Attachment 2 - Ihi Southwest Inner Diameter Examination Data, Part 2 of 2 ML14314A0332014-11-0505 November 2014 Submittal Special Report (Spr) 2014-006 ML14261A2762014-09-18018 September 2014 Draft September 23, 2014, Category 1 Public Meeting with V. C. Summer - Draft License Amendment Related to Approval of the Technical Support Center ML14141A4612014-06-0606 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14051A3632014-03-0505 March 2014 Closure Letter Concerning 2012 Annual Emergency Core Cooling System Evaluation Model Revisions Report (TAC No. Mf 2722) ML14010A4152014-01-30030 January 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML13309B0532013-11-0404 November 2013 ECCS Evaluation Model Revisions 30-Day Report RC-12-0104, Submittal of ECCS Evaluation Model Revisions 30-Day Report2012-10-16016 October 2012 Submittal of ECCS Evaluation Model Revisions 30-Day Report RC-15-0020, Attachment 5 - Westinghouse LTR-PAFM-12-86 Flaw Tolerance Evaluation to Support Re-Categorization of V.C. Summer Unit 1 Steam Generator Nozzle to Safe End Dissimilar Metal Weld Inspection Requirements2012-07-31031 July 2012 Attachment 5 - Westinghouse LTR-PAFM-12-86 Flaw Tolerance Evaluation to Support Re-Categorization of V.C. Summer Unit 1 Steam Generator Nozzle to Safe End Dissimilar Metal Weld Inspection Requirements RC-11-0178, Submittal of Twenty-Second Report to 10 CFR 50.59(d)(2) Changes2011-11-0404 November 2011 Submittal of Twenty-Second Report to 10 CFR 50.59(d)(2) Changes RC-11-0119, 30-Day Special Report (Rt 2800) Groundwater Protection Initiative (GPI) - Voluntary Special Report for On-Site Liquid Effluent Line Leak2011-08-0303 August 2011 30-Day Special Report (Rt 2800) Groundwater Protection Initiative (GPI) - Voluntary Special Report for On-Site Liquid Effluent Line Leak ML1019304172010-05-0606 May 2010 Tritium Database Report ML0921001332009-07-24024 July 2009 Submittal of Special Report (Spr) 09-0001 RC-09-0050, Submittal of 2008 Emergency Core Cooling System (ECCS) Evaluation Model Revisions Report2009-05-14014 May 2009 Submittal of 2008 Emergency Core Cooling System (ECCS) Evaluation Model Revisions Report ML0810701772008-04-11011 April 2008 Submittal of Special Report 2008-001, Pursuant to Requirements of Technical Specification 3.3.3.10.a RC-08-0019, Request for Use of Higher Assigned Protection Factors with Use of French-Designed Air-Line Respirator Equipment2008-02-19019 February 2008 Request for Use of Higher Assigned Protection Factors with Use of French-Designed Air-Line Respirator Equipment RC-08-0006, Application to Use Weighting Factors for External Exposure2008-02-19019 February 2008 Application to Use Weighting Factors for External Exposure RC-07-0169, Virgil Summer - 10 CFR 50.59 Biennial Report Covering the Period from October 1, 2005 Until October 1, 20072007-11-0606 November 2007 Virgil Summer - 10 CFR 50.59 Biennial Report Covering the Period from October 1, 2005 Until October 1, 2007 RC-07-0081, ECCS Evaluation Model Revisions Annual Report2007-06-0101 June 2007 ECCS Evaluation Model Revisions Annual Report RC-06-0216, License Condition 2.C(5), 14-Day Report on Exceeding Surveillance Frequency2006-12-13013 December 2006 License Condition 2.C(5), 14-Day Report on Exceeding Surveillance Frequency RC-06-0205, ECCS Evaluation Model Revisions Report2006-12-0404 December 2006 ECCS Evaluation Model Revisions Report RC-06-0202, V. C. Summer - ECCS Evaluation Model Revisions Report, Addresses the Effect of Using a Finer Break Spectrum2006-11-15015 November 2006 V. C. Summer - ECCS Evaluation Model Revisions Report, Addresses the Effect of Using a Finer Break Spectrum RC-06-0196, Special Report (Spr) 2006-0052006-10-27027 October 2006 Special Report (Spr) 2006-005 RC-05-0076, Operating License, Special Report (Spr 2005-001)2005-05-19019 May 2005 Operating License, Special Report (Spr 2005-001) ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement ML0406307712004-02-25025 February 2004 Transmittal of Semi-Annual Fitness-for-Duty Report for Period Ending December 31, 2003 2022-02-18
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Text
Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 November 5, 2014 A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir / Madam:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 SPECIAL REPORT (SPR) 2014-006
Reference:
Letter from S. A. Byrne (VCSNS) to Document Control Desk (NRC),"SPECIAL REPORT (SPR 2000-005)," dated November 8, 2000[ML003769321]
South Carolina Electric & Gas Company (SCE&G) hereby submits the Steam Generator Tube Inspection Report pursuant to the requirements of Technical Specifications 6.9.1.12.
This report summarizes the examination conducted during the 2014 Spring Refueling (RF-21) and the state of the three steam generators installed at V. C. Summer Nuclear Station (VCSNS), Unit 1.In preparation for the RF-21 inspection, an error was identified.
The referenced letter incorrectly identified the location of two of the three tubes that were plugged in 'A" Steam Generator during RF-12. This error has been corrected in Table 9 of the attached Steam Generator Tube Inspection Report. The error has been entered in the station's corrective action program. No tube plugging was required during RF-21.Should you have any questions, please call Mr. Bruce Thompson at (803) 931-5042.Very truly yours, Thomas D. Gatlin WLT/TDG/ts Attachment c: K. B. Marsh S. A. Byrne J. B. Archie N. S. Carns J. H. Hamilton J. W. Williams W. M. Cherry V. M. McCree S. A. Williams NRC Resident Inspector K. M. Sutton INPO Records Center NSRC RTS (LTD 322)File (818.08)PRSF (RC-14-0171)
Virgil C. Summer Station
- F (803) 941-9776 Document Control Desk Attachment LTD 322 RC-14-0171 Page 1 of 11 Steam Generator Tube Inspection Report VC Summer Nuclear Station Jenkinsville SC 29065 RF-21 April 2014 1.0 Background Commissioned in January 1984, VC Summer originally had three Westinghouse Model D3 steam generators.
The original steam generators were replaced with Westinghouse Delta-75 steam generators in the fall of 1994.Each" of the replacement steam, generators contains 6307 thermally treated alloy 690 tubes with an outside diameter, of 0.6875 inches and a nominal wall thickness of 0.040 inches. Stress relief was performed during fabrication on the U-bends of the first 17 rows of tubing. The straight sections of the tubing are supported by nine tube support plates made of 1.125 inch thick, SA-240, Type 405 stainless steel. The tube support plates (TSP) have trefoil, broached holes which reduce dryout and the collection of impurities at the tube-to-support plate intersections.
The u-bend sections of the tubing are supported by four sets of anti-vibration bars (AVB) made of SA-240, Type 405 stainless steel. A flow distribution baffle (FDB) plate is located between the tubesheet and the lowest support plate. The flow distribution baffle increases the secondary side flow velocity near the tubesheet, thus reducing the accumulation of sludge at the top of the tubesheet.
The D-75 steam generators contain a sludge collector above each primary separator which is designed to reduce the accumulation of sludge at the tubesheet.
The tube-to-tubesheet joints are formed by a full depth hydraulic expansion and a weld to the primary cladding on the tubesheet.
2.0 Scope of Inspections Performed on Each Steam Generator (SG)The RF-21 inspection program, as required by the EPRI PWR SG Examination Guidelines Revision 7, addressed the existing and potential degradation mechanisms for V. C. Summer. The defined scope for each steam generator included the following:
- 1) 55% (plus tubes with prior indications) full length bobbin inspection
- 2) U-bend +Point inspection of tubes that a bobbin probe would not pass through 3) 100% +Point inspection of peripheral tubes and at the tube lane (two tubes closest to the boundary in both cases) in the hot leg (HL) and cold leg (CL) from 3 inches above to 3 inches below Top of the Tubesheet (TTS)4) 20% +Point inspection of non-peripheral tubes in hot leg from 3 inches above to 3 inches below TTS Document Control Desk Attachment LTD 322 RC-14-0171 Page 2 of 11 5) 50% +Point inspection of all dents/dings
> 5 volts 6) +Point inspection of legacy loose parts 7) Special interest RPC (freespan signals without historical resolution, bobbin I-code indications, etc.)8) 100% tube plug video inspection in HL and CL (8 tubes total)9) Video scan of channel head bowl in HL and CL as recommended by Westinghouse Nuclear Safety Advisory Letter (NSAL) 12-1 10)TTS secondary side video inspection including Foreign Object Search & Retrieval (FOSAR)1 1)Upper bundle video inspection in 'A' SG 12)Steam drum inspection.Tables 1 and 2 provide the RF-21 total eddy current examination scope.Table 1: Base Scope. Exam --Base Scope Exam Type SG A SG B SG C Cold Leg Full Length Bobbin 2889 2895 2885 Cold Leg Low Row Straight Section 619 545 322 Hot Leg Full Length Bobbin 0 80 298 Hot Leg Candy Cane 619 545 322 Cold Leg RPC Tubesheet Exams (Periphery and Intejior) 680 679 678 Hot Leg RPC Tubesheet Exams (Periphery and Interior) 1832 1893 1882 TOTAL 6639 6637 6387 Table 2: +Point Diagnostic Exams for Bobbin Indications (Special Interest Additional Scope)Exam Type SG A SG B SG C Hot Leg Bobbin Dent/Ding
> 5V 0 2 0 Hot Leg Bobbin I-Code Indications 6 1 4 Hot Leg PLPNOL and Box In 18 14 0 Hot Leg Mag Bias Special Interest 0 1 2 Cold Leg Bobbin Dent/Ding
> 5V 34 47 31 Cold Leg Bobbin I-Code Indications 5 13 8 Cold Leg PLPNOL and Box In 0 0 0 Cold Leg New Bobbin Wear 10 20 13 Cold Leg Mag Bias Special Interest 0 0 0 U-bend Bobbin I-Code Indications 0 0 2 Hot Leg U-Bend Dent/Ding
> 5V 2 2 0 Cold Leg U-Bend Dent/Ding
> 5V 1 1 0 TOTAL 76 101 60 Document Control Desk Attachment.'
LTD 322 RC-14-0171 Page 3 of 11 3.0 Degradation Mechanisms Found Existing degradation mechanisms are those mechanical or corrosive processes previously and/or currently observed in a Steam Generator.
Based on this definition of existing degradation mechanisms and prior inspection results, there are three existing degradation mechanisms within the VC Summer replacement steam generators.
3.1 Tube Wear at AVBs Wear related to tube interaction with AVB supports results from flow-induced vibration in the upper bundle. This mechanical process is related to the tightness of the upper bundle assembly as expressed in the distribution of tube-to-AVB gaps. Indications of AVB wear in-the replacement steam generators were first reported during RF-12. Very few AVB wear indications were reported in subsequent inspections during RF-1 5 and RF-1 8. No new AVB wear indications were reported in RF-21. The deepest wear indication during RF-21 was 10% through-wall depth (TWD). To date, no tubes have been plugged for this mechanism.
3.2 Tube Wear at TSPs and FDB Flow-induced vibration that causes tube-wear at TSPs and the FDB is governed primarily by secondary side thermal hydraulic characteristics and the geometry (sizes of the tube-to-support gaps, TSP spacing, etc.). Tube wear at TSPs was first reported in the replacement steam generators during RF-1 5. The deepest TSP wear reported during RF-21 was, 29% TW. One indication of tube wear at FDB was-reported during RF-18. The FDB wear reported during RF-21 was 16% TW. Both wear mechanisms are located and sized using identical eddy current testing techniques.
With TSP wear being significantly deeper than FDB wear, TSP wear bounds the condition monitoring analysis results for FDB wear. To date, no tubes have been plugged for this mechanism.
3.3 Tube Wear Due to Foreign Objects Foreign objects may enter the SG through the feedwater stream and may cause tube wear. Foreign object wear was detected for the first time during RF-21 inspections.
During the bobbin probe base scope inspection program, an indication was noted at the top of the hot leg baffle plate in two tubes. When these locations were tested with a rotating coil probe, small indications indicative of wear were noted above the baffle plate. It was surmised that the wear was caused by a transient loose part. The tube at row 32, column 89 was sized at 7% TWD and the tube at row 35, column 92 was sized at 10% TWD. A review of the eddy current data from surrounding tubes was conducted with no additional indications discovered.
Due to the lack of similar indications reported during the inspection and the shallow nature of the indications, examination expansion was deemed unnecessary.
No loose parts associated with this wear were found in the Document Control Desk Attachment LTD 322 RC-14-0171 Page 4 of 11 vicinity of these indications during FOSAR activities so it is believed that the loose part no longer remains in the SG. No tubes were plugged for this mechanism.
4.0 Nondestructive Examination Techniques for Each Degradation Mechanism Inspection programs and Examination Technique Specification Sheets (ETSS) identified in the Degradation Assessment for each degradation mechanism are identified in Table 3.Table 3: SG Tube Degradation Mechanisms and Associated NDE Techniques Degradation Inspection Program Detection Sizing Mechanisms
% Sample Technique ETSS Technique__ _ _ I____ETSS
- Exidting Mechanisms
_Wear at AVBs 55% Bobbin 96004.1 96004.1 Wear at TSPs and 55% Bobbin, 96004.1 96004.1 FDB +Point Special Interest 96910.1 96910.1 Potential Mechanisms
___Wear due to Foreign 55% Bobbin, 27091.2 Objects +Point Special Interest, TTS +Point of 21998.1 21998.1 peripheral tubes in Alternate Alternate-HL and CL 2790X.3 2 2790X.3 2 Diagnostic Inspection Inspection of Row 55% Bobbin 96004.1 96004.1 1-2 U-bends ODSCC at HL TTS 20% + Point HL TTS 128424 128431 Expansion 128425 128432 Transition.
21410.1 21410.1 ODSCC at 55% Bobbin, 10013.1 Dings/Dents 50% +Point dents/dings
> 5V 24013.1 128413 128424 128431 128425 128432 21410.1 21410.1 S22841.3 Note 1: Prior to RF-21, wear due to foreign objects was classified as a potential degradation mechanism.
Note 2: The applicable ETSS are numbered 2790X.3 where X is a variable between 1 and 7. Techniques and corresponding uncertainty used for sizing of foreign object wear is dependent on indication geometry.
Document Control Desk Attachment LTD 322 RC-14-0171 Page 5 of 11 5.0 Location, Orientation (if Linear), and Measured Sizes (if Available) of Service Induced Indications Tables 4-8 provide the location and measured sizes of service induced indications.
The majority of the indications were from tube wear by support structures.
Two indications of wear due to a migratory loose part were detected in SG C.Table 4: AVB Wear SG Row Col Supp Inch Depth A 19 140 AV7 0.43 10 A 26 139 AV2 -0.75 7 A 26 139 AV7 0.75 9 C 26 3 AV2 -0.75 9 Table 5: SG A TSP Wear Row Col Supp Inch Depth 17 2 05C 0.31 13 1.4... 10 06C -0.57 15 1 10 05C 0.00 20 1 10 04C 0.00 13 67 24 07C 0.41 12 91 28 07C -0.70 11 1 60 03C -0.43 8 1 60 03C 0.47 16 115 62 09H ,:0.45 12 2 77 08H -0.43 26 2 77 07C 0.00 29 4 81 07C 0.52 10 4 99 05C 0.36 14 4 103 05C -0.56 8 4 103 05C 0.45 10 77 122 07C 0.50 12 Table 6: SG B TSP Wear Row Col Supp Inch Depth 1 2 06C -0.63 15 1 2 06C 0.43 15 1 2 05C -0.68 7 1 2 05C 0.38 14 34 9 06C 0.43 11 2 17 07H 0.49 11 1 28 07H -0.55 14 Document Control Desk Attachment LTD 322 RC-14-0171 Page 6 of 11 Row Col Supp Inch Depth 1 28 07H 0.49 11 2 31 06C 0.58 12 2 31 05C 0.48 17 99 44 09C 0.43 10 94 45 09C 0.43 8 3 46 08C -0.53 14 3 46 07C -0.49 19 108 47 07H -0.61 11 108 47 07H 0.41 11 3 54 06C 0.56 13 2 61 04C -0.59 14 115 74 06H 0.43 12 6 75 07C -0.54 13 114 75 04H 0.49 6 3 84 .07C 0.47 16 5 84 07C 0.52 11 5 88 08C -0.57 12 5 88 08C 0.48 12 5 88 07C '0.54 18 5 88 07C 0.48 11 5 88 05C -0.47 10 5 90 08C 0.50 13 5 90 07C 0.57 13 99 106 07C -0.61 12 2 113 07C 0.47 15 2 113 06C 0.50 16 5 116 07C 0.43 14 39 120 08C 0.36 8 10 121 08C --0.52 10 10 121 08C 0.58 7 54 121 07C -0.61 13 20 139 03H -0.58 11 1 140 06C -0.58 10 1 140 06C 0.41 14 Table 7: SG C TSP Wear and FDB Wear Row Col Supp Inch Depth 1 4 06C 0.38 11 1 8 07C 0.45 14 52 19 08C 0.47 10 25 36 07C 0.47 12 5 52 06C -0.56 15 7 54 06C -0.52 15 Document Control Desk Attachment LTD 322 RC-14-0171 Page 7 of 11 Row Col Supp Inch Depth 104 55 08C 0.38 25 110 73 08C 0.33 20 1 82 08H 0.53 14 114 83 09H 0.39 8 34 91 BPH 0.51 16 1 94 06H 0.59 21 6 107 05C -0.45 9 1 112 06C -0.52 19 1 112 06C 0.48 13 83 116 07C -0.55 13 30 137 08C 0.43 12 Table 8: SG C Foreign Object Wear Row Col .Supp Inch Depth 32 89 BPH 0.42 7 35 92 BPH 0.50 10 6.0 Tubes Plugged during the Inspection Outage No tube plugging was required during RF-21.7.0 Number and Percentage of Tubes Plugged to Date A total of eight (8) tubes have been removed from service by plugging in the VC Summer Steam Generators as shown in Table 9. The percentage of tubes plugged to date is less than 0.07% for each SG as shown in Table 10.Table 9: List of all Plugged Tubes through RF-21 SG Outage Year Row Col Ind Plugging Attribute A 7 RF-12 2000 26 25 NTE Anomalous tube-end expansion AK' RF-12 2000 26 31 NTE Anomalous tube-end expansion A RF-12 2000 94 51 NTE Anomalous tube-end expansion B Factory 1994 76 123 Factory-installed weld plug C Factory 1994 104 41 Factory-installed weld plug C RF-12 2000 57 96 NTE Anomalous tube-end expansion C Factory 1994 108 97 Factory-installed weld plug C RF-12 12000 99 100 NTE Anomalous tube-end expansion Note 1: Report RC-00-0345, dated 11/8/2000, incorrectly identified that the tubes at Row 25 Column 26 and Row 25 Column 31 were plugged in "A" Steam Generator during RF-12.-11 Document Control Desk Attachment LTD 322 RC-14-0171 Page 8 of 11 Table 10: SG Tube Plugging Percent ISG A SG B SG C Total Tubes 6307 6307 6307 Tubes Plugged 3 1 4 Effective Plugging 0.048% 0.016% 0.063%Percentage 8.0 Results of Condition Monitoring, Including Tube Pulls and In-Situ Testing Condition monitoring results for each existing degradation mechanism is outlined in the sections that follow. Based on inspection data, no tube pulls or in-situ testing was required in RF Condition monitoring concluded that SG performance criteria for leakage and structural integrity were satisfied for the preceding three'cycle operating interval prior to RF-21.8.1 Tube Wear at AVBs The maximum AVB wear indication from RF-21 was 10% TWD. This observed wear is well within the predicted wear from the operating assessment performed in RF-1 8 (approximated to be 21% TWD).ETSS 96004.1, Rev. 13, was utilized for sizing AVB wear. The ETSS correlation for true depth to indicated depth is: T = 0.98*1 + 2.89, where T is predicted depth and I is the indicated depth from EC.From this correlation, the predicted depth of the maximum indicated wear (10%) is 12.69%.The standard deviation for the correlation, a = 4.19. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645. Therefore, at 95% probability, the sizing uncertainty is 6.89% (1.645x4.19).
The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%probability is: Ut = 1.12
Document Control Desk Attachment LTD 322 RC-14-0171 Page 9 of 11 The potential maximum depth of the indication at a confidence level > 95/50 is: Maximum Predicted Indication 12.69% TWD NDE Sizing Uncertainty (95/50 CL) 7.72% TWD Total (>95/50) 20.41% TWD Structural Limit for AVB Wear 72% TWD Since the maximum observed AVB wear indication of 20.41 % TWD at 95% probability and 50% confidence is less than the structural limit of 72% TWD (defined in the RF-21 Degradation Assessment), the requirements for condition monitoring are met for AVB wear.8.2 Tube Wear at TSPs and FDB The maximum TSP and FDB wear indication from RF-21 was 29% TWD and 16%TWD, respectively.
This observed wear is well within the predicted wear from the operating assessment performed at RF-18 (approximated to be 40% TWD and 24%TWD, respectively).
..ETSS 96004.1, Rev. 13, was utilized for sizing TSP and FDB wear. The ETSS correlationh for true depth to indicated depth is: T = 0.98*1 + 2.89, where T is predicted depth and I is the indicated depth from EC.From this correlation, the predicted depth of the maximum indicatedwear (29%) is 31.31%.The standard deviation for the correlation, a = 4.19. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645. Therefore, at 95% probability, the sizing uncertainty is 6.89% (1.645 x 4.19).The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%probability is: Ut = 1.12
Document Control Desk Attachment LTD 322 RC-14-0171 Page 10 of 11 The potential maximum depth of the indication at a confidence level > 95/50 is: Maximum Predicted Indication 31.31% TWD NDE Sizing Uncertainty (95/50 CL) 7.72% TWD Total (>95/50) 39.03% TWD Structural Limit for TSP and FDB Wear 66% TWD Since the maximum observed TSP and FDB wear indication of 39.03% TWD at 95%probability and 50% confidence is less than the structural limit of 66% TWD (defined in the RF-21 Degradation Assessment), the requirements for condition monitoring are met for TSP and FDB wear.8.3 Tube Wear Due to Foreign Objects The maximum foreign object wear indication from RF-21 was 10% TWD.ETSS 27903.3, Rev. 1, was utilized for sizing foreign object wear. The ETSS correlation for true depth to indicated depth is: .T =1.21*1 + 26.3; .where T is predicted depth and I is the indicated depth from EC.From this correlation, the predicted depth of the maximum indicated wear (10%) is 38.4%. The standard deviation for the correlation, a = 4.99. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645.Therefore, at 95% probability, the sizing uncertainty is 8.21% (1.645x4.99).
The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%probability is: Ut = 1.12
- 8.21% = 9.19%The potential maximum depth of the indication at a confidence level > 95/50 is: Maximum Predicted Indication 38.4% TWD NDE Sizing Uncertainty (95/50 CL) 9.19% TWD Total (>95/50) 47.6% TWD Structural Limit for Foreign Object Wear 65% TWD Document Control Desk Attachment
-LTD 322 RC-14-0171 Page 11 of 11 Since the maximum observed foreign object wear indication of 47.6% TWD at 95%probability and 50% confidence is less than the structural limit of 65% TWD, the requirements for condition monitoring are met for foreign object wear.9.0 References 9.1 Westinghouse Document SG-SGMP-14-1, V.C. Summer 1 RF21 (April 2014)Steam Generator Degradation Assessment, Rev. 2, April 2014.9.2 Westinghouse Document SG-SGMP-14-15, V.C. Summer Unit 1 Steam Generator Cycle 21 Condition Monitoring and Cycles 22, 23, and 24 Operational Assessment, Rev. 0, May 2014.9.3 V.C. Summer Ul R21 SG Eddy Current Inspection Report, April 2014.'.' J.