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MONTHYEARML12236A1442012-08-28028 August 2012 Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.45 Project stage: RAI ML12311A0292012-11-0101 November 2012 Response to Request for Additional Information Regarding the 30-Day Report for Emergency Core Cooling System Model Changes Project stage: Response to RAI ML13074A7952013-03-0707 March 2013 Supplement to Response to Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes Project stage: Supplement ML13071A1022013-03-13013 March 2013 Request for Withholding Information from Public Disclosure for Millstone Power Station, Unit No. 2 Project stage: Withholding Request Acceptance ML13080A3062013-03-28028 March 2013 Request for Withholding Information from Public Disclosure for Millstone Power Station, Unit No. 2 Project stage: Withholding Request Acceptance ML13192A1022013-07-18018 July 2013 Closure Evaluation for 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 Project stage: Other 2013-03-13
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Category:Letter
MONTHYEARML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status ML24260A2192024-09-16016 September 2024 Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification ML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24248A2272024-09-0404 September 2024 Operator Licensing Examination Approval ML24240A1532024-09-0303 September 2024 Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel IR 05000336/20240052024-08-29029 August 2024 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005 IR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 ML24221A2872024-08-0808 August 2024 Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage IR 05000336/20244412024-08-0606 August 2024 Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only) ML24212A0742024-08-0505 August 2024 Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification ML24211A1712024-07-25025 July 2024 Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change IR 05000336/20244032024-07-22022 July 2024 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403 IR 05000336/20245012024-07-0101 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501 ML24180A0932024-06-28028 June 2024 Readiness for Additional Inspection: EA-23-144 IR 05000336/20240102024-06-26026 June 2024 Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010 ML24178A2422024-06-25025 June 2024 2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46 IR 05000336/20244402024-06-24024 June 2024 Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only) ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations ML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML24162A0882024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications ML24165A1292024-06-0505 June 2024 ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023 ML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities ML24110A0562024-05-21021 May 2024 Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML24141A2432024-05-20020 May 2024 Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15 ML24141A1502024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24142A0952024-05-20020 May 2024 End of Cycle 22 Steam Generator Tube Inspection Report IR 05000336/20240012024-05-14014 May 2024 Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation ML24123A2272024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24123A2042024-05-0202 May 2024 Pre-Decisional Replay to EA-23-144 IR 05000336/20244012024-04-30030 April 2024 Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only) ML24123A1222024-04-30030 April 2024 Inservice Inspection Program - Owners Activity Report, Refueling Outage 22 ML24116A0452024-04-25025 April 2024 Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only) ML24116A1742024-04-24024 April 2024 Annual Radiological Environmental Operating Report ML24114A2662024-04-24024 April 2024 Submittal of 2023 Annual Radioactive Effluent Release Report ML24103A0202024-04-22022 April 2024 Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits ML24106A2032024-04-15015 April 2024 2023 Annual Environmental Operating Report ML24088A3302024-04-0404 April 2024 Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel ML24093A1022024-04-0101 April 2024 Alternative Request IR-4-13, Proposed Alternative Request to Support Steam Genera Tor Channel Head Drain Modification ML24093A2162024-04-0101 April 2024 Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits IR 05000336/20240112024-04-0101 April 2024 Comprehensive Engineering Team Inspection - Inspection Report 05000336/2024011 and 05000423/2024011 ML24092A0752024-03-28028 March 2024 3R22 Refueling Outage Inservice Inspection (ISI) Owners Activity Report Extension ML24088A2352024-03-26026 March 2024 Decommissioning Funding Status Report ML24086A4762024-03-22022 March 2024 Application for Technical Specification Change to Extend the Inspection Interval for Reactor Coolant Pump Flywheels Using the Consolidated Line-Item Improvement Process ML24086A4802024-03-22022 March 2024 Alternative Request IR-4-14, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examination for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles ML24051A1922024-03-0808 March 2024 – Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) IR 05000336/20230062024-02-28028 February 2024 Annual Assessment Letter for Millstone Power Station, Units 2 and 3, (Reports 05000336/2023006 and 05000423/2023006) ML24053A2632024-02-21021 February 2024 Unit 3, and Independent Spent Fuel Storage Installation, Notification Pursuant to 10 CFR 72.212(b)(1) Prior to First Storage of Spent Fuel Under a General License ML24057A0612024-02-19019 February 2024 and Virgil C. Summer Power Nuclear Stations - Nuclear Property Insurance Coverage 2024-09-04
[Table view] Category:Report
MONTHYEARML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML24093A1022024-04-0101 April 2024 Alternative Request IR-4-13, Proposed Alternative Request to Support Steam Genera Tor Channel Head Drain Modification ML24086A4802024-03-22022 March 2024 Alternative Request IR-4-14, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examination for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles ML23324A4222023-11-20020 November 2023 Reactor Vessel Internals Inspections Aging Management Program Submittal Related to License Renewal Commitment 13 ML23324A4302023-11-20020 November 2023 Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment 15 ML23188A0202023-07-26026 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23151A0742023-06-12012 June 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML23103A2282023-04-12012 April 2023 Stations Units 1 and 2; Millstone Power Station Units 2 and 3, DOM-NAF-2-P/NP-A, Revision 0.4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML22353A6202022-12-19019 December 2022 Request for Approval of Appendix F Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code ML22193A1432022-06-23023 June 2022 5 to Updated Final Safety Analysis Report, Technical Requirements Manual Current Through Change No. 207 ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML21113A1352021-04-27027 April 2021 Review of the Spring 2017 Steam Generator Tube Inspection Report ML21042B3212021-02-11011 February 2021 Stations, Units 1 & 2; Millstone Power Station, Units 2 & 3 - Request for Approval of Fleet Report DOM-NAF-2 Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML20352A3342020-12-17017 December 2020 Proposed Alternative Requests RR-05-04 and IR-4-02, Use of Alternative Pressure/Flow Testing Requirements for Service Water System Supply Piping ML20345A3682020-12-16016 December 2020 Review of the Fall 2017 and Spring 2019 Steam Generator Tube Inspection Reports 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 ML20171A5342020-08-13013 August 2020 Staff Assessment of Flood Hazard Focused Evaluation and Integrated Assessment 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 ML20105A0782020-04-14014 April 2020 Supplement to License Amendment Request to Revise TS 3.8.1.1, A.C. Sources - Operating, to Support Maintenance and Replacement of the 'A' Reserve Station Service Transformer and 345 Kv South Bus Switchyard Components ML19352B8982019-12-17017 December 2019 Proposed Alternative Request RR-05-05, Containment Unbonded Post-Tensioning System Inservice Inspection Requirements ML19246A1162019-10-0707 October 2019 Supplement to Staff Assessment of Response to 10 CFR 50.54(F) Information Request - Flood Causing Mechanism Reevaluation(Epid Nos. 000495\05000336\L- 2015-JLD-0011 and 000495\05000423\L-2015-JLD-0012 - (2019Aug21) ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19211B1682019-07-24024 July 2019 Day Special Report for One Train of Reactor Vessel Level Monitor Inoperable ML19070A2172019-04-0303 April 2019 Supplement to Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request - Flood - Causing Mechanism Reevaluation ML19064A5902019-02-28028 February 2019 Proposed Alternative Request IR-3-39, Alternative to ASME Code, Section XI, IWA-4221(C), to Permit Two Fillet Welds Not in Compliance with the Construction Code to Remain in Service ML19011A1722019-01-0404 January 2019 Enclosure 3, Attachments 2C-3C - MPS3 EAL Technical Bases Document (Marked-Up) ML19011A1732019-01-0404 January 2019 Enclosure 4 - North Anna Power Station Units 1 & 2, EAL Scheme Revisions-Supporting Documents ML19011A1742019-01-0404 January 2019 Enclosure 5 - Surry Power Station, EAL Scheme Revisions-Supporting Documents ML18256A2002018-10-0303 October 2018 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation (EPID L00495\05000336 L-2015-JLD-0011 and 00495\05000423\L-2015-JLD-0012) ML18247A2752018-06-18018 June 2018 Technical Requirements Manual ML17187A1692017-06-28028 June 2017 Mitigating Strategies Assessment (MSA) Report) ML17108A3272017-04-0606 April 2017 Reactor Vessel Standby Surveillance Capsule Z Dosimetry Analysis and Storage Confirmation ML17051A0012017-02-27027 February 2017 Summary of the NRC Staff'S Review of the Spring 2015 Steam Generator Tube Inservice Inspections ML16193A6702016-06-30030 June 2016 ISFSI - 10 CFR 50.59, 10 CFR 72.48 Change Report for 2014 and 2015, and Commitment Change Report for 2015 ML15328A2682015-12-15015 December 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 for Recommendation 2.1 of the Near-Term Task Force Review ML15275A2882015-10-19019 October 2015 Summary of the NRC Staff'S Review of the Fall 2014 Steam Generator Tube Inservice Inspections ML15253A2062015-09-0101 September 2015 ANP-3315NP, Revision 0, Millstone Unit 2 M5 Upgrade, Small Break LOCA Analysis Licensing Report. ML15194A0572015-06-30030 June 2015 ISFSI - NRC Commitment Change Report for 2014 ML15078A2062015-03-12012 March 2015 to Engineering Evaluation 14-E16, Dominion Flooding Hazard Reevaluation Report for Millstone, Units 2 and 3, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Pp. 2-58 Through 2-125 ML15078A2082015-03-12012 March 2015 to Engineering Evaluation 14-E16, Dominion Flooding Hazard Reevaluation Report for Millstone, Units 2 and 3, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Pp. 2-172 Through the End ML15078A2072015-03-12012 March 2015 to Engineering Evaluation 14-E16, Dominion Flooding Hazard Reevaluation Report for Millstone, Units 2 and 3, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Pp. 2-126 Through 2-171 ML15078A2052015-03-12012 March 2015 to Engineering Evaluation 14-E16, Dominion Flooding Hazard Reevaluation Report for Millstone, Units 2 and 3, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Pp. 1 Through 2-57 ML14220A0172014-07-30030 July 2014 Startup Test Report for Cycle 23 ML13338A4332014-01-31031 January 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14013A2712014-01-30030 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 ML13357A3982014-01-24024 January 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Milestone Power Station, Unit 2, TAC No.: MF0858 ML14006A1592014-01-0808 January 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Millstone Power Station, Unit 3, TAC No.: MF0859 2024-06-10
[Table view] Category:Miscellaneous
MONTHYEARML23324A4302023-11-20020 November 2023 Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment 15 ML23188A0202023-07-26026 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23151A0742023-06-12012 June 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML22353A6202022-12-19019 December 2022 Request for Approval of Appendix F Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code ML21113A1352021-04-27027 April 2021 Review of the Spring 2017 Steam Generator Tube Inspection Report ML20345A3682020-12-16016 December 2020 Review of the Fall 2017 and Spring 2019 Steam Generator Tube Inspection Reports 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 ML20171A5342020-08-13013 August 2020 Staff Assessment of Flood Hazard Focused Evaluation and Integrated Assessment ML20105A0782020-04-14014 April 2020 Supplement to License Amendment Request to Revise TS 3.8.1.1, A.C. Sources - Operating, to Support Maintenance and Replacement of the 'A' Reserve Station Service Transformer and 345 Kv South Bus Switchyard Components ML19211B1682019-07-24024 July 2019 Day Special Report for One Train of Reactor Vessel Level Monitor Inoperable ML17187A1692017-06-28028 June 2017 Mitigating Strategies Assessment (MSA) Report) ML17108A3272017-04-0606 April 2017 Reactor Vessel Standby Surveillance Capsule Z Dosimetry Analysis and Storage Confirmation ML17051A0012017-02-27027 February 2017 Summary of the NRC Staff'S Review of the Spring 2015 Steam Generator Tube Inservice Inspections ML16193A6702016-06-30030 June 2016 ISFSI - 10 CFR 50.59, 10 CFR 72.48 Change Report for 2014 and 2015, and Commitment Change Report for 2015 ML15328A2682015-12-15015 December 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 for Recommendation 2.1 of the Near-Term Task Force Review ML15275A2882015-10-19019 October 2015 Summary of the NRC Staff'S Review of the Fall 2014 Steam Generator Tube Inservice Inspections ML15194A0572015-06-30030 June 2015 ISFSI - NRC Commitment Change Report for 2014 ML14220A0172014-07-30030 July 2014 Startup Test Report for Cycle 23 ML14013A2712014-01-30030 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 ML13303B9072013-10-17017 October 2013 10 CFR 71.95 Report - 8-120B Cask Certificate of Compliance Noncompliance Due to an Inadequate Vendor Leak Test Procedure ML13192A1022013-07-18018 July 2013 Closure Evaluation for 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML12362A3922012-11-30030 November 2012 Spent Fuel Pool Criticality Analysis with No Credit for Boraflex. ML12353A2422012-11-27027 November 2012 Report in Response to March 12, 2012 Information Request Regarding Seismic Aspects of Recommendation 2.3 ML12187A1752012-06-25025 June 2012 ISFSI - 10 CFR 50.59, 10 CFR 72.48 Change Report for 2010 and 2011, and Commitment Change Report for 2010 and 2011 ML12151A3592012-06-0707 June 2012 End of Cycle 20 Steam Generator Tube Inservice Inspection Report Review ML12046A8362012-01-31031 January 2012 Inservice Inspection Program - Owner'S Activity Report ML12031A1472012-01-25025 January 2012 Day Report for Emergency Core Cooling System Model Changes of 10 CFR 50.46 ML1116002202011-06-13013 June 2011 Exhibit D (Part 3): Millstone 2001 Safstor Cost Analysis (Normal Dollars) ML1028504382010-09-30030 September 2010 Attachments 2, 3, 5, 7, 8 & 9, to 10-579A, Westinghouse Electric Co. LLC, WCAP-17071-NP, H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F), Revision 2, Dated Sep ML1019403352010-06-30030 June 2010 Units, 1, 2, 3, & ISFSI, 10 CFR 50.59, 10 CFR 72.48 Change Report for 2008 and 2009, and the Commitment Change Report for 2009 ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants ML1009702062010-03-29029 March 2010 Chemistry Procedure, Liquid Waste Sample Sink, Rev. 003-00 ML0931701762009-11-12012 November 2009 Part 3 of 4--Weld Overlay Examination Report 2R19 ML0931701732009-11-12012 November 2009 Part 2 of 4--Weld Overlay Examination Report 2R19 ML0931701662009-11-12012 November 2009 Part 1 of 4--Weld Overlay Examination Report 2R19 ML0931701852009-09-23023 September 2009 Part 4 of 4--Weld Overlay Examination Report 2R19 ML0921908942009-08-0404 August 2009 Units 1 & 2, Millstone, Units 2 and 3 and Kewaunee - Approved Topical Report DOM-NAF-2, Revision 0.1-A ML0913204602009-04-28028 April 2009 2008 Occupational Radiation Exposure Information ML0901303822008-12-26026 December 2008 Revision 002 to CP 3804L Procedure Action Request, Pass Containment Air Sample, Dated 12/26/2008 ML0830903962008-10-24024 October 2008 End of Cycle 18 Steam Generator Tube Inspection Report ML0827701132008-09-30030 September 2008 Response to Request for Additional Information Regarding Spent Fuel Pool Storage License Amendment Request ML0819800752008-07-0707 July 2008 Revision to Control Room and OSC Assembly Area Building 475 1st Floor, and SAP Locker ML0818406072008-07-0101 July 2008 Final Evaluation of Weld Overlay Examinations in Refueling Outage 18 ML0727504022007-09-27027 September 2007 Alternative Request RR-89-64 for Use of a Limited One-Sided Ultrasonic Examination Technique LBDCR 07-MP2-034 ML0720003902007-07-13013 July 2007 Attachment 1, Millstone Power Station Unit 3, License Amendment Request, Stretch Power Uprate, Descriptions, Technical Analysis and Regulatory Analysis for the Proposed Operating License and Technical Specifications Changes ML0715203992007-05-30030 May 2007 Fy 2007 Final Fee Rule Workpapers ML0713501882007-05-17017 May 2007 Review of Steam Generator Tube Inservice Inspection Report for the 2005 Refueling Outage at Millstone Power Station, Unit No. 2 ML0713502492007-05-0808 May 2007 Steam Generator Tube Plugging Report ML0708805652007-03-28028 March 2007 Response to Request for Additional Information Regarding an Alternative for the Weld Overlay of Pressurizer Nozzle Welds ML0705903272007-02-27027 February 2007 Fitness-For-Duty Program Semi-Annual Performance Data Report 2023-07-26
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UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 July 18, 2013 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 MILLSTONE POWER STATION, UNIT NO.2 -CLOSURE EVALUATION FOR 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 (TAC NO. ME7881) Dear Mr. Heacock: By letter dated January 25,2012, as supplemented by letters dated November 1, 2012, and March 7,2013, Dominion Nuclear Connecticut Inc. (Dominion), submitted a report describing cumulative errors and changes identified in the small break loss of coolant accident emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the changes on the predicted peak cladding temperature for Millstone Power Station Unit 2. This report was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 46 (10 CFR 50.46), paragraph (a)(3). A meeting was held between the Nuclear Regulatory Commission (NRC) staff and Dominion personnel along with their vendor, AREVA, on May 14, 2013 to discuss the sensitivity studies performed to support the review. The NRC staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR), for the realistic ECCS evaluations revision of 10 CFR 50.46 (53 FR 35996). The NRC staff's closure evaluation of the report is enclosed. Please contact me at (301) 415-4125 if you have any questions on this issue. Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336 Enclosure: As stated cc w/encl: Distribution via Listserv UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 1.0 INTRODUCTION By letter dated January 25. 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12031A147). Dominion Nuclear Connecticut Inc. (Dominion), submitted a report describing cumulative errors and changes identified in the small break loss-of-coolant accident (SBLOCA) emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the changes on the predicted peak cladding temperature (PCT) for Millstone Power Station Unit 2. This report was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 46 (10 CFR 50.46), paragraph (a)(3). The report was supplemented by two additional letters dated November 1. 2012 (ADAMS Accession No. ML 12311A029) and March 7, 2013 (ADAMS Accession No. ML 13074A795). There was also a meeting held between the NRC staff and Dominion personnel along with their vendor, AREVA, on May 14, 2013 to discuss the sensitivity studies performed to support the review. The Nuclear Regulatory Commission (NRC) staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3). and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR), for the realistic ECCS evaluations revision of 10 CFR 50.46 (53 FR 35996). The staff review is discussed in the following sections of this closure evaluation. 2.0 REGULATORY EVALUATION 2.1 Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCS for light water nuclear power reactors are promulgated at 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change to, or error in, an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature different by more than 50 degrees Fahrenheit CF) from the temperature calculated for the limiting transient using the last acceptable model, or is an accumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 OF. Enclosure
-2 For each change to or error discovered in an acceptable evaluation model or in the application of such a model, paragraph (a)(3)(ii) to 10 CFR 50.46 requires the affected licensee to report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually. If the change or error is significant, the licensee is required to provide this report within 30 days and include with the report a proposed schedule for providing a re-analysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements. 2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the statement of considerations published on September 16, 1988, in the FR for the best estimate loss-of-coolant-accident (LOCA) revision of 10 CFR 50.46 (53 FR 35996): [Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation modeL .. Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model. .. More timely reporting (30 days) is required for significant errors or changes ... the final rule revision also allows the NRC to determine the schedule for reanalYSis based on the importance to safety relative to other applicant or licensee requirements. The NRC staff considered the discussion in the Federal Register in its evaluation of the changes in the report submitted by the licensee. 3.0 TECHNICAL EVALUATION The report submitted by the licensee described the effects of changes in the SBLOCA ECCS evaluation model associated with the S-RELAP5 kinetics and heat conduction model changes and the Sieicher-Rouse single phase vapor heat transfer correlation. Based on the nature of the reported changes, and on the magnitude of their effect on the PCT calculation, the NRC staff determined that a detailed technical review is necessary. Based on the regulatory evaluation discussed above, the staff's review was performed to ensure that the NRC staff agrees with the licensee's assessment of the Significance of the changes, and to enable the staff to verify that the evaluation model, as a whole, remains adequate. Finally, the NRC staff's review also establishes that the licensee's proposed schedule for reanalysis is acceptable in light of the safety significance of the reported error.
-3.1 Summary Of Technical Information In The Report The licensee's report indicated that the effect of the model changes was 87 OF for Millstone Power Station Unit 2. The nature of the changes, and the method used to estimate its effect on the calculated peak fuel cladding temperature, are briefly discussed in the original report and significantly more detail is provided in the response to the NRC staff request for additional information dated January 25, 2012. S-RELAP5 Kinetics and Heat Conduction Model Changes Corrections to the coding of the point kinetics model in SRELAP-5 were previously provided by Idaho National Lab (INL) and incorporated into the code. The INL recently announced that the corrections were incorrect and that the recommended convergence criteria supplied with the corrections should be retained. The INL also announced that the heat conduction solution was incorrectly programmed. AREVA entered both of these items into their corrective action system. The programming error in the heat conduction solution was associated with using the incorrect heat capacity when evaluating the right boundary mesh point, the code incorrectly used the next to last instead of the last mesh interval heat capacity. This error is minimized by the SBLOCA methodology guidelines requiring close mesh spacing. The corrections for the two errors were installed into a new S-RELAP5 version and new evaluations were performed. The impact of the changes on PCT for Millstone Power Station Unit 2 was determined to be +4 OF. Sleicher-Rouse Single Phase Vapor Heat Transfer Correlation AREVA noticed that the behavior of the Sieicher-Rouse single phase vapor heat transfer correlation differed from other correlations while developing a boiling water reactor LOCA analysis methodology using S-RELAP5. It was discovered that the formulation of the correlation in S-RELAP5 differed from the formulation used in other industry codes. AREVA prepared a version of the code with the alternative formulation and found that it more closely agreed with the formulation used in other industry codes. The results of the sample problems run with the corrected Sieicher-Rouse correlation indicate that the predicted SBLOCA PCT for Millstone Power Station Unit 2 should be increased 83 oF. Reported Results Following the changes to the S-RELAP5 models and the Sieicher-Rouse correlation, the current predicted PCT for Millstone Power Station Unit 2 is 1801 OF. The margin utilization table attached to the report also included RODEX2 thermal conductivity degradation. This error was reported to be a 0 OF change to PCT. 3.2 Summary of Staff Evaluation In its evaluation, the NRC staff reviewed (1) the approach used to estimate the effects of the changes, (2) the estimated effect of the changes, and (3) the licensee's proposal for re-analysis in consideration of the approach used to estimate the effects of the changes. As discussed in
-the following paragraphs, the NRC staff determined that the licensee's estimate and proposal for reanalysis are acceptable. The changes proposed to the S-RELAP5 point kinetics and heat conduction models were simple, straight-forward corrections to errors reported by INL. The staff questioned the bundling of the errors and reporting a single value for the delta effect on the PCT. Dominion responded that none of the other changes and errors provided are bundles of more than one error or change and were reported separately to confirm to the staff that the errors did not offset each other and that the absolute value of the effects separately was not greater that the 4 of delta reported. This also led the staff to question whether any of the magnitudes of the other changes reported to the analysis in the past had been bundled. Dominion confirmed that no other delta PCTs listed in the report were a bundle of multiple changes. The initial report contained a vague description of the changes made to the Sieicher-Rouse correlation in S-RELAP5. The staff requested that Dominion provide more information detailing the changes that were made to the correlation and how it was modified to more closely agree with other industry code formulations of the correlation. The S-RELAP5 form and the alternative form of the correlation were both provided to the staff. AREVA performed a comparison of the two formulations and the staff confirmed that the alternate form of the correlation that was implemented in S-RELAP5 more accurately models the temperature degradation factor. Dominion reported, in the rack-up table, that the effect of the lack of thermal conductivity degradation consideration in the RODEX2 code produced a zero effect on SBLOCA PCT. A section of explanation was not provided in the report regarding this error. At the request of the NRC staff, Dominion provided a justification to support the reported estimate. The basis of the argument is that the PCT for SBLOCA does not occur until later in the transient, is therefore dependent on the decay heat versus heat transfer, and is not coupled to the initial stored energy within the fuel. The staff accepts the licensee's evaluation of thermal conductivity degradation with respect to SBLOCA. Recent information gained through other reviews led the staff to question other aspects of the Millstone Power Station Unit 2 SBLOCA analysis. The specific areas of concern were with the coarseness of the break spectrum and the credit for the hot leg nozzle gaps and core barrel leakage path modeling in the S-RELAP5 nodalization. Dominion had AREVA perform thorough sensitivity studies to investigate the effects of the staffs concerns on the outcome of the SBLOCA analysis. The studies showed that the PCT reported in the Dominion letter dated January 25,2012 remained conservative. The staff also questioned the limiting break analysis which had multiple loop seal clearing behavior in the suction legs. Typically, PCT is maximized when only the broken loop seal clears due to the increased resistance of vapor flow through only one loop versus multiple venting loops. Because of these concerns, the staff requested the licensee to re-analyze the limiting break with credit for the leakage paths removed and with the constraint that only a single loop seal clear during the transient. The reanalysis demonstrated that the PCT was less than 1801 OF for the limiting break size of 0.083 fe. This current analysis of record showed that multiple loop seals partially cleared allowing vapor from the intact loop to entrain downcomer liquid and expel this liquid out the break. This caused the downcomer head of water to decrease, which increased the depth of core uncovery and a higher PCT. The staff agrees that the multiple loop seal partial clearing produces a higher PCT than that for the case with only the
-broken loop seal cleared. With vapor venting through only the broken loop, the potential for entraining downcomer liquid and expelling it out the break is precluded. The staff agrees that the current model with partial loop seal clearing produces a conservative PCT result for this limiting break. The staff further notes that the licensee should assure that this behavior remains dominant in all future analyses by comparison to the case where only one loop seal is allowed to clear for the limiting break. In the supplemental letter dated November 1,2012, Dominion proposed to submit a reanalysis of the SBLOCA event within one year of NRC approval of Supplement 1 to EMF-2328(P)(A). The NRC staff determines herewith, that the licensee's proposed schedule for reanalysis is acceptable and that the reanalysis requirement of 10 CFR 50.46 is presently satisfied. In summary, the NRC staff reviewed the licensee's report estimating the effect of changes on the small break LOCA analyses for Millstone Power Station Unit 2. Based on the technical rigor employed by the licensee, which included performing significant sensitivity studies on the SBLOCA analysis, the NRC staff concluded that the change estimate was acceptable. Also, the NRC staff reviewed the licensee's proposed schedule for reanalysis and determined that the licensee satisfied the reanalysis requirement set forth in 10 CFR 50.46(a)(3)(ii). 4.0 CONCLUSION Based on the considerations discussed above, the NRC staff finds that the report submitted pursuant to 10 CFR 50.46(a)(3), concerning multiple ECCS evaluation model errors, satisfies the intent of the 10 CFR 50.46 reporting requirements. The report and supplemental information enabled the staff to (1) determine that it agrees with the licensee's assessment of the significance of the error, (2) confirm that the evaluation model remains adequate, and (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b). The NRC staff concludes that the licensee's sensitivity studies and proposed schedule for reanalysis is acceptable and, therefore, the requirements of 10 CFR 50.46 are presently satisfied. Principal Contributors: A. Proffitt L. Ward Date: July 18, 2013 July 18, 2013 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 SUBJECT: MILLSTONE POWER STATION, UNIT NO.2 -CLOSURE EVALUATION FOR 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 (TAC NO. ME7881) Dear Mr. Heacock: By letter dated January 25,2012, as supplemented by letters dated November 1,2012, and March 7,2013, Dominion Nuclear Connecticut Inc. (Dominion), submitted a report describing cumulative errors and changes identified in the small break loss of coolant accident emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the changes on the predicted peak cladding temperature for Millstone Power Station Unit 2. This report was submitted pursuant to Title 10 of the Code ofFederal Regulations (10 CFR), Part 50, Section 46 (10 CFR 50.46), paragraph (a)(3). A meeting was held between the Nuclear Regulatory Commission (NRC) staff and Dominion personnel along with their vendor, AREVA, on May 14, 2013 to discuss the sensitivity studies performed to support the review. The NRC staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published on September 16, 1988, in the Federal Register (FR), for the realistic ECCS evaluations revision of 10 CFR 50.46 (53 FR 35996). The NRC staff's closure evaluation of the report is enclosed. Please contact me at (301) 415-4125 if you have any questions on this issue. Docket No. 50-336 Enclosure: As stated cc w/encl: Distribution via Listserv DISTRIBUTION: PUBLIC RidsAcrsAcnwMailCenter RidsNrrLAKGoldstein (paper copy) RidsNrrDssSrxb ADAMS ACCESSION NO.. ML13192A102 Sincerely, Ira! James Kim. Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPL 1-1 R/F RidsNrrDorlLpl-1 RidsNrrPMMilistone (paper copy) RidsNrrDssSnpb FBower, RI RidsRgn'1 MailCenter AProffitt, NRR AGuzzetta, NRR *See memo daet d J I 10, 2013UIY OFFICE NRR/LPL1-1/PM NRR/LPL1-1/LA NRR/DSS/SNPB/BC NRR/LPL1-11BC (A) NAME JKim KGoldstein SWhaley* RBeall (DPickett for) DATE 7/17113 7/12/2013 7/10/2013 7/18/13 OFFICIAL RECORD COpy