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MONTHYEARML13120A1582013-04-25025 April 2013 License Amendment Request to Revise Technical Specification 6.8.4.F for Peak Calculated Containment Internal Pressure Project stage: Request ML13150A2292013-05-28028 May 2013 Acceptance Review Results Regarding Millstone 3 - Revised Peak Calculated Containment Internal Pressure LAR (MF1731) Project stage: Acceptance Review ML13218A3102013-08-0808 August 2013 Request for Additional Information Regarding Amendment for a Revised Calculated Peak Containment Internal Pressure Project stage: RAI ML13275A2402013-09-19019 September 2013 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 6.8.4.F for Peak Calculated Containment Internal Pressure Project stage: Response to RAI ML13305B0052013-11-0606 November 2013 Request for Additional Information Regarding License Amendment Request to Revise Technical Specification for Calculated Peak Containment Internal Pressure Project stage: RAI ML13353A2962013-12-11011 December 2013 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 6.8.4.F for Peak Calculated Containment Internal Pressure Project stage: Response to RAI ML14073A0552014-04-0808 April 2014 Issuance of Amendment Calculated Containment Internal Pressure Project stage: Approval ML14142A0962014-06-10010 June 2014 Correction Letter to License Amendment No. 259 Project stage: Other 2013-05-28
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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:License-Operating (New/Renewal/Amendments) DKT 50
MONTHYEARML24128A2772024-06-0404 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) 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) ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML23058A4542023-03-16016 March 2023 Issuance of Amendment Nos. 345 and 285 Regarding Adoption of Technical Specification Task Force-359, Increase Flexibility in Mode Restraints ML23005A1702023-01-17017 January 2023 Correction to Issuance of Amendment No. 283 to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definition ML22095A1072022-07-11011 July 2022 Issuance of Amendment Nos. 120, 344, & 284, 293 & 276, & 307 & 307 to Relocate Requirements to the QAPD ML22041A0102022-03-0101 March 2022 V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition ML22007A1512022-02-16016 February 2022 Issuance of Amendment No. 282 Regarding Shutdown Bank Technical Specification Requirements and Alternate Control Rod Position Monitoring Requirements ML21326A0992022-01-0707 January 2022 Issuance of Amendment No. 281 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-177642-P-A, Revision 1 ML21262A0012021-11-0909 November 2021 Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate ML21227A0002021-10-0505 October 2021 Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident ML21222A2302021-09-0909 September 2021 Issuance of Amendment No. 343 Revision to Technical Specifications for Steam Generator Inspection Frequency (L-2020-LLA-0227) ML21043A1622021-03-25025 March 2021 Issuance of Amendment No. 278 Regarding Revision to Battery Surveillance Requirements ML21026A1422021-02-23023 February 2021 Issuance of Amendment No. 342 Revision to Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20275A0002020-10-14014 October 2020 Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections ML20237H9952020-09-29029 September 2020 Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program ML20191A0042020-08-0707 August 2020 Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity ML20161A0002020-07-15015 July 2020 Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML20140A3692020-06-24024 June 2020 Issuance of Amendment No. 339 Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time ML19340A0252020-01-30030 January 2020 Issuance of Amendment No. 337 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components of Nuclear Power Reactors ML19305D2482019-12-31031 December 2019 Issuance of Amendments Adoption of Emergency Action Level Schemes Per NEI 99-01, Rev. 6 ML19150A3672019-05-30030 May 2019 Current Facility Operating License No. DPR-21, Tech Specs, Revised 5/30/2019 ML19126A0002019-05-28028 May 2019 Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage ML19042A2772019-03-21021 March 2019 Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML18246A0072018-09-25025 September 2018 Issuance of Amendment No. 335 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML18038B2002018-02-26026 February 2018 Issuance of Amendment Nos. 118, 334, and 271 to Revise Licensee'S Name (CAC Nos. MF9844, MF9845, and MF9848; EPID L-2017-LLA-0245 and EPID L-2017-LLA-0346) ML17025A2182017-01-24024 January 2017 Issuance of Amendment Realistic Large Break Loss-of-Coolant Accident Analysis ML16308A4852016-12-22022 December 2016 Issuance of Amendment No. 331 Revision to Emergency Core Cooling System Technical Specification and Final Safety Analysis Report Ch. 14 to Remove Charging ML16249A0012016-09-30030 September 2016 Issuance of Amendments Small Break Loss of Coolant Accident Reanalysis ML16206A0012016-08-0404 August 2016 Issuance of Amendment No. 270 to Revise Technical Specification 5.6.3, Fuel Storage Capacity ML16193A0012016-07-28028 July 2016 Issuance of Amendments Removal of Severe Line Outage Detection from the Offsite Power System ML16131A7282016-07-28028 July 2016 Issuance of Amendment 268 Adopting Dominion Core Design and Safety Analysis Methods and Addressing the Issues Identified in Three Westinghouse Communication Documents ML16189A0762016-06-29029 June 2016 Supplement to Information Regarding License Amendment Request for Removal of Severe Line Outage Detection from the Offsite Power System ML16003A0082016-06-23023 June 2016 Issuance of Amendment No. 327 Proposed Technical Specification Changes for Spent Fuel Storage ML16068A3122016-03-31031 March 2016 Issuance of Amendment No. 326 to Revise Technical Specifications for Containment Leak Rate Testing ML16011A4002016-01-29029 January 2016 Issuance of Amendments Nos. 325 and 267 to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation ML15288A0042015-11-30030 November 2015 Issuance of Amendment No. 266: Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times ML15280A2422015-10-29029 October 2015 Issuance of Amendment No. 324 Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, Adoption of TSTF-425,Rev 3 ML15245A4822015-10-0707 October 2015 Millstone Power Station, Surry Power Station, and North Anna Power Station - Issuance of Amendments to Revise the Cyber Security Milestone 8 Completion Date in the Renewed Facility Operating Licenses ML15225A0102015-08-28028 August 2015 Issuance of Amendment No. 264 Surveillance Requirement 4.4.4.2, Reactor Coolant System Relief Valves ML15246A1422015-08-27027 August 2015 Connecticut, Inc. Millstone Power Station Unit 3 Supplement to License Amendment Request to Revise Surveillance Requirement 4.4.4.2, Reactor Coolant System Relief Valves ML15225A0082015-08-26026 August 2015 Issuance of Amendment No. 322 Revision to the Final Safety Analysis Report - Examination Requirements for ANSI B31.1.0 Piping Welds ML15187A3262015-07-29029 July 2015 Issuance of Amendment No. 321, Adoption of TSTF-426, Revision 5, Revise or Add Actions to Preclude Entry Into LCO 3.0.3 - RITSTF Initiatives 6B & 6C ML15187A1862015-07-28028 July 2015 Issuance of Amendment No. 263, Regarding Technical Specification Changes to Auxiliary Feedwater System (MF4692) ML15187A0112015-07-27027 July 2015 Issuance of Amendment No. 262, Refueling Water Storage Tank Allowable Temperatures ML15093A0022015-05-20020 May 2015 Issuance of Amendment No. 320, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15098A0342015-05-20020 May 2015 Issuance of Amendment No. 261, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15093A4412015-05-18018 May 2015 Issuance of Amendment No. 319, Use of Areva M5 Alloy Clad Fuel Assemblies ML14178A5992014-07-11011 July 2014 Issuance of Amendment Proposed Changes to Technical Specification 3/4 7.5, Ultimate Heat Sink (Tac MF1780) 2024-06-04
[Table view] Category:Safety Evaluation
MONTHYEARML24128A2772024-06-0404 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) 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) ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23226A0052023-09-26026 September 2023 Issuance of Amendment No. 287 Supplement to Spent Fuel Pool Criticality Safety Analysis ML23188A0432023-07-31031 July 2023 Authorization and Safety Evaluation for Alternative Request No. IR-04-11 ML23175A0052023-07-12012 July 2023 Alternative Request P-07 for Pump Periodic Verification Testing Program for Containment Recirculation Spray System Pumps ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML23058A4542023-03-16016 March 2023 Issuance of Amendment Nos. 345 and 285 Regarding Adoption of Technical Specification Task Force-359, Increase Flexibility in Mode Restraints ML21320A0072022-09-0707 September 2022 Review of Appendix E to DOM-NAF-2, Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code (EPID L-2021-LLT-0000) (Non-Proprietary) ML22201A5082022-07-28028 July 2022 Authorization and Safety Evaluation for Alternative Request No. IR-04-09 ML22095A1072022-07-11011 July 2022 Issuance of Amendment Nos. 120, 344, & 284, 293 & 276, & 307 & 307 to Relocate Requirements to the QAPD ML22039A3392022-03-0303 March 2022 Request for Alternative Frequency to Supplemental Valve Position Verification Testing Requirements in the Fourth 10-year Valve Inservice Testing Program ML22041A0102022-03-0101 March 2022 V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition ML22007A1512022-02-16016 February 2022 Issuance of Amendment No. 282 Regarding Shutdown Bank Technical Specification Requirements and Alternate Control Rod Position Monitoring Requirements ML21326A0992022-01-0707 January 2022 Issuance of Amendment No. 281 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-177642-P-A, Revision 1 ML21262A0012021-11-0909 November 2021 Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate ML21284A0062021-10-29029 October 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-04 and IR-4-02 ML21227A0002021-10-0505 October 2021 Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident ML21222A2302021-09-0909 September 2021 Issuance of Amendment No. 343 Revision to Technical Specifications for Steam Generator Inspection Frequency (L-2020-LLA-0227) ML21174A0202021-08-0202 August 2021 Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML21167A3552021-07-16016 July 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-06 ML21167A2112021-06-30030 June 2021 Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0081 Through L-2020-LLR-0088) ML21075A0452021-03-26026 March 2021 Request to Utilize Code Case N-885 ML21043A1622021-03-25025 March 2021 Issuance of Amendment No. 278 Regarding Revision to Battery Surveillance Requirements ML21026A1422021-02-23023 February 2021 Issuance of Amendment No. 342 Revision to Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20312A0022020-12-10010 December 2020 Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 Through L-2020-LLR-0032) ML20312A0012020-12-10010 December 2020 Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML20287A4712020-10-20020 October 2020 Proposed Alternative RR-05-05 to the Requirements of the ASME Code Containment Unbonded Post-Tensioning System Inservice Inspection Requirements ML20275A0002020-10-14014 October 2020 Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections ML20237H9952020-09-29029 September 2020 Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program ML20252A0072020-09-15015 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 ML20191A0042020-08-0707 August 2020 Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity ML20189A2062020-07-16016 July 2020 Relief Request IR-4-03 Concerning Non-Code Methodology to Demonstrate Structural Integrity of Class 3 Moderate-Energy Piping ML20161A0002020-07-15015 July 2020 Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML20140A3692020-06-24024 June 2020 Issuance of Amendment No. 339 Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time ML20080K5082020-03-24024 March 2020 Alternative Request RR-05-03 for the Fifth 10-Year Inservice Inspection Interval ML19340A0252020-01-30030 January 2020 Issuance of Amendment No. 337 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components of Nuclear Power Reactors ML19305D2482019-12-31031 December 2019 Issuance of Amendments Adoption of Emergency Action Level Schemes Per NEI 99-01, Rev. 6 ML19340A0012019-12-18018 December 2019 Proposed Alternative Request IR-4-01 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML19338G3722019-12-18018 December 2019 Alternative Requests RR-05-01 and RR-05-02 for the Fifth 10-Year Inservice Inspection Interval ML19340A0002019-12-13013 December 2019 Relief Request IR-3-39, Proposed Alternative to ASME Code Weld Preheat Requirements ML19126A0002019-05-28028 May 2019 Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins ML19042A2772019-03-21021 March 2019 Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML18290A6022018-11-13013 November 2018 Alternative Requests Related to the Fifth and Fourth 10-Year Interval Pump, Valve, and Snubber Inservice Testing Programs, Respectively (EPID L-2018-LLR- 0012 Through EPID L-2018-LLR-0022) ML18275A0122018-10-0404 October 2018 Alternative Request P-06 for the 'C' Charging Pump Test Frequency ML18246A0072018-09-25025 September 2018 Issuance of Amendment No. 335 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML18252A0032018-09-17017 September 2018 Alternative Requests RR-04-27 and IR-3-38 for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography 2024-06-04
[Table view] Category:Technical Specifications
MONTHYEARML24128A2772024-06-0404 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) ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML21280A3282021-10-0707 October 2021 Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions ML20275A0002020-10-14014 October 2020 Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections ML20237H9952020-09-29029 September 2020 Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program ML20261H5982020-09-17017 September 2020 Response to Request for Additional Information for License Amendment Request to Revise Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20191A0042020-08-0707 August 2020 Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity ML19126A0002019-05-28028 May 2019 Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage ML19042A2772019-03-21021 March 2019 Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML18100A0552018-04-0404 April 2018 License Amendment Request to Revise Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML17038A0062017-02-0101 February 2017 Technical Specifications 2016 Annual Report ML17025A2182017-01-24024 January 2017 Issuance of Amendment Realistic Large Break Loss-of-Coolant Accident Analysis ML16354A4242016-12-14014 December 2016 License Amendment Request to Revise Technical Specification Surveillance Requirement 4.1.3.1.2 for Control Element Assembly 39 for the Remainder of Cycle 24 ML16208A0512016-07-26026 July 2016 Clean Technical Specifications (TS) Pages for Millstone Power Station 1 (Docket No. 05000245), Administrative Changes for TS (L53121) ML16068A3122016-03-31031 March 2016 Issuance of Amendment No. 326 to Revise Technical Specifications for Containment Leak Rate Testing ML16011A4002016-01-29029 January 2016 Issuance of Amendments Nos. 325 and 267 to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation ML15288A0042015-11-30030 November 2015 Issuance of Amendment No. 266: Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times ML15280A2422015-10-29029 October 2015 Issuance of Amendment No. 324 Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, Adoption of TSTF-425,Rev 3 ML15245A4822015-10-0707 October 2015 Millstone Power Station, Surry Power Station, and North Anna Power Station - Issuance of Amendments to Revise the Cyber Security Milestone 8 Completion Date in the Renewed Facility Operating Licenses ML15225A0102015-08-28028 August 2015 Issuance of Amendment No. 264 Surveillance Requirement 4.4.4.2, Reactor Coolant System Relief Valves ML15225A0082015-08-26026 August 2015 Issuance of Amendment No. 322 Revision to the Final Safety Analysis Report - Examination Requirements for ANSI B31.1.0 Piping Welds ML15187A3262015-07-29029 July 2015 Issuance of Amendment No. 321, Adoption of TSTF-426, Revision 5, Revise or Add Actions to Preclude Entry Into LCO 3.0.3 - RITSTF Initiatives 6B & 6C ML15187A1862015-07-28028 July 2015 Issuance of Amendment No. 263, Regarding Technical Specification Changes to Auxiliary Feedwater System (MF4692) ML15187A0112015-07-27027 July 2015 Issuance of Amendment No. 262, Refueling Water Storage Tank Allowable Temperatures ML15209A7292015-07-21021 July 2015 Response to Second Request for Additional Information Regarding Proposed Technical Specification Change for Spent Fuel Storage ML15098A0342015-05-20020 May 2015 Issuance of Amendment No. 261, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15093A0022015-05-20020 May 2015 Issuance of Amendment No. 320, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15093A4412015-05-18018 May 2015 Issuance of Amendment No. 319, Use of Areva M5 Alloy Clad Fuel Assemblies ML15044A0442015-02-0909 February 2015 Revised Response to Request for Additional Information Regarding License Amendment Request for Proposed Technical Specification Changes of the Refueling Water Storage Tank Allowable Temperatures ML15021A1282015-01-15015 January 2015 Proposed License Amendment Requests to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation ML14301A1122014-10-22022 October 2014 License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3 ML14154A0912014-05-28028 May 2014 Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specification 3/4.7.5, Ultimate Heat Sink. ML14073A0552014-04-0808 April 2014 Issuance of Amendment Calculated Containment Internal Pressure ML13346A7672013-12-18018 December 2013 Correction Letter to License Amendment Nos. 256 and 257 ML13155A1402013-05-28028 May 2013 Response to Request for Additional Information Regarding Proposed Technical Specifications Changes for Spent Fuel Storage ML13133A0322013-05-0303 May 2013 License Amendment Request for Changes to Technical Specification 3/4.7.5 Ultimate Heat Sink. ML13133A0332013-05-0303 May 2013 License Amendment Request for Changes to Technical Specification 3/4.7.11, Ultimate Heat Sink. ML12284A2132012-10-0404 October 2012 License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3 ML12250A6642012-08-28028 August 2012 Correction to Proposed Technical Specifications to Adopt TSTF-510 ML12220A0122012-07-31031 July 2012 Proposed Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process ML12219A0732012-07-31031 July 2012 Proposed Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process ML12202A0402012-07-17017 July 2012 License Amendment Request Regarding Removal of a License Condition, Proposed Technical Specifications Bases Change and FSAR Change for Ultimate Heat Sink Temperature Measurement ML12110A1142012-04-12012 April 2012 License Amendment Request for Permanent Alternate Repair Criteria for Steam Generator Tube Inspection and Repair ML12097A2002012-04-0202 April 2012 License Amendment Request to Revise Snubber Surveillance Requirements ML12032A2242012-01-25025 January 2012 License Amendment Request to Relocate TS Surveillance Frequencies to License Controlled Program in Accordance with TSTF-425, Revision 3 ML11329A0032011-11-17017 November 2011 Proposed License Amendment and Exemption Request for Use of Optimized Zirlo Rod Cladding ML11193A2242011-07-0505 July 2011 License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3, Attachment 3 Through Attachment 6, Marked-up TS Pages and TS Bases Pages ML11193A2232011-07-0505 July 2011 License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3, Attachment 1 Through Attachment 3, Marked-up TS Page Changes 3/4 6-22 ML1102802082011-01-20020 January 2011 License Amendment Request to Revise Technical Specification (TS) 6.8.4.g, Steam Generator (SG) Program, and TS 6.9.1.7, Steam Generator Tube Inspection Report for Temporary Alternate Repair Criteria (H*) 2024-06-04
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 April 8, 2014 SUBJECT: MILLSTONE POWER STATION, UNIT NO. 3-ISSUANCE OF AMENDMENT REGARDING CALCULATED CONTAINMENT INTERNAL PRESSURE (TAC NO. MF1731) Dear Mr. Heacock: The Commission has issued the enclosed Amendment No. 259 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3. This amendment is in response to your application dated April 25, 2013, as supplemented by letters dated September 19, and December 11, 2013. The amendment revises the Technical Specification Section 6.8.4.f, "Containment Leakage Rate Testing Program," to change the peak calculated containment internal pressure for the design basis loss-of-coolant accident. The Amendment approves an increase in the value of the peak calculated containment internal pressure, Pa, from 41.4 pounds per square inch gage (psig) to 41.9 psig. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-423 Sincerely, Mohan C. Thadani, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Enclosures: 1. Amendment No. 259 to NPF-49 2. Safety Evaluation cc w/encls: Distribution via Listserv UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT. INC .. ET AL. DOCKET NO. 50-423 MILLSTONE POWER STATION. UNIT NO.3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. NPF-49 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Dominion Nuclear Connecticut, Inc. et. al., dated April 25, 2013, as supplemented by letters dated September 19, and December 11, 2013 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
-2 -2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 259, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. DNC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance. Attachment: Changes to the License and Technical Specifications Date of Issuance: Apr i 1 8, 2014 FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 259 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove 4 4 Replace the following page of the Appendix A Technical Specification, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove 6-17 6-17
-4-(2) Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No.259 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DNC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) DNC shall not take any action that would cause Dominion Resources, Inc. (DR I) or its parent companies to void, cancel, or diminish DNC=s commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3. (4) Immediately after the transfer of interests in MPS Unit No. 3 to DNC, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC would then hold, be at a level no less than the formula amount under 10 CFR 50.75. (5) The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC is effected and thereafter is subject to the following: (a) The decommissioning trust agreement must be in a form acceptable to the NRC. (b) With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Resources, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited. Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.
(c) The decommissiong trust agreement for MPS Unit No. 3 must provide that no disbursement or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC. (d) The decommissioning trust agreements must provide that the agreement can not be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation. Renewed License No. NPF-49 Amendment No. 259 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and 3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable. f. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54( o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Performance Based Option of 10 CFR Part 50 Appendix J": The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013. The peak calculated containment internal pressure for the design basis loss of coolant accident, P a, is 41.9 psig. The maximum allowable containment leakage rate La, at P a' shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Leakage rate acceptance criteria are: 1) Containment overall leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are< 0.60 La for the combined Type Band Type C tests, 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and< 0.75 La for Type A tests; 2) Air lock testing acceptance criteria are: a. Overall air lock leakage rate is 0.05 La when tested at ;::: P a* b. For each door, seal leakage rate is < 0.01 La when pressurized to;::: Pa. The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.
- An exemption to Appendix J, Option A, paragraph III.D.2(b )(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985. MILLSTONE -UNIT 3 6-17 Amendment No. @, -l-84, 259 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 259 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 1.0 INTRODUCTION DOMINION NUCLEAR CONNECTICUT. INC. MILLSTONE POWER STATION. UNIT NO.3 DOCKET NO. 50-423 By letter dated April 25, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13120A 158), Dominion Nuclear Connecticut, Inc. (DNC or the licensee) proposed a change to Technical Specifications (TSs) for Millstone Power Station, Unit No. 3 (MPS3). The proposed change would revise the TS Section 6.8.4.f, "Containment Leakage Rate Testing Program" to increase the value of the peak calculated containment internal pressure for the design basis loss-of-coolant accident, P 8, from 41.4 pounds per square inch gage (psig) to 41.9 psig. According to the licensee, this increase is needed to address an increase in the calculated mass and energy (M&E) release during the blowdown phase of the design basis Loss-of-Coolant Accident (LOCA). The licensee provided responses to two sets of requests for additional information (RAis) from the U.S. Nuclear Regulatory Commission (NRC) staff related to this license amendment request (LAR). By a letter dated September 19, 2013, the licensee provided supplemental information in response to the NRC staff's first RAI (ADAMS Accession No. ML 13275A240). By a letter dated December 11, 2013, the licensee provided supplemental information in response to the NRC staff's second RAI (ADAMS Accession No. ML 13353A296). The supplemental letter of September 19 and December 11, 2013 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 25, 2013 (78 FR 38081). 2.0 REGULATORY EVALUATION The regulations at Title 10 of the Code of Federal Regulations ( 1 0 CFR) Part 50 Appendix A, General Design Criteria (GDC) 16 and 50 address the capability of the containment to withstand the containment pressure resulting from a postulated design basis LOCA.
-2 -Criterion 16, Containment Design, requires that the reactor containment and associated systems shall be provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. Criterion 19, Control Room, specifies that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in safe shutdown under accident conditions, including LOCAs, and that adequate radiation protection shall be provided. Criterion 38, Containment Heat Removal, specifies that a system to remove heat from the reactor containment shall be provided that rapidly reduces, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintains them at acceptably low levels. Criterion 50, Containment Design Basis, specifies that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartment can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a design basis LOCA. The regulations at 10 CFR Part 50, Appendix J Option 8 define calculated peak containment internal pressure as the calculated peak containment internal pressure related to the design basis LOCA as specified in the TS and specify the requirements for containment leakage rate testing. The requirements of MPS3 TS 6.8.4.f, "Containment Leakage Rate Testing" provide more detailed containment leakage rate testing requirements. As additional background, the NRC staff has previously issued license amendments to a number of reactor units to implement increased Pa values when revised containment analyses were performed for reasons such as correcting the calculation input errors in power uprates. For instance, the licensee cited Palisades Nuclear Plant as precedent for a similar license amendment approved by the NRC on January 19, 2012 (ADAMS Accession Nos. ML 113220370 and ML 120600415). The regulations at 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," require, in part, that licensees establish programs to qualify electric equipment important to safety located in a harsh environment. Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants," describes a method acceptable to the NRC staff for complying with 10 CFR 50.49 with regard to qualification of electric equipment important to safety for service in nuclear power plants to ensure that the equipment can perform its safety function during and after a design basis accident.
-3 -3.0 TECHNICAL EVALUATION Westinghouse provided the licensee with the mass and energy release data, which is used as an input to the LOCA containment response analysis of record (AOR), and is the basis for the current value of P a* The licensee used these mass and energy release values to calculate P a using the DOM-NAF-3-0.0-P-A GOTHIC (GOTHIC) containment analysis methodology previously approved by the NRC. This methodology is described in MPS3's Final Safety Analysis Report (FSAR) Chapter 6. The AOR and the revised analysis performed in support of this LAR utilize the same version of GOTHIC. In its April 25, 2013, letter the licensee stated that it had identified four errors in the MPS3 FSAR Chapter 6 analyses for large break LOCA M&E releases. The M&E releases were calculated by Westinghouse and input to the MPS3 FSAR Chapter 6 containment response analyses that were performed by DNC. A total of six issues were identified in Westinghouse Nuclear Safety Advisory Letter (NSAL)-11-5, "Westinghouse LOCA Mass and Energy Release Calculation Issues," dated July 25, 2011, out of which three were determined to affect large break LOCA containment M&E responses. The fourth error was independent of NSAL-11-5 and specific to MPS3 (see Item 1 below). The licensee stated that the four errors applicable to the MPS3 LOCA M&E analysis are: 1. Lower steam generator (SG) primary side pressure was used in the AOR (948 pounds per square inch absolute (psia)), rather than the correct value of 984 psia. This error under predicts the initial stored energy in the four MPS3 SGs. This error was specific to the MPS3 analysis of record and was discovered independent of the issues identified in NSAL-11-5. 2. Reactor vessel model in the AOR did not include the appropriate reactor vessel metal mass available, which affects the amount of reactor vessel stored energy initially available in the M&E model. 3. Reactor vessel model did not include the appropriate reactor vessel metal mass in the reactor vessel barrel/baffle downcomer region, this impacts the initial energy stored within the reactor vessel. MPS3 is an upflow plant. 4. The break flow was initialized at a non-conservative SG secondary pressure condition. This input value determines the initial SG secondary side temperature and pressure used in the large break LOCA M&E release calculations. The licensee stated that the Westinghouse errors only affected large break LOCA M&E releases and that steam line break and small break LOCA M&E releases are unaffected. Westinghouse reanalyzed the large break LOCA M&E releases with the errors corrected and no other design input changes. The large break LOCA M&E analysis methods that were applied are consistent with those referenced in MPS3 FSAR Section 6.2.1.3 (see below).
- WCAP-8264-P-A, Revision 1, "Topical Report: Westinghouse Mass and Energy Release Data for Containment Design," August 1975.
-4-* WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design -March 1979 Version," May 1983 (Proprietary). The NRC staff reviewed the licensee's response in a letter dated September 19, 2013, to the Staff's RAI. The current TS Section 3.6.1.4 value for containment initial pressure is a range of 10.6 psia-14.0 psia. The staff concurs with the licensee's approach to use an initial containment pressure of 14.2 psia as it would result in a conservative value for peak containment pressure. Therefore, with respect to containment safety analyses, there is no need to change the limiting safety system setting contained in TS Section 3.6.1.4 "Containment Pressure." Through the licensee letter dated September 19, 2013, the staff was able to get confirmation from the licensee that there were no additional generic errors that affect the MPS3 large break LOCA containment M&E release analysis, including the EPITOME computer modeling error discovered by Westinghouse. The proposed value of Pa, which includes the corrected mass and energy release data and the most adverse analytical initial containment pressure, is 41.9 psig. The revised calculated Pa remains below the containment design pressure of 45 psig, as referenced in FSAR Sections 6.2.1 and 3.8.1. Thus, the increase in peak calculated containment internal pressure does not affect systems and components in containment because these are designed for a containment design pressure limit of 45 psig. Therefore the proposed change in the Pavalue from41.4 psig to 41.9 psig is acceptable. The maximum allowable containment leakage rate (La). at Pa, for MPS3 is 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the first 2 minutes post-LOCA, 1 00 percent of the containment leak rate is assumed to bypass the secondary containment and release unfiltered at ground. After the Supplementary Leak Collection and Release System (SLCRS) drawdown is effective at 2 minutes, the bypass leakage rate is 0.06 of La (or 0.018 percent by weight per day) as defined in TS 6.8.4.f. The remaining containment leakage is filtered and releases through SLCRS. The LAR further states that the containment leak rate, La, is reduced from 0.3 to 0.15 percent by weight at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for offsite dose calculations, and at one hour for control room and technical support center dose calculations. The licensee stated that the long-term LOCA containment response AOR demonstrates that containment pressure meets the radiological analysis requirement in the FSAR for a 50 percent reduction in containment leakage after one hour. As stated in the licensee's April 25, 2013, letter, the containment leakage rate "Type A" test is performed in accordance with the requirements of 10 CFR Part 50, Appendix J to demonstrate that leakage of systems and components penetrating the primary containment do not exceed the allowable leakage rates specified in MPS3 TS 6.8.4.f. Specifically, the Type A test verifies that the measured containment leakage rate at Pa does not exceed the maximum allowable leakage rate, La, which is used to calculate the dose consequences following a postulated LOCA. The most recent Type A test at MPS3 was completed on November 7, 2011. The containment pressure during the test was measured at 42.5 psig, which exceeds the proposed peak calculated containment internal pressure of 41.9 psig. The containment leakage rate during the
-5 -test was calculated to correspond to 0.0531 weight percent per day, which is significantly less than the maximum containment leakage rate of 0.30 weight percent of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as specified in TS 6.8.4.f and used in the offsite dose calculations. The licensee also stated in the September 19, 2013, letter that Appendix J Type B and C test procedures do not require revision upon approval of this proposed LAR. ANSI 56.8-1994 Section 3.3.2 requires that Type Band C testing be performed at a pressure not less than Pa (except for airlock door seals, which may have a lower pressure specified) and not more than 1.1 times Pa when a higher differential pressure results in increased sealing. MPS3 site procedures for Type B and C testing require that testing be performed within a range of pressures that, with the revised Pa, will continue to be within the range of pressures required by ANSI 56.8-1994. Therefore, the Type B and C test procedures will not require revision upon approval of the license amendment request. The LOCA offsite, control room, and technical support center dose calculations are based on the allowable leakage rate (i.e., 0.3 percent by weight of containment air) in the TSs. Based on the significant margin between the tested and allowed leakage rates exhibited by the most recent Type A test completed on November 7, 2011, the staff concludes that the proposed change to increase the peak calculated containment pressure, P a. from 41.4 psig to 41.9 psig will have no adverse impact on the dose calculations and the plant's compliance with GDC-19. The NRC staff concludes that the proposed change meets the requirements of 10 CFR Part 50 Appendix A, (1) GDC 16, because the licensee showed that the containment design conditions important to safety are not exceeded during a postulated Design-Basis Accident (DBA); (2) GDC 19, because the licensee showed that control room dose analysis is unaffected by the proposed change; (3) GDC 38, because the licensee showed that the containment heat removal system would reduce the containment pressure and temperature rapidly following a DBA and would maintain them at acceptable levels; and (4) GDC 50, because the licensee showed that the containment heat removal system is designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a design basis LOCA. Therefore, the proposed increase in the Pa, from 41.4 psig to 41.9 psig is acceptable. The licensee stated that the calculated peak containment internal pressure, Pa, would increase from 41.4 psig to 41.9 psig. This increase in Pais due to an increase in the calculated M&E released into containment during the blowdown phase of the design basis LOCA event. The licensee stated that it has reanalyzed MPS3's FSAR Chapter 6 containment analyses with corrected large break LOCA M&E data. The licensee also stated that the large break LOCA containment pressure analysis uses NRC-approved methods already described in the MPS3 FSAR. The licensee stated that the calculated peak containment internal pressure will remain below the design limit of 45 psig. In the LAR, Section 4, the licensee provided a review of the environmental qualification (EQ) of equipment in the containment addressing the effect of the increase in M&E on EQ. The licensee determined that the increase in Pa to 41.9 psig does not adversely affect environmentally qualified equipment within the containment because this equipment is qualified
-6-for the containment design pressure of 45 psig. The licensee also determined that the containment temperatures, using the corrected large break LOCA M&E releases, remain within the bounding containment temperature profile used to qualify the equipment and concluded that the post-accident operating time of the environmentally qualified equipment remains unaffected. By letter dated December 11, 2013, the licensee responded to the NRC staff's request for additional information regarding the EQ reanalysis to show equipment qualification inside containment meets the requirements of 10 CFR 50.49. The licensee stated that the EQ bounding temperature and pressure profiles taken from the MPS3 Environmental Specification are used for the EQ of plant equipment. The licensee also stated that the most limiting LOCA for peak containment pressure and temperature is a double-ended guillotine break of the hot leg. The licensee provided Figures 1 and 2, which compared the EQ bounding temperature and pressure profiles and containment temperature and pressure profiles (EQ overlays) from the double-ended hot leg break analysis. The NRC staff reviewed the figures and concluded that the EQ bounding temperature and pressure profiles envelope the containment temperature and pressure profiles from the double-ended hot leg break analysis respectively. The licensee further stated in the letter dated December 11, 2013, that the temperature and pressure margins in the MPS3 EQ program are applied in accordance with 10 CFR 50.49 and Regulatory Guide 1.89. Also, the licensee stated that these margins are in addition to the EQ bounding temperature and pressure profiles for equipment qualified according to 10 CFR 50.49. Since the EQ bounding temperature and pressure profiles exceed the expected peak containment temperature and pressure from the double-ended hot leg break analysis, and the margins are applied in addition to the EQ bounding temperature and pressure, the NRC staff concluded that there will be no adverse impact on the EQ of electric equipment for the revised temperature and pressure inside containment due to the proposed amendment request. Additionally, the licensee stated that the proposed change in calculated peak containment internal pressure for the design basis LOCA does not result in any adverse impact on any area of the plant outside containment, and as a result, environmental conditions outside containment remain bounded by existing calculations. Based on this information, the NRC staff concludes that there will be no adverse impact on the qualifications of electrical equipment for both temperature and pressure outside containment. Based on the technical evaluation provided above, the NRC staff concludes that the EQ bounding profiles for temperature and pressure are bounded by the EQ test profiles under harsh environmental conditions. The NRC staff concludes that the revised LOCA containment pressure and temperature profiles remain enveloped by the current MPS3 EQ profiles and therefore the electric equipment inside and outside containment remains environmentally qualified in accordance with 10 CFR 50.49. Therefore, the NRC staff concludes that the licensee's application for the amendment to increase peak calculated containment internal pressure P a, due to reanalyzed design basis LOCA is acceptable, because it continues to satisfy GDC 16, GDC 19, GDC 38, and GDC 50, and the potential impact on the environmental qualifications of electrical equipment under the newly created environmental conditions of increased mass and energy release is not significant, because the equipment continues to remain qualified in accordance with the requirements of 1 0 CFR 50.49.
-7 -4.0 STATE CONSULTATION In accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official had no comments. 5.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding as published in the Federal Register on June 25, 2013 (78 FR 38081). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment. 6.0 CONCLUSION The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: R. Torres S.Som Date: April 8, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 April 8, 2014 SUBJECT: MILLSTONE POWER STATION, UNIT NO.3 -ISSUANCE OF AMENDMENT REGARDING CALCULATED CONTAINMENT INTERNAL PRESSURE (TAC NO. MF1731) Dear Mr. Heacock: The Commission has issued the enclosed Amendment No. 259 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3. This amendment is in response to your application dated April 25, 2013, as supplemented by letters dated September 19, and December 11, 2013. The amendment revises the Technical Specification Section 6.8.4.f, "Containment Leakage Rate Testing Program," to change the peak calculated containment internal pressure for the design basis loss-of-coolant accident. The Amendment approves an increase in the value of the peak calculated containment internal pressure, P a, from 41.4 pounds per square inch gage (psig) to 41.9 psig. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-423 Sincerely, Ira/ Mohan C. Thadani, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Enclosures: 1. Amendment No. 259 to NPF-49 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC RidsAcrsAcnw_MaiiCTR Resource RidsNrrDorllpl1-1 Resource RidsNrrLAKGoldstein Resource ADAMS Accession NO* ML14073A055 .. LPLI-1 R/F RidsNrrDssStsb Resource RidsNrrDirsSCVB Resource RidsRgn1 MaiiCenter Resource *Via Memo ML14007A663 RidsNrrDeEEEB Resource RidsNrrDoriDpr Resource RidsNrrPMMillstone Resource **Via Memo ML13329A444 OFFICE NRR/LPLI-1/PM NRR/LPLI-1/LA OGC NRR//EEEB/BC NRR/SCVB/BC NRR/LPLI-1/BC NAME MThadani KGoldstein JWachutka JZimmerman* RDennina** BBeasley DATE 03/25/2014 03/21/2014 03/31/2014 01/27/2014 02/10/2014 04/08/2014 OFFICIAL RECORD COPY