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Category:Letter type:NL
MONTHYEARNL-24-0261, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20232024-07-19019 July 2024 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2023 NL-24-0281, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions2024-07-18018 July 2024 License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions NL-24-0227, Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2024-07-0303 July 2024 Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) NL-24-0234, Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown2024-06-28028 June 2024 Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown NL-24-0230, Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-3602024-06-28028 June 2024 Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-360 NL-24-0143, Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in2024-06-27027 June 2024 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in NL-24-0201, Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)2024-06-18018 June 2024 Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) NL-24-0238, RIPE License Amendment Request to Change Containment Air Temperature Actions, Supplemental Information2024-06-14014 June 2024 RIPE License Amendment Request to Change Containment Air Temperature Actions, Supplemental Information NL-24-0229, Notice of Intent to Pursue Subsequent License Renewal2024-06-0707 June 2024 Notice of Intent to Pursue Subsequent License Renewal NL-24-0202, SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations2024-05-24024 May 2024 SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations NL-24-0191, Annual Radiological Environmental Operating Reports for 20232024-05-10010 May 2024 Annual Radiological Environmental Operating Reports for 2023 NL-24-0064, Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.52024-05-0303 May 2024 Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5 NL-24-0188, Cycle 33 Core Operating Limits Report2024-05-0101 May 2024 Cycle 33 Core Operating Limits Report NL-24-0165, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20232024-04-25025 April 2024 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2023 NL-24-0137, RIPE License Amendment Request to Change Containment Air Temperature Actions2024-04-19019 April 2024 RIPE License Amendment Request to Change Containment Air Temperature Actions NL-24-0011, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2024-01-11011 January 2024 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis NL-23-0901, 30-Day 10 CFR 21 Notification - Framatome Supplied Siemens Medium Voltage (Mv) Circuit Breakers2023-12-15015 December 2023 30-Day 10 CFR 21 Notification - Framatome Supplied Siemens Medium Voltage (Mv) Circuit Breakers NL-23-0908, Cycle 30 Core Operating Limits Report2023-12-13013 December 2023 Cycle 30 Core Operating Limits Report NL-23-0877, Request for Exemption from Security Event Notification Implementation2023-11-29029 November 2023 Request for Exemption from Security Event Notification Implementation NL-23-0825, Reply to Notice of Violation EA-23-080 and Readiness for 95001 Inspection2023-11-14014 November 2023 Reply to Notice of Violation EA-23-080 and Readiness for 95001 Inspection NL-23-0739, Response to NRC Inspection Report and Preliminary White Finding2023-09-0808 September 2023 Response to NRC Inspection Report and Preliminary White Finding NL-23-0713, Response to Request for Additional Information Regarding Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit2023-08-23023 August 2023 Response to Request for Additional Information Regarding Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit NL-23-0716, Response to Request for Additional Information Regarding Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit2023-08-23023 August 2023 Response to Request for Additional Information Regarding Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit NL-23-0704, Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit2023-08-22022 August 2023 Emergency License Amendment Request: Technical Specification 3.6.5, Containment Air Temperature, One-Time Temporary Change to Limit NL-23-0658, Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal2023-08-11011 August 2023 Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal NL-23-0542, CFR 50.46 ECCS Evaluation Model Annual Report for 20222023-08-0909 August 2023 CFR 50.46 ECCS Evaluation Model Annual Report for 2022 NL-23-0624, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2023-08-0404 August 2023 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis NL-23-0628, Readiness for Supplemental Inspection EA-22-1012023-07-26026 July 2023 Readiness for Supplemental Inspection EA-22-101 NL-23-0566, ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use2023-07-13013 July 2023 ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use NL-23-0555, Request for Exemption from Physical Barrier Requirement2023-07-0707 July 2023 Request for Exemption from Physical Barrier Requirement NL-23-0506, to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications2023-07-0505 July 2023 to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC and CAOC Fq Surveillance Technical Specifications NL-23-0444, Quality Assurance Topical Report Submittal2023-06-15015 June 2023 Quality Assurance Topical Report Submittal NL-23-0457, ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use2023-06-12012 June 2023 ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use NL-23-0449, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application2023-06-0202 June 2023 National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application NL-23-0383, SNC Response to Regulatory Issue Summary 2023-01:Preparation And.2023-05-19019 May 2023 SNC Response to Regulatory Issue Summary 2023-01:Preparation And. NL-23-0372, Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 20222023-05-10010 May 2023 Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 2022 NL-23-0337, Response to Request for Additional Information Related to License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.52023-05-0505 May 2023 Response to Request for Additional Information Related to License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.5 NL-23-0295, Reply to a Notice of Violation; EA-22-1012023-05-0101 May 2023 Reply to a Notice of Violation; EA-22-101 NL-23-0310, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20222023-04-25025 April 2023 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2022 NL-23-0019, GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2023-04-12012 April 2023 GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI NL-23-0263, Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report2023-04-0505 April 2023 Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report NL-23-0014, Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding2023-03-29029 March 2023 Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding NL-23-0208, Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update2023-03-29029 March 2023 Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update NL-23-0213, Inservice Inspection Program Owner'S Activity Report for Outage 1R312023-03-21021 March 2023 Inservice Inspection Program Owner'S Activity Report for Outage 1R31 NL-23-0228, Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21)2023-03-20020 March 2023 Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21) NL-23-0080, Response to Request for Additional Information (RAI) Related to Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.02023-02-0202 February 2023 Response to Request for Additional Information (RAI) Related to Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0 NL-23-0008, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting.2023-01-17017 January 2023 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting. NL-23-0011, Update to Supporting Documentation Regulatory Conference EA-22-101. Cover Letter Only2023-01-0505 January 2023 Update to Supporting Documentation Regulatory Conference EA-22-101. Cover Letter Only NL-22-0799, License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.52022-12-20020 December 2022 License Amendment Request to Revise the Frequency of Surveillance Requirement 3.6.3.5 NL-22-0897, Supplement to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A,2022-12-0909 December 2022 Supplement to Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, 2024-07-03
[Table view] Category:Technical Specification
MONTHYEARNL-24-0230, Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-3602024-06-28028 June 2024 Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-360 NL-24-0064, Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.52024-05-0303 May 2024 Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5 ML23318A0702023-10-31031 October 2023 Technical Specifications Bases ML23076A0502023-03-21021 March 2023 Correction of Amendment Nos. 244 and 241 Issuance of Amendments Regarding Revision to Technical Specifications to Relocate Augmented Piping Inspection Program Details to a License Controlled Document NL-20-0170, Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC2022-10-14014 October 2022 Non-Voluntary License Amendment Request: Technical Specification Revision to Adopt WCAP-17661-P-A, Improved RAOC NL-22-0289, License Amendment Request to Revise Technical Specification 4.3 Fuel Storage to Correct Tabulated Values from the Associated Spent Fuel Pool (SFP) Criticality Analysis2022-09-21021 September 2022 License Amendment Request to Revise Technical Specification 4.3 Fuel Storage to Correct Tabulated Values from the Associated Spent Fuel Pool (SFP) Criticality Analysis NL-21-0925, Adoption of TSTF-269-A, Revision 2, Allow Administrative Means of Position Verification for Locked or Sealed Valves2021-12-22022 December 2021 Adoption of TSTF-269-A, Revision 2, Allow Administrative Means of Position Verification for Locked or Sealed Valves NL-21-1017, License Amendment Request to Revise Technical Specification 5.5.17, Containment Leakage Rate Testing Program to Increase Calculated Peak Containment Pressure2021-12-13013 December 2021 License Amendment Request to Revise Technical Specification 5.5.17, Containment Leakage Rate Testing Program to Increase Calculated Peak Containment Pressure NL-21-0984, Southern Nuclear Operating Company - Second Supplement to License Amendment Request to Revise Technical Specification 5.7, High Radiation Area, Administrative Controls2021-11-10010 November 2021 Southern Nuclear Operating Company - Second Supplement to License Amendment Request to Revise Technical Specification 5.7, High Radiation Area, Administrative Controls ML21313A3272021-10-28028 October 2021 0, Technical Specification Bases ML21313A3262021-10-28028 October 2021 0 to Updated Final Safety Analysis Report, Chapter 16, Technical Specifications NL-21-0613, Supplement to License Amendment Request to Revise Technical Specification 5.7, High Radiation Area, .2021-06-30030 June 2021 Supplement to License Amendment Request to Revise Technical Specification 5.7, High Radiation Area, . NL-21-0160, License Amendment Request to Remove Table of Contents from Technical Specifications2021-06-22022 June 2021 License Amendment Request to Remove Table of Contents from Technical Specifications ML20125A2122020-04-28028 April 2020 Revisions to Technical Specification Bases Changes NL-19-0331, License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors2019-12-12012 December 2019 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors NL-19-1063, Units 1 and 2 and Vogtle Electric Generating Plant, Units 1 and 2 - Application to Revise Technical Specifications to Adopt TSTF-569, Revise Response Time Testing Definition2019-12-10010 December 2019 Units 1 and 2 and Vogtle Electric Generating Plant, Units 1 and 2 - Application to Revise Technical Specifications to Adopt TSTF-569, Revise Response Time Testing Definition NL-19-0006, Application to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program2019-07-15015 July 2019 Application to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program NL-19-0771, Risk Informed Technical Specification Information Only Bases Changes2019-06-27027 June 2019 Risk Informed Technical Specification Information Only Bases Changes NL-19-0221, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications, SNC Response to NRC Request for Additional Information (RAI)2019-05-0303 May 2019 Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications, SNC Response to NRC Request for Additional Information (RAI) ML18312A0772018-10-30030 October 2018 Technical Specifications Bases ML18226A0942018-08-0909 August 2018 Edwin I. Hatch, Units 1 and 2; and Vogtle, Units 1 and 2 - License Amendment Request to Revise Technical Specification 5.2.2.g and Update Emergency Plan Minimum On-shift Staff Tables NL-17-2095, Supplement Submittal of Corrected Technical Specification Definition of Iodine 1-131 for the Ast. TSTF-448 and TSTF-312 Amendment Package2017-12-18018 December 2017 Supplement Submittal of Corrected Technical Specification Definition of Iodine 1-131 for the Ast. TSTF-448 and TSTF-312 Amendment Package NL-17-0534, Joseph M. Farley Nuclear Plant, Technical Specifications Bases2017-04-20020 April 2017 Joseph M. Farley Nuclear Plant, Technical Specifications Bases ML17117A3742017-04-20020 April 2017 Technical Specifications Bases NL-16-1026, Units 1 & 2, Vogtle Electric Generating Plant- Units 1 & 2 - Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2016-11-21021 November 2016 Units 1 & 2, Vogtle Electric Generating Plant- Units 1 & 2 - Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements NL-16-0091, License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application.2016-07-28028 July 2016 License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application. NL-15-1761, Response to Request for Additional Information Regarding Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294)2015-09-17017 September 2015 Response to Request for Additional Information Regarding Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294) NL-15-0421, J. M. Farley, Units 1 and 2 - License Amendment Request to Revise Technical Specifications Regarding Generic Letter 2008-01, Managing Gas Accumulation in Accordance with TSTF-523, Revision 2, Using the Consolidated Line Item Improvement P2015-05-12012 May 2015 J. M. Farley, Units 1 and 2 - License Amendment Request to Revise Technical Specifications Regarding Generic Letter 2008-01, Managing Gas Accumulation in Accordance with TSTF-523, Revision 2, Using the Consolidated Line Item Improvement Pro NL-14-1531, License Amendment Request for Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294), Using the Consolidated Line Item Improvement Process2015-04-13013 April 2015 License Amendment Request for Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294), Using the Consolidated Line Item Improvement Process ML14335A6292014-11-24024 November 2014 Basis for Proposed Changes. Part 2 of 2 NL-14-1385, Basis for Proposed Changes. Part 1 of 22014-11-24024 November 2014 Basis for Proposed Changes. Part 1 of 2 NL-14-0635, License Amendment Request to Revise Technical Specifications Reactor Trip System Instrumentation2014-06-0303 June 2014 License Amendment Request to Revise Technical Specifications Reactor Trip System Instrumentation NL-13-0081, Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement...2013-01-23023 January 2013 Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement... NL-12-0403, Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST)2012-02-28028 February 2012 Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST) NL-12-0346, Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST)2012-02-20020 February 2012 Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST) NL-11-0299, Application to Amend Surveillance Requirement 3.4.11.1 and 3.4.11.42012-01-18018 January 2012 Application to Amend Surveillance Requirement 3.4.11.1 and 3.4.11.4 NL-11-1057, License Amendment Request for Technical Specification Table 3.3.1-12011-09-0909 September 2011 License Amendment Request for Technical Specification Table 3.3.1-1 NL-10-0272, License Amendment Request for Adoption of TSTF-425-A, Revision 3, Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Using the Consolidated..2010-10-29029 October 2010 License Amendment Request for Adoption of TSTF-425-A, Revision 3, Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Using the Consolidated.. NL-09-1644, Emergency Technical Specification (TS) Revision Request Regarding TS 3.7.8, Service Water System (SWS)2009-10-0808 October 2009 Emergency Technical Specification (TS) Revision Request Regarding TS 3.7.8, Service Water System (SWS) NL-09-0154, Request to Revise Technical Specifications to Delete Reactor Trip System, Function 11, Reactor Coolant Pump Breaker Position2009-03-30030 March 2009 Request to Revise Technical Specifications to Delete Reactor Trip System, Function 11, Reactor Coolant Pump Breaker Position NL-08-0535, Application for Technical Specification Change TSTF-374, Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil Using Consolidated Line Item Improvement Process2008-10-0808 October 2008 Application for Technical Specification Change TSTF-374, Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil Using Consolidated Line Item Improvement Process NL-08-1519, Response to Request for Information Regarding Revision to Technical Specifications 3.3.1, 3.3.2, 3.3.6, 3.3.7, and 3.3.82008-10-0808 October 2008 Response to Request for Information Regarding Revision to Technical Specifications 3.3.1, 3.3.2, 3.3.6, 3.3.7, and 3.3.8 ML0611504462006-04-14014 April 2006 Technical Specifications, Revise TS Section 5.5.6, Pre-Stressed Concrete Containment Tendon Surveillance Program, for Consistency with the Requirements of 10 CFR 50.55a(g)(4) for Components Classified as Code Class CC NL-05-0856, Request to Revise Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process2005-11-0202 November 2005 Request to Revise Technical Specifications Regarding Mode Change Limitations Using the Consolidated Line Item Improvement Process NL-05-0068, Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using Astrum2005-10-0606 October 2005 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using Astrum NL-05-0740, Submittal of Joseph M. Farley Nuclear Plant, Technical Specifications Revision Spent Fuel Cask Loading Requirements2005-05-17017 May 2005 Submittal of Joseph M. Farley Nuclear Plant, Technical Specifications Revision Spent Fuel Cask Loading Requirements ML0512300672005-04-22022 April 2005 Enclosure, Correction Ltr. on Amendment Nos. 167 & 159 for Elimination of Hydrogen Recombiners and Monitors (Tac No. MC3224, MC3225) ML0507303042005-03-0808 March 2005 Enclosure, Amendment 168 & 160, TS Revision on 5.6 Reporting Requirements NL-04-0564, Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, Application for Technical Specification Amendment to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors2004-05-21021 May 2004 Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, Application for Technical Specification Amendment to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors NL-03-2454, Exigent Technical Specification Revision Request Regarding Electrical Power Systems AC Sources - Operating2003-12-0303 December 2003 Exigent Technical Specification Revision Request Regarding Electrical Power Systems AC Sources - Operating 2024-06-28
[Table view] Category:Bases Change
MONTHYEARNL-24-0230, Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-3602024-06-28028 June 2024 Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-360 ML23318A0702023-10-31031 October 2023 Technical Specifications Bases ML21313A3262021-10-28028 October 2021 0 to Updated Final Safety Analysis Report, Chapter 16, Technical Specifications ML21313A3272021-10-28028 October 2021 0, Technical Specification Bases ML20125A2122020-04-28028 April 2020 Revisions to Technical Specification Bases Changes NL-19-0331, License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors2019-12-12012 December 2019 License Amendment Request: Revise Measurement Units Associated with Reactor Trip System Nuclear Instrument Comparison Surveillance and Trip Setpoint for Control Room Air Intake Radiation Monitors NL-19-0006, Application to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program2019-07-15015 July 2019 Application to Revise Technical Specifications to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program NL-19-0771, Risk Informed Technical Specification Information Only Bases Changes2019-06-27027 June 2019 Risk Informed Technical Specification Information Only Bases Changes ML18312A0772018-10-30030 October 2018 Technical Specifications Bases ML17117A3742017-04-20020 April 2017 Technical Specifications Bases NL-17-0534, Joseph M. Farley Nuclear Plant, Technical Specifications Bases2017-04-20020 April 2017 Joseph M. Farley Nuclear Plant, Technical Specifications Bases NL-16-0091, License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application.2016-07-28028 July 2016 License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application. NL-15-1761, Response to Request for Additional Information Regarding Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294)2015-09-17017 September 2015 Response to Request for Additional Information Regarding Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294) NL-14-1531, License Amendment Request for Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294), Using the Consolidated Line Item Improvement Process2015-04-13013 April 2015 License Amendment Request for Adoption of TSTF-432-A, Rev. 1, Change in Technical Specifications End States (WCAP-16294), Using the Consolidated Line Item Improvement Process NL-14-1385, Basis for Proposed Changes. Part 1 of 22014-11-24024 November 2014 Basis for Proposed Changes. Part 1 of 2 ML14335A6292014-11-24024 November 2014 Basis for Proposed Changes. Part 2 of 2 NL-13-0081, Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement...2013-01-23023 January 2013 Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement... NL-12-0403, Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST)2012-02-28028 February 2012 Exigent Technical Specification Revision Request for TS 3.5.4 Refueling Water Storage Tank (RWST) NL-11-0299, Application to Amend Surveillance Requirement 3.4.11.1 and 3.4.11.42012-01-18018 January 2012 Application to Amend Surveillance Requirement 3.4.11.1 and 3.4.11.4 NL-10-0272, License Amendment Request for Adoption of TSTF-425-A, Revision 3, Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Using the Consolidated..2010-10-29029 October 2010 License Amendment Request for Adoption of TSTF-425-A, Revision 3, Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Using the Consolidated.. NL-05-0068, Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using Astrum2005-10-0606 October 2005 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using Astrum NL-04-0564, Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, Application for Technical Specification Amendment to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors2004-05-21021 May 2004 Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, Application for Technical Specification Amendment to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors NL-03-1511, Joseph A. Farley, Amendment Request to Revise Technical Specifications Control Room Emergency Filtration/Pressurization System and Penetration Room Filtration System, Revision 22003-07-16016 July 2003 Joseph A. Farley, Amendment Request to Revise Technical Specifications Control Room Emergency Filtration/Pressurization System and Penetration Room Filtration System, Revision 2 NL-03-0605, Request for Technical Specification Changes Unit Two Pressurizer Power Operated Relief Valve Block Valve Surveillance Testing2003-03-31031 March 2003 Request for Technical Specification Changes Unit Two Pressurizer Power Operated Relief Valve Block Valve Surveillance Testing NL-03-0372, Technical Specifications Revision Request Removal of Reference to Plant Operations Review Committee2003-03-21021 March 2003 Technical Specifications Revision Request Removal of Reference to Plant Operations Review Committee ML0234505732002-12-0303 December 2002 Transmittal of Revision 19 of the Technical Specifications Bases ML0203701742002-01-10010 January 2002 Revision 11 to Technical Specification Bases 2024-06-28
[Table view] |
Text
1.M. Stinson (Mike) Southern Nuclear Vice President Operating Company, Inc.
40 lnverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 SOUTHERN-October 6 , 2005 COMPANY Energy t o Serve YourWorldm Docket Nos.: 50-348 NL-05-0068 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analvsis Using ASTRUM In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is proposing a change to the Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2 Technical Specifications (TS). This change is to support a revision to the Best Estimate Loss of Coolant Accident (LOCA) for FNP. The NRC recently approved a new Westinghouse Best Estimate LOCA (BELOCA) methodology, ASTRUM.
ASTRUM (Automated Statistical Treatment of Uncertainty Method) was submitted in WCAP-16009-P. The NRC issued a Safety Evaluation Report in a letter dated November 5,2004. Westinghouse issued WCAP-16009-P-A in January 2005. SNC has completed the analysis for FNP and the enclosed proposed amendment is to incorporate a reference to WCAP-16009-P-A in TS section 5.6.5 Core Operating Limits Report (COLR).
Enclosure 1 provides the basis for the proposed change, including an evaluation determining that the proposed change involves no significant hazards consideration as defined in 10 CFR 50.92 and an evaluation that determines this change satisfies the criteria of 10 CFR 5 1.22 for categorical exclusion from the requirements for an environmental assessment. Marked-up TS page is provided in Enclosures 2, and clean-typed pages are provided in Enclosure 3.
SNC requests approval of the proposed license amendments by October 14,2006. The proposed changes would be implemented within 60 days of issuance of the amendment.
Upon approval of this proposed license amendment, the results presented in this letter will become the large break LOCA analysis of record for FNP Units 1 and 2.
(Affirmation and signature are provided on the following page.)
U. S. Nuclear Regulatory Commission NL-05 -0068 Page 2 Mr. L. M. Stinson states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.
This letter contains no NRC commitments. If you have any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY S j q n to and subscribed before me this 6 day of &f 0 be r ,2005.
-
Notary Public dqdP-My commission expires:
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MY COMMISSlON-E Jmac 10, M)pm&D'IRIIUmrrARYHJBwC~~
Enclosures:
- 1. Basis for Proposed Change
- 2. Marked-Up Technical Specifications Page
- 3. Clean Typed Technical Specifications Pages cc: Southern Nuclear Operating Companv Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, General Manager - Plant Farley RTYPE: CFA04.054; LC# 14206 U. S. Nuclear Rermlatorv Commission Dr. W. D. Travers, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley Mr. C. A. Patterson, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer
Joseph M. Farley Nuclear Plant Units 1 and 2 Technical SpecificationAmendment Request to Incorporate Best Estimate LOCA Analysis Using ASTRUM Enclosure 1 Basis for Proposed Change
Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using ASTRUM Enclosure 1 Basis for Proposed Change In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is proposing a change to the Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2 Technical Specifications (TS). This change is to support a revision to the Best Estimate Loss of Coolant Accident (LOCA) for FNP. The NRC recently approved a new Westinghouse Best Estimate LOCA (BELOCA) methodology, ASTRUM. ASTRUM (Automated Statistical Treatment of Uncertainty
- Method) was submitted in WCAP-16009-P. The NRC issued a Safety Evaluation Report in a letter dated November 5,2004. Westinghouse issued WCAP-16009-P-A in January 2005. SNC has completed the analysis for FNP and the enclosed proposed amendment is to incorporate a reference to WCAP-16009-P-A in TS section 5.6.5 Core Operating;Limits Report (COLR).
2.0 Proposed Chan~e The current FNP Technical Specification section 5.6.5, Core Operating Limits Report (COLR), contains references to the analytical methods used to determine the core operating limits as follows:
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 (W Proprietary).
(Methodology for LCOs 3.1.1 - SHUTDOWN MARGIN, 3.1.3 -
Moderator Temperature Coefficient, 3.1.5 - Shutdown Bank Insertion Limit, 3.1.6 - Control Bank Insertion Limits, 3.2.3 - Axial Flux Difference, 3.2.1 - Heat Flux Hot Channel Factor, 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor and 3.9.1 - Boron Concentration)
- 2. WCAP- 10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control / FQSurveillance Technical Specification," February 1994 (W Proprietary).
(Methodology for LCOs 3.2.3 - Axial Flux Difference and 3.2.1 - Heat Flux Hot Channel Factor.)
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 2 of 14 3a. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (W Proprietary).
3b. WCAP- 12610-P-A, "Vantage+ Fuel Assembly Reference Core Report,"
April 1995 (W Proprietary).
(Methodology for LC0 3.2.1 - Heat Flux Hot Channel Factor and LC0 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
- 4. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower AT and Thermal Overtemperature AT Trip Functions)
- 5. WCAP-14750-P-A Revision 1, "RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop PWRs. (Westinghouse Proprietary)
(Methodology for minimum RCS flow determination using the elbow tap measurement.)
- 6. WCAP-11596-P-A, "Qualification of the Phoenix-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Methodology for LC0 3.9.1 - Boron Concentration.)
- 7. WCAP-11397-P-A "Revised Thermal Design Procedure," April 1989 (Methodology for LC0 2.1.1-Reactor Core Safety Limits, LC0 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
This proposed amendment will add the following item:
3c. WCAP- 16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," M.E. Nissley, et al., January 2005 (Proprietary).
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 3 of 14 3.0 Background Farley Units 1 and 2 are currently operating under the W O B M R A C best estimate methodology approved by the NRC in 1998. This is reflected in the current COLR Reference in TS 5.6.5.b.3a. The NRC recently approved a new Westinghouse BELOCA methodology, ASTRUM. ASTRUM was submitted in WCAP-16009-P (ref. 3). The NRC issued a Safety Evaluation Report in a letter dated November 5,2004 (ref. 4). Westinghouse issued WCAP-16009-P-A in January 2005 (ref. 5).
Westinghouse recently underwent a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is still based on the Code Qualification Document (CQD) methodology (ref. 2) and follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology. However, the uncertainty analysis (element 3 in CSAU) is replaced by a technique based on order statistics.
The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case.
4.0 Technical Analvsis Westinghouse reanalyzed the FNP BELOCA using the new approved ASTRUM methodology. The reanalysis was performed and met all the NRC Safety Evaluation Report conditions and limitations identified in NRC letter dated November 5,2004 (ref. 4).
The WCOBRA/TRAC models for Farley Units 1 and 2 were originally developed for the power uprate which was approved by the NRC in 1998 (ref. 1). Two BELOCA models were utilized in the original analysis of record (AOR), mainly because Unit 1 had an upflow barrelbaffle (B/B) configuration, whereas Unit 2 had a downflow B/B configuration. A parametric study was performed at that time to determine the limiting unit. Unit 2 was determined to be the limiting unit at that time. Therefore, the Unit 2 model was utilized for the subsequent steps of the original application of the best estimate large break LOCA evaluation model.
Subsequent to the original analysis, FNP performed a Replacement Steam Generators (RSG) project for both units, and Unit 2 has been converted to an upflow B/B configuration. These changes were incorporated into the ASTRUM analysis. Moreover, investigations revealed that the remaining differences in the vessels were small enough to justify the use of a single W C O B M R A C geometric model for both Units 1 and 2.
Table 1 lists the major plant parameter assumptions used in the BE LOCA analysis for FNP and Table 2 summarizes the results of the ASTRUM analysis.
The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95' percentile of the Peak Clad Temperature (PCT),
Local Maximum Oxidation (LMO), and Core Wide Oxidation (CWO) with 95%
BE LOCA with ASTRLTM NL-05-0068 Enclosure 1 Page 4 of 14 confidence level. These parameters are needed to satisfy the 10 CFR 50.46 criteria with regard to PCT, LMO, and CWO. From these 124 calculations, run 104 proved to be the limiting PCT transient, run 05 1 the limiting LMO transient, and run 014 the limiting CWO transient.
The scatter plot presented on Figure 2 shows the influence of the effective break area on the final PCT. The effective break area is calculated by multiplying the discharge coefficient (CD) with the sample value of the break area, normalized to the cold-leg cross sectional area. Figure 2 is provided to illustrate that the break area is a significant contributor to the variation in PCT.
Figures 3,4 and 5 are presented to show the limiting cladding transient for each criterion. Figure 3 shows the predicted clad temperature transient at the PCT limiting elevation for run 104. Figure 4 presents the clad temperature transient predicted at the LMO elevation for run 051. Figure 5 shows the PCT trace for the CWO limiting transient (run 014).
Based on the results as presented in Table 2, it is concluded that the FNP Units 1 and 2 continue to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.
5.0 Regulatorv Analvsis 5.1 10 CFR 50.46 Evaluation In accordance with 10 CFR 50.46, the conclusions of the best estimate large break LOCA analysis show that there is a high level probability the following criteria are met.
- 1. The calculated maximum fuel element cladding temperature (i.e.,
peak cladding temperature (PCT)) will not exceed 2,200°F.
- 2. The calculated total oxidation of the cladding (i.e., maximum cladding oxidation) will nowhere exceed 0.17 times the total cladding thickness before oxidation.
3 The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam (i.e.,
maximum hydrogen generation) will not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
- 4. The calculated changes in core geometry are such that the core remains amenable to cooling.
- 5. After successful initial operation of the Emergency Core Cooling System (ECCS), the core temperature will be maintained at an acceptably low value and decay heat will be removed for the
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 5 of 14 extended period of time required by the long-lived radioactivity remaining in the core.
5.2 No Significant Hazards Consideration In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is proposing a change to the Joseph M.
Farley Nuclear Plant (FNP) Unit 1 and Unit 2 Technical Specifications (TS). This change is to support a revision to the Best Estimate Loss of Coolant Accident (LOCA) for FNP. The NRC recently approved a new Westinghouse Best Estimate LOCA (BELOCA) methodology, ASTRUM. ASTRUM automated Statistical Treatment of Uncertainty Method) was submitted in WCAP-16009-P. The NRC issued a Safety Evaluation Report in a letter dated November 5,2004. Westinghouse issued WCAP-16009-P-A in January 2005. SNC has completed the analysis for FNP and the enclosed proposed amendment is to incorporate a reference to WCAP-16009-P-A in TS section 5.6.5 Core Operating Limits Revort (COLRZ SNC has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
No physical plant changes are being made as a result of using the Westinghouse Best Estimate Large Break LOCA (BELOCA) analysis methodology. The proposed TS changes simply involve updating the references in TS 5.6.5.b, Core Operating;Limits Revort (COLR), to reference the Westinghouse BELOCA analysis methodology. The plant conditions assumed in the analysis are bounded by the design conditions for all equipment in the plant; therefore, there will be no increase in the probability of a LOCA.
The consequences of a LOCA are not being increased, since the analysis has shown that the Emergency Core Cooling System (ECCS) is designed such that its calculated cooling performance conforms to the criteria contained in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors." No other accident consequence is potentially affected by this change.
All systems will continue to be operated in accordance with current design requirements under the new analysis, therefore no new components or system interactions have been identified that could lead to an increase in the probability of any accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR). No changes were required to the Reactor Protection System (RPS) or Engineering Safety Features (ESF) setpoints because of the new analysis methodology.
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 6 of 14 Therefore, it is concluded that this change does not significantly increase the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
There are no physical changes being made to the plant as a result of using the Westinghouse Best Estimate Large Break LOCA analysis methodology. No new modes of plant operation are being introduced. The configuration, operation and accident response of the structures or components are unchanged by utilization of the new analysis methodology. Analyses of transient events have confirmed that no transient event results in a new sequence of events that could lead to a new accident scenario. The parameters assumed in the analysis are within the design limits of existing plant equipment.
In addition, employing the Westinghouse Best Estimate Large Break LOCA analysis methodology does not create any new failure modes that could lead to a different kind of accident. The design of all systems remains unchanged and no new equipment or systems have been installed which could potentially introduce new failure modes or accident sequences. No changes have been made to any RPS or ESF actuation setpoints.
Based on this review, it is concluded that no new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes.
Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
It has been shown that the analytic technique used in the Westinghouse Best Estimate Large Break LOCA analysis methodology realistically describes the expected behavior of the reactor system during a postulated LOCA. Uncertainties have been accounted for as required by 10 CFR 50.46. A sufficient number of LOCAs with different break sizes, different locations, and other variations in properties have been considered to provide assurance that the most severe postulated LOCAs have been evaluated. The analysis has demonstrated that all acceptance criteria contained in 10 CFR 50.46 paragraph b continue to be satisfied.
Therefore, it is concluded that this change does not involve a significant reduction in the margin of safety.
Based on the above, SNC concludes that the proposed change presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
6.0 Environmental Consideration SNC has reviewed the proposed change pursuant to 10 CFR 50.92 and determined that it does not involve a significant hazards consideration. In addition, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite and there is no significant increase in individual or cumulative occupational radiation exposure.
Consequently, the proposed TS change has no significant effect on the human environment and satisfies the criteria of 10 CFR 5 1.22 for categorical exclusion from the requirements for an environmental assessment.
7.0 References
- 1. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 137 to Facility Operating License No. NPF-2 and Amendment No. 129 to Facility Operating License No. NPF-8 Southern Nuclear Operating Company, Inc., Et. Al., Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364, April 29,1998.
- 2. Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).
- 3. Nissley, M. E., et. al., 2003, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP- 16009-P.
- 4. Letter from H. N. Berkow (NRC) to J. A. Gresham (W) dated November 5, 2004, RE: Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," TAC No.
MB9483.
- 5. Nissley, M. E., et. al., January 2005, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP- 16009-P-A
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 8 of 14 Table 1 Major Plant Parameter Assumptions Used in the BE LOCA Analysis Parameter Value Plant Physical Description SG Tube Plugging Plant Initial Operating Conditions Reactor Power 5 102 % of 2,775 MWt FQ5 2.5 Peaking Factors FdH5 1.7 Axial Power Distribution See Figure 1 Fluid Conditions TAVG 567.2 + 6 OF 5 TAVGj 577.2 -+ 6 OF Pressurizer Pressure 2,200 psia 5 PRCsI 2,300 psia Reactor Coolant Flow L 86,000 gprnlloop Accumulator Temperature 90 OF 5 TACC 5 120 OF Accumulator Pressure 600 psia 5 PACC5 680 psia Accumulator Water Volume 965 ft3 5 VAcC5 995 ft3 Accident Boundary Conditions Single Failure Assumptions Loss of one ECCS train Safety Injection Flow Minimum Safety Injection Temperature 70 OF 5 TsI 5 100 OF 5 12 sec (with offsite power)
Safety Injection Initiation Delay Time 5 27 sec (without offsite power)
Containment Pressure Bounded
Table 2 Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement Value Criteria 95/95 PCT ("F) 1,836. 2,200 95/95 LMO (%) 2.9 17.0 95/95 cwo (%) 0.22 1.oo Core remains Coolable Geometry coolable Core remains cool Long Term Cooling in long term PCT - Peak Clad Temperature LMO - Local Maximum Oxidation CWO - Core Wide Oxidation Separate from the ASTRUM methodology, an evaluation was performed to assess the ECCS performance during the quarterly Residual Heat Removal (RHR) surveillance testing. The assessment concluded that a 25 O F PCT penalty applies to the licensing basis PCT during the testing period. This value will be tracked as a temporary PCT and will apply only during the period of the test.
BE LOCA with ASTRLM NL-05-0068 Enclosure 1 Page 10 of 14 Figure 1 Farley BELOCA Analysis Axial Power Shape Operating Space Envelope PBOT: integrated power fraction in the lower third of the core PMID: integrated power fraction in the middle third of the core
BE LOCA with ASTRLTM NL-05-0068Enclosure 1 Page 11 of 14 Figure 2 Farley BELOCA Analysis PCT v. Effective Break Area Scatter Plot 0 0 PCT-DEG 0 0 0 PCT DEGCL [deg F]
X X PCT-SPL 0 0 0 PCT S P L I T [deg F]
800 - X
-
-
-
I l l 1 I I I I I l l 1 I l l I I I I I I 600 I I I I I 0 .5 1 1.5 2 25 CD
BE LOCA with ASTRUM NL-05-0068Enclosure 1 Page 12 of 14 Figure 3 Farley Units 112 BELOCA Analysis Clad Temperature Transient at the Limiting Elevation for the Limiting PCT Case (Run 104)
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Paee 13 of 14 Figure 4 Farley Units 1/2 BELOCA Analysis Clad Temperature Transient at the Limiting Elevation for the Limiting LMO Case (Run051)
BE LOCA with ASTRUM NL-05-0068 Enclosure 1 Page 14 of 14 Figure 5 Farley Units Y2 BELOCA Analysis PCT Transient for the Limiting CWO Case (Run 014)
Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using ASTRUM Enclosure 2 Marked-Up Technical Specifications Page
Reporting Requirements 5.6 5.6 Reporting Requirements CORE OPERATING LIMITS REPORT (COLR) (continued) 3a. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (W Proprietary).
3b. WCAP-12610-P-A, "Vantage+ Fuel Assembly Reference Core Report," April 1995 & l Proprietary).
(Methodology for LC0 3.2.1 - Heat Flux Hot Channel Factor and LC0 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary)
I (Methodology for Overpower AT and Thermal Overtemperature AT Trip Functions)
I 5. WCAP-14750-P-A Revision 1, "RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop PWRs. (Westinghouse Proprietary)
I (Methodology for minimum RCS flow determination using the elbow tap measurement.)
II 6. WCAP-11596-P-A, "Qualification of the Phoenix-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Methodology for LC0 3.9.1 - Boron Concentration.)
I 7. WCAP-11397-P-A "Revised 'Thermal Design Procedure," April 1989 (Methodology for LC0 2.1 .l-Reactor Core Safety Limits, LC0 3.4.1 -
RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
II
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
3c. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using 1
Automated Statistical Treatment of Uncertainty Method (AS-TRUM)," M.E. Nissley, et al., January 2005 (Proprietary).
(continued)
Farley Units 1 and 2 Amendment No. 151 (Unit 1)
Amendment No. 143 (Unit 2)
Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Amendment Request to Incorporate Best Estimate LOCA Analysis Using ASTRUM Enclosure 3 Clean Typed Technical Specifications Pages Affected Pages
Reporting Requirements 5.6 5.6 Re~ortinaReauirements CORE OPERATING LIMITS REPORT (COLR) (continued) 3a. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (W Proprietary).
3b. WCAP-12610-P-A, "Vantage+ Fuel Assembly Reference Core Report," April 1995 (W Proprietary).
(Methodology for LC0 3.2.1 - Heat Flux Hot Channel Factor and LC0 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
3c. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" M.E. Nissley, et al., January 2005 (Proprietary).
- 4. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower AT and Thermal Overtemperature AT Trip Functions)
- 5. WCAP-14750-P-A Revision 1, "RCS Flow Verification Using Elbow Taps at Westinghouse %Loop PWRs. (Westinghouse Proprietary)
(Methodology for minimum RCS flow determination using the elbow tap measurement.)
- 6. WCAP-11596-P-A, "Qualification of the Phoenix-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Methodology for LC0 3.9.1 - Boron Concentration.)
- 7. WCAP-11397-P-A "Revised Thermal Design Procedure," April 1989 (Methodology for LC0 2.1.1 -Reactor Core Safety Limits, LC0 3.4.1 -
RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
(continued)
Farley Units 1 and 2 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING I-IMITS REPORT (COLR) (continued)
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant Svstem (RCS) PRESSLIRE AND TEMPERATURE LIMITS REPORT [PTLR)
- a. The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LC0 3.4.3.
- b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated March 31, 1998 and April 3, 1998.
- c. The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.
5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures shall be reported within 30 days.
Reports on EDG failures shall include a description of the failures, underlying causes, and corrective actions taken per the Emergency Diesel Generator Reliability Monitoring Program.
5.6.8 PAM Report When a report is required by Condition B or G of LC0 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Farley Units 1 and 2 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Reporting Requirements 5.6 5.6 Reportina Reauirements 5.6.9 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.
5.6.1 0 Steam Generator (SG) Tube Inspection R e ~ o r t A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
- a. The scope of inspections performed on each SG,
- b. Active degradation mechanisms found,
- c. Nondestructive examination techniques utilized for each degradation mechanism,
- d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e. Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f. Total number and percentage of tubes plugged to date, and
- g. The results of condition monitoring, including the results of tube pulls and in-situ testing.
5.6.1 1 Alternate AC ( M C ) Source Out of Service Report The NRC shall be notified if the M C source is out of service for greater than 10 days.
Farley Units 1 and 2 Amendment No.
Amendment No.
(Unit 1)
(Unit 2) I