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Category:Letter type:JAFP
MONTHYEARJAFP-23-0065, License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications2023-12-14014 December 2023 License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications JAFP-23-0069, Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements JAFP-23-0057, and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation JAFP-23-0064, Emergency Plan Document Revision2023-11-15015 November 2023 Emergency Plan Document Revision JAFP-23-0063, Registration of Spent Fuel Cask Use2023-11-13013 November 2023 Registration of Spent Fuel Cask Use JAFP-23-0059, Registration of Spent Fuel Cask Use2023-10-24024 October 2023 Registration of Spent Fuel Cask Use JAFP-23-0048, Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-31031 August 2023 Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0050, Physical Security Plan, Revision 242023-08-31031 August 2023 Physical Security Plan, Revision 24 JAFP-23-0047, Correction to the 2022 Annual Radioactive Effluent Release Report2023-08-30030 August 2023 Correction to the 2022 Annual Radioactive Effluent Release Report JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0043, 10 CFR 50.46 Annual Report2023-07-31031 July 2023 10 CFR 50.46 Annual Report JAFP-23-0038, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance2023-07-28028 July 2023 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance JAFP-23-0033, License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation2023-06-28028 June 2023 License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation JAFP-23-0025, 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023 2022 Annual Radiological Environmental Operating Report JAFP-23-0023, 2022 Annual Radioactive Effluent Release Report2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report JAFP-23-0010, 2022 REIRS Transmittal of NRC Form 52023-03-20020 March 2023 2022 REIRS Transmittal of NRC Form 5 JAFP-23-0008, Supplement to Inservice Inspection Summary Report Cycle 252023-02-22022 February 2023 Supplement to Inservice Inspection Summary Report Cycle 25 JAFP-22-0053, Inservice Inspection Summary Report Cycle 252022-12-20020 December 2022 Inservice Inspection Summary Report Cycle 25 JAFP-22-0046, Core Operating Limits Report Cycle 262022-10-17017 October 2022 Core Operating Limits Report Cycle 26 JAFP-22-0040, 10 CFR 50.46 Annual Report2022-07-29029 July 2022 10 CFR 50.46 Annual Report JAFP-22-0033, Core Operating Limits Report Mid-Cycle 252022-06-23023 June 2022 Core Operating Limits Report Mid-Cycle 25 JAFP-22-0032, Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b..2022-06-16016 June 2022 Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.. JAFP-22-0030, Oswego County and New York State Participation in the Emergency Plan2022-05-13013 May 2022 Oswego County and New York State Participation in the Emergency Plan JAFP-22-0029, 2021 Annual Radiological Environmental Operating Report2022-05-11011 May 2022 2021 Annual Radiological Environmental Operating Report JAFP-22-0028, 2021 Annual Radioactive Effluent Release Report2022-04-27027 April 2022 2021 Annual Radioactive Effluent Release Report JAFP-22-0026, 2021 REIRS Transmittal of NRC Form 52022-04-0707 April 2022 2021 REIRS Transmittal of NRC Form 5 JAFP-22-2020, Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Info2022-03-0404 March 2022 Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Informed JAFP-22-0017, Amendments to Indemnity Agreements2022-02-15015 February 2022 Amendments to Indemnity Agreements JAFP-22-0007, Summary of Changes to Exelon Generation Company, LLC, Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-102022-01-31031 January 2022 Summary of Changes to Exelon Generation Company, LLC, Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10 JAFP-22-0010, Supplemental Information in Response to Order Consenting to License Transfers and Approval of Draft Conforming License Amendments2022-01-24024 January 2022 Supplemental Information in Response to Order Consenting to License Transfers and Approval of Draft Conforming License Amendments JAFP-22-0008, Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary2022-01-14014 January 2022 Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary JAFP-21-0093, Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation2021-10-18018 October 2021 Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation JAFP-21-0089, Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2021-09-27027 September 2021 Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position JAFP-21-0087, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-09-16016 September 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0083, Notification of Readiness for NRC 95001 Inspection2021-09-0909 September 2021 Notification of Readiness for NRC 95001 Inspection JAFP-21-0081, Supplement to Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements2021-09-0303 September 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements JAFP-21-0075, Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs2021-08-12012 August 2021 Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs JAFP-21-0073, Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance.2021-08-0909 August 2021 Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance. JAFP-21-0069, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2021-07-30030 July 2021 Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors JAFP-21-0070, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b2021-07-30030 July 2021 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b JAFP-21-0071, 10 CFR 50.46 Annual Report2021-07-29029 July 2021 10 CFR 50.46 Annual Report JAFP-21-0064, Supplemental Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting2021-07-0707 July 2021 Supplemental Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting JAFP-21-0053, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-14014 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0052, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0051, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0050, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-12012 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0044, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-06-11011 June 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0042, Reply to a Notice of Violation; EA-20-1382021-06-0303 June 2021 Reply to a Notice of Violation; EA-20-138 JAFP-21-0041, Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing2021-05-17017 May 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing JAFP-21-0040, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2021-05-14014 May 2021 Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs 2023-08-31
[Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50
MONTHYEARJAFP-23-0065, License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications2023-12-14014 December 2023 License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications JAFP-23-0048, Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-31031 August 2023 Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0038, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance2023-07-28028 July 2023 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance JAFP-23-0033, License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation2023-06-28028 June 2023 License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling RS-22-004, Supplement to Application to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements2022-01-0404 January 2022 Supplement to Application to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements ML21277A1932021-11-16016 November 2021 Enclosure 2, Draft Conforming License Amendments ML21181A0572021-11-0404 November 2021 Issuance of Amendment No. 344 Adoption of TSTF-545 JAFP-21-0093, Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation2021-10-18018 October 2021 Propose Change to Eliminate Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation JAFP-21-0089, Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2021-09-27027 September 2021 Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position JAFP-21-0069, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2021-07-30030 July 2021 Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors JAFP-21-0053, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-14014 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0051, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0050, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-12012 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement ML21144A1072021-05-17017 May 2021 Acceptance of Requested Licensing Action License Amendment Request to Adopt TSTF-582 JAFP-21-0041, Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing2021-05-17017 May 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing JAFP-21-0040, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2021-05-14014 May 2021 Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs RS-21-039, Supplemental Information Regarding Application for Order Approving Transfers and Proposed Conforming License Amendments2021-03-25025 March 2021 Supplemental Information Regarding Application for Order Approving Transfers and Proposed Conforming License Amendments ML21057A2732021-02-25025 February 2021 Application for Order Approving License Transfers and Proposed Conforming License Amendments JAFP-20-0080, Application to Revise Surveillance Requirement (SR) 3.5.1.62020-12-11011 December 2020 Application to Revise Surveillance Requirement (SR) 3.5.1.6 JAFP-20-0083, Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule2020-12-11011 December 2020 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing JAFP-20-0047, License Amendment Request: Application to Modify Technical Specifications 3.6.1.3, Primary Containment Isolation Valves2020-06-30030 June 2020 License Amendment Request: Application to Modify Technical Specifications 3.6.1.3, Primary Containment Isolation Valves JAFP-20-0046, License Amendment Request for Adoption of TSTF-478-A, Revision 2, BWR Technical Specification Changes That Implement the Revised Rule for Combustible Gas Control, Using the Consolidated Line Item.2020-06-30030 June 2020 License Amendment Request for Adoption of TSTF-478-A, Revision 2, BWR Technical Specification Changes That Implement the Revised Rule for Combustible Gas Control, Using the Consolidated Line Item. RS-20-025, Supplement to Exelon Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2020-02-24024 February 2020 Supplement to Exelon Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls JAFP-20-0012, License Amendment Request - Application to Adopt Technical Specification Task Force Traveler TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, Using the Consolidated Line...2020-01-23023 January 2020 License Amendment Request - Application to Adopt Technical Specification Task Force Traveler TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, Using the Consolidated Line... ML19295G7832019-12-19019 December 2019 Issuance of Amendment No. 331 Regarding Change to Technical Specifications to Remove Ultimate Heat Sink Bar Rack Heaters ML19277D5522019-10-0404 October 2019 Draft James A. FitzPatrick Nuclear Power Plant - Emergency License Amendment Request - One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source ML19277G0852019-10-0404 October 2019 Emergency License Amendment Request: One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source ML19276G3132019-10-0202 October 2019 Draft Exigent License Amendment Request - One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source JAFP-19-0092, Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers2019-09-26026 September 2019 Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers JAFP-19-0089, License Amendment Request - Proposed Changes to the Technical Specifications Related to Primary Containment Hydrodynamic Loads2019-09-12012 September 2019 License Amendment Request - Proposed Changes to the Technical Specifications Related to Primary Containment Hydrodynamic Loads JAFP-19-0083, License Amendment Request - Proposed Change to the Technical Specifications to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation2019-09-0505 September 2019 License Amendment Request - Proposed Change to the Technical Specifications to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation RS-19-044, License Amendment Request - Deletion of Facility Operating License Conditions Related to Decommissioning Trust Provisions2019-08-28028 August 2019 License Amendment Request - Deletion of Facility Operating License Conditions Related to Decommissioning Trust Provisions ML19157A2032019-07-11011 July 2019 Issuance of Amendment No. 325 Reactivity Anomalies Surveillance JAFP-19-0067, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2019-06-27027 June 2019 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-19-040, Application to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program2019-06-25025 June 2019 Application to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program RS-19-038, Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR2019-03-0707 March 2019 Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR JAFP-19-0033, License Amendment Request - Revise Technical Specifications to Adopt TSTF-522, Revise Ventilation System Surveillance Requirements to Operate for 10 Hours Per Month, Revision 02019-03-0707 March 2019 License Amendment Request - Revise Technical Specifications to Adopt TSTF-522, Revise Ventilation System Surveillance Requirements to Operate for 10 Hours Per Month, Revision 0 RS-19-003, Application to Adopt TSTF-564, Safety Limit Mcpr.2019-02-0101 February 2019 Application to Adopt TSTF-564, Safety Limit Mcpr. JAFP-18-0106, License Amendment Request: Proposed Change to the Technical Specifications to Remove Ultimate Heat Sink Bar Rack Heaters from Technical Specifications2018-11-29029 November 2018 License Amendment Request: Proposed Change to the Technical Specifications to Remove Ultimate Heat Sink Bar Rack Heaters from Technical Specifications RS-18-130, Application to Revise Technical Specifications to Adopt TSTF-439, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO2018-11-0101 November 2018 Application to Revise Technical Specifications to Adopt TSTF-439, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO JAFP-18-0093, License Amendment Request - Reactivity Anomalies Surveillance2018-10-0202 October 2018 License Amendment Request - Reactivity Anomalies Surveillance ML18292A5282018-09-25025 September 2018 Technical Evaluation Input for License Amendment Request to Remove Requirements for Deicing Heaters from Technical Specification 3.7.2 RS-18-020, Application to Revise Technical Specifications to Modify the APRM Channel Adjustment Surveillance Requirement2018-06-15015 June 2018 Application to Revise Technical Specifications to Modify the APRM Channel Adjustment Surveillance Requirement JAFP-18-0048, James A. Fitzpatrick, Unit 1, License Amendment Request - Safety Limit Minimum Critical Power Ration Change2018-05-17017 May 2018 James A. Fitzpatrick, Unit 1, License Amendment Request - Safety Limit Minimum Critical Power Ration Change RS-17-161, Fleet License Amendment Request to Relocate Technical Specification Unit/Facility/Plant Staff Qualification ANSI N18.1-1971 and ANSl/ANS-3.1-1978 Requirements to the Exelon Quality Assurance Topical Report (QATR)2018-03-0101 March 2018 Fleet License Amendment Request to Relocate Technical Specification Unit/Facility/Plant Staff Qualification ANSI N18.1-1971 and ANSl/ANS-3.1-1978 Requirements to the Exelon Quality Assurance Topical Report (QATR) JAFP-18-0014, James A. FitzPatrick Nuclear Power Plant - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors2018-01-31031 January 2018 James A. FitzPatrick Nuclear Power Plant - License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors JAFP-18-0014, Attachment 1: License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors2018-01-31031 January 2018 Attachment 1: License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors ML18037A7872018-01-31031 January 2018 Attachment 1: License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 2023-08-31
[Table view] Category:Technical Specification
MONTHYEARJAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling ML23095A3722023-04-0505 April 2023 2023 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes Transmittal ML23095A3742023-04-0505 April 2023 Enclosure 2: Technical Specification (TS) Bases Revision 58 JAFP-22-2020, Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Info2022-03-0404 March 2022 Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Informed JAFP-21-0070, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b2021-07-30030 July 2021 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b JAFP-21-0053, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-14014 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0052, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0051, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement ML21095A2972021-04-0505 April 2021 Revised Technical Specifications Bases JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability ML19277G0852019-10-0404 October 2019 Emergency License Amendment Request: One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source ML19276G3132019-10-0202 October 2019 Draft Exigent License Amendment Request - One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source JAFP-19-0092, Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers2019-09-26026 September 2019 Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers JAFP-19-0089, License Amendment Request - Proposed Changes to the Technical Specifications Related to Primary Containment Hydrodynamic Loads2019-09-12012 September 2019 License Amendment Request - Proposed Changes to the Technical Specifications Related to Primary Containment Hydrodynamic Loads JAFP-19-0067, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2019-06-27027 June 2019 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls RS-19-038, Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR2019-03-0707 March 2019 Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR JAFP-17-0093, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22017-10-0202 October 2017 Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 ML17104A0122017-04-10010 April 2017 Submittal of 2017 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes JAFP-17-0030, Submittal of 2017 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes2017-04-10010 April 2017 Submittal of 2017 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes JAFP-15-0047, 2015 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes Transmittal2015-04-0808 April 2015 2015 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes Transmittal JAFP-11-0102, Application for Change to the Current Licensing Basis, Authorizing Use of on Load Tap Changers with the Reserve Station Service Transformers2011-08-16016 August 2011 Application for Change to the Current Licensing Basis, Authorizing Use of on Load Tap Changers with the Reserve Station Service Transformers JAFP-10-0079, Application for Amendment to Modify the Technical Specifications Requirements for Testing of Low Pressure Coolant Injection Inverter2010-06-25025 June 2010 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Low Pressure Coolant Injection Inverter JAFP-10-0050, Proposed Change to the Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio2010-04-21021 April 2010 Proposed Change to the Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio JAFP-09-0086, Application for Amendment for James A. FitzPatrick Nuclear Power Plant to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function2009-07-31031 July 2009 Application for Amendment for James A. FitzPatrick Nuclear Power Plant to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function JAFP-09-0087, Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves2009-07-31031 July 2009 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves ENOC-09-00009, Entergy Nuclear Operations, Inc. - License Amendment Request; Proposed Adoption of TSTF-511 Regarding Elimination of Technical Specification Work Hour Restrictions2009-04-27027 April 2009 Entergy Nuclear Operations, Inc. - License Amendment Request; Proposed Adoption of TSTF-511 Regarding Elimination of Technical Specification Work Hour Restrictions JAFP-08-0098, Supplement to the Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent...2008-09-24024 September 2008 Supplement to the Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent... ML0816900942008-08-0707 August 2008 Technical Specification JAFP-08-0011, License Amendment Request Application for Technical Specification Changes Using the Consolidated Line- Item Improvement Process (CLIIP)2008-02-0707 February 2008 License Amendment Request Application for Technical Specification Changes Using the Consolidated Line- Item Improvement Process (CLIIP) JAFP-08-0006, Application for Amendment to Technical Specifications Regarding Diesel Fuel Oil, Lube Oil, and Starting Air Consistent with TSTF-501 and Diesel Fuel Oil Testing Program Consistent with TSTF-3742008-01-22022 January 2008 Application for Amendment to Technical Specifications Regarding Diesel Fuel Oil, Lube Oil, and Starting Air Consistent with TSTF-501 and Diesel Fuel Oil Testing Program Consistent with TSTF-374 JAFP-07-0091, Application for Technical Specifications Change Based on TSTF-477, Add Action Statement for Two Inoperable Control Room Air Conditioning Subsystems, Using the Consolidated Line Item Improvement Process (CLIIP)2007-07-25025 July 2007 Application for Technical Specifications Change Based on TSTF-477, Add Action Statement for Two Inoperable Control Room Air Conditioning Subsystems, Using the Consolidated Line Item Improvement Process (CLIIP) JAFP-07-0088, James A. FitzPatrick, Application for Amendment to Technical Specifications Regarding Control Room Envelope Habitability, Consistent with TSTF-448 Revision 32007-07-17017 July 2007 James A. FitzPatrick, Application for Amendment to Technical Specifications Regarding Control Room Envelope Habitability, Consistent with TSTF-448 Revision 3 JAFP-05-0084, Application for Technical Specifications (TS) Change Regarding Mode Change Limitations (TSTF-359) Using the Consolidated Line Item Improvement Process (CLIIP)2005-05-31031 May 2005 Application for Technical Specifications (TS) Change Regarding Mode Change Limitations (TSTF-359) Using the Consolidated Line Item Improvement Process (CLIIP) JAFP-04-0083, Proposed License Amendment Re Safety Limit Minimum Critical Power Ratio2004-06-0404 June 2004 Proposed License Amendment Re Safety Limit Minimum Critical Power Ratio JAFP-03-0108, Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension2003-07-28028 July 2003 Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension ML0235000102002-12-0606 December 2002 Proposed Change to TS Regarding Local Power Range Monitor Calibration Frequency (JPTS-02-002) ML0222606432002-08-13013 August 2002 Technical Specifications for Amendment No. 275, Changes to the Technical Specification Leakage Limit for the Main Steam Isolation Valves from an Individual to an Aggregate Limit (Tac No. MB3333) JAFP-02-0146, Updated Markup and Retyped Technical Specification Pages for Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling...2002-07-17017 July 2002 Updated Markup and Retyped Technical Specification Pages for Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling... ML0219700842002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Table of Contents ML0219700282002-07-0303 July 2002 Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Attachment Table a - Administrative Changes Matrix ML0219700712002-07-0303 July 2002 Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Attachment 4, Table L - Less Restrictive Changes Matrix ML0219700952002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), 3.6 Containment Systems ML0219701022002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B 3.7 Plant Systems ML0219701382002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Table of Contents ML0219701422002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B 3.3 Instrumentation ML0219700922002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B. 3.4 Reactor Coolant System (RCS) JPN-02-016, Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling Dose Consequences2002-06-0707 June 2002 Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling Dose Consequences ML18031A0412000-04-14014 April 2000 Issuance of Amendment 261 Changes to the Technical Specifications Regarding the Allowed Containment Leakage Rate 2023-08-03
[Table view] Category:Amendment
MONTHYEARRS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling JAFP-22-2020, Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Info2022-03-0404 March 2022 Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10CFR 50.69, Risk-Informed JAFP-21-0070, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b2021-07-30030 July 2021 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b JAFP-21-0053, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-14014 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0052, Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request, One Time Extension of Condition a to TS 3.5.1, TS 3.6.1.9, and TS 3.6.4.1 Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-21-0051, Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement2021-06-13013 June 2021 Emergency License Amendment Request - One Time Extension to TS 3.5.1 Condition a, TS 3.6.1.9 Condition a, and TS 3.6.4.1 Condition a Completion Time to Support Residual Heat Removal (RHR) Pump Motor Replacement JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability ML19277G0852019-10-0404 October 2019 Emergency License Amendment Request: One Time Extension to the TS 3.8.1 Action A.3 Completion Time for an Inoperable Offsite Source JAFP-19-0092, Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers2019-09-26026 September 2019 Application to Adopt TSTF-372, Addition of LCO 3.0.8, Inoperability of Snubbers RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls RS-19-038, Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR2019-03-0707 March 2019 Supplement to the Application to Adopt TSTF-564, Safety Limit MCPR JAFP-10-0079, Application for Amendment to Modify the Technical Specifications Requirements for Testing of Low Pressure Coolant Injection Inverter2010-06-25025 June 2010 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Low Pressure Coolant Injection Inverter JAFP-10-0050, Proposed Change to the Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio2010-04-21021 April 2010 Proposed Change to the Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio JAFP-09-0086, Application for Amendment for James A. FitzPatrick Nuclear Power Plant to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function2009-07-31031 July 2009 Application for Amendment for James A. FitzPatrick Nuclear Power Plant to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function JAFP-09-0087, Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves2009-07-31031 July 2009 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves ENOC-09-00009, Entergy Nuclear Operations, Inc. - License Amendment Request; Proposed Adoption of TSTF-511 Regarding Elimination of Technical Specification Work Hour Restrictions2009-04-27027 April 2009 Entergy Nuclear Operations, Inc. - License Amendment Request; Proposed Adoption of TSTF-511 Regarding Elimination of Technical Specification Work Hour Restrictions JAFP-08-0098, Supplement to the Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent...2008-09-24024 September 2008 Supplement to the Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent... ML0816900942008-08-0707 August 2008 Technical Specification JAFP-08-0011, License Amendment Request Application for Technical Specification Changes Using the Consolidated Line- Item Improvement Process (CLIIP)2008-02-0707 February 2008 License Amendment Request Application for Technical Specification Changes Using the Consolidated Line- Item Improvement Process (CLIIP) JAFP-07-0091, Application for Technical Specifications Change Based on TSTF-477, Add Action Statement for Two Inoperable Control Room Air Conditioning Subsystems, Using the Consolidated Line Item Improvement Process (CLIIP)2007-07-25025 July 2007 Application for Technical Specifications Change Based on TSTF-477, Add Action Statement for Two Inoperable Control Room Air Conditioning Subsystems, Using the Consolidated Line Item Improvement Process (CLIIP) JAFP-05-0084, Application for Technical Specifications (TS) Change Regarding Mode Change Limitations (TSTF-359) Using the Consolidated Line Item Improvement Process (CLIIP)2005-05-31031 May 2005 Application for Technical Specifications (TS) Change Regarding Mode Change Limitations (TSTF-359) Using the Consolidated Line Item Improvement Process (CLIIP) JAFP-03-0108, Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension2003-07-28028 July 2003 Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension ML0235000102002-12-0606 December 2002 Proposed Change to TS Regarding Local Power Range Monitor Calibration Frequency (JPTS-02-002) ML0222606432002-08-13013 August 2002 Technical Specifications for Amendment No. 275, Changes to the Technical Specification Leakage Limit for the Main Steam Isolation Valves from an Individual to an Aggregate Limit (Tac No. MB3333) ML0219701382002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Table of Contents ML0219701022002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B 3.7 Plant Systems ML0219700952002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), 3.6 Containment Systems ML0219700842002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Table of Contents ML0219700712002-07-0303 July 2002 Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Attachment 4, Table L - Less Restrictive Changes Matrix ML0219700282002-07-0303 July 2002 Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), Attachment Table a - Administrative Changes Matrix ML0219701422002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B 3.3 Instrumentation ML0219700922002-07-0303 July 2002 Technical Specifications, Amendment Conversion to Improved Technical Specifications (TAC No. MA5049), B. 3.4 Reactor Coolant System (RCS) JPN-02-016, Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling Dose Consequences2002-06-0707 June 2002 Proposed License Amendment for a Limited Scope Application of the Alternate Source Term Guidelines in NUREG-1465 Related to the Re-Evaluation of the Fuel Handling Dose Consequences ML18031A0412000-04-14014 April 2000 Issuance of Amendment 261 Changes to the Technical Specifications Regarding the Allowed Containment Leakage Rate JAFP-01-0133, ITS: 3.3.4.1 Anticipated Transient Without Scram Recirulation Pump Trip (ATWS-RPT) Instrumentation1995-04-0707 April 1995 ITS: 3.3.4.1 Anticipated Transient Without Scram Recirulation Pump Trip (ATWS-RPT) Instrumentation ML17037C4561975-08-11011 August 1975 Letter Transmitting Application for Amendment to Technical Specifications to Provide Reporting Intervals and Organizational Information Consistent with Appendix a of License Nos. DPR-63 and DPR-59 2023-05-25
[Table view] |
Text
v-w10NrEnheW Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Pete Dietrich Site Vice President JAFP-09-0087 July 31, 2009 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests an amendment to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power Plant (JAF).
This license amendment submittal requests modifying the TS requirements for testing of the Safety/Relief Valves (SRVs) by replacing the current requirement to manually actuate each SRV during plant startup with a series of overlapping tests that demonstrate the required functions of successive valve stages. provides the Application for Amendment to Modify the Technical Specifications Requirements On Testing of Safety/Relief Valves. provides the proposed TS changes as marked up pages. provides the proposed TS changes in final typed format with change bars. provides the proposed TS Bases changes as marked up pages. provides a copy of JAF 4 th IST Interval Relief Request VRR-06 (Info Only).
The TS Bases changes are provided for NRC information only. The final TS Bases pages will be submitted with a future update in accordance with TS 5.5.11, "Technical Specifications (TS)
Bases Control Program."
Entergy requests NRC approval of the proposed TS amendment by July 31, 2010, with the amendment being implemented within 60 days from approval.
In accordance with 10 CFR 50.91, a copy of this application, with the associated attachments, is being provided to the designated New York State official.
There are no new commitments made in this letter.
,&04-7 KA "?
JAFP-09-0087 Page 2 of 3 Questions concerning this report may be addressed to Mr. Joseph Pechacek, Licensing Manager, at (315) 349-6766.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on the 31st dao*"uiPly 2009.
Pete 'Dietrich --.
Site Vice President PD/JP/ed Attachments: 1. Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves
- 2. Proposed TS changes, on current marked up pages
- 3. Proposed TS changes, on typed final format pages
- 4. Proposed TS Bases change, as marked up pages (Info Only)
- 5. Proposed James A. FitzPatrick Fourth IST Interval Relief Request VRR-06 (Info Only) cc: next page
JAFP-09-0087 Page 3 of 3 cc:
Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. Bhalchandra Vaidya, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2A Washington, DC 20555-0001 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 1 0 th Floor Albany, NY 12223 Mr. Francis J. Murray Jr., President NYSERDA 17 Columbia Circle Albany, NY 12203-6399
JAFP-09-0087 Attachment 1 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves
JAFP-09-0087 Attachment 1
1.0 DESCRIPTION
The proposed amendment would modify the Technical Specifications (TS) requirements for testing of the Safety/Relief Valves (SRVs) by replacing the current requirement to manually actuate each SRV during plant startup with a series of overlapping tests that demonstrate the required functions of successive valve stages. Elimination of the manual actuation requirement is desirable to decrease the potential for SRV leakage and spurious SRV openings.
2.0 PROPOSED CHANGE
S Current TS Surveillance Requirement (TSSR) 3.4.3.2 states, "Verify each required SRV opens when manually actuated." TSSR 3.5.1.13 likewise states, "Verify each required ADS valve opens when manually actuated." The proposed amendment would change both TSSRs to verify each required valve "is capable of being opened." The current Frequency for both TSSRs is "24 months on a STAGGERED TEST BASIS for each valve solenoid"; this would be changed to state, "In accordance with the Inservice Testing Program."
Both TSSRs are modified by a NOTE that states, "Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test." This allowance would no longer be needed, and thus, would be deleted.
TS Bases associated with these Surveillance Requirements will be revised to describe the new testing method as discussed below. Revised Bases pages are attached for information, but do not require NRC approval.
3.0 BACKGROUND
SRVs installed at FitzPatrick are Target Rock model 7567F two-stage safety/relief valves. Eleven SRVs are installed on the main steam lines between the reactor vessel and the inboard main steam isolation 'Valves. Each SRV discharges via a separate tailpipe to a point below the water level in the suppression pool. SRVs open:
- In the safety mode on high reactor pressure, to provide primary overpressure protection to the reactor coolant pressure boundary.
- In the relief mode when actuated by the SRV Electric Lift logic on high reactor pressure, as a backup to the safety mode actuation.
In the relief mode when manually actuated by individual control switches in the Control Room, or by individual control switches in the Remote Shutdown system.
For seven of the eleven SRVs, in the relief mode when actuated by the Automatic Depressurization System (ADS) logic of the Emergency Core Cooling Systems (ECCS). The ADS function is to rapidly reduce reactor pressure to within the capacity of low pressure ECCS pumps in the event of a small or intermediate break Loss of Coolant Accident with the High Pressure Coolant Injection System (HPCI) unable to maintain level due to equipment failure or break size.
Experience in the industry and at FitzPatrick has shown that manual actuation of SRVs Page 1 of 6
JAFP-09-0087 Attachment 1 during plant operation leads to valve seat leakage. In particular, manual actuation testing has been the principle cause of main stage seat leakage at FitzPatrick. SRV leakage is routed to the suppression pool; the increased heat and fluid additions to the suppression pool require more frequent suppression pool cooling and pump-down operations. Main stage seat leakage also tends to mask the indications of pilot stage seat leakage; pilot stage leakage can cause maloperation of the SRV, including spurious actuation and/or failure to reclose after actuation. Excessive leakage of either stage requires plant shutdown to replace the leaking SRV.
The Boiling Water Reactor Owners' Group (BWROG) Evaluation of NUREG-0737, "Clarification of TMI Action Plan Requirements," Item I1.K.3.16, "Reduction of Challenges and Failures of Relief Valves," recommends that the number of SRV openings be reduced as much as possible and that unnecessary challenges should be avoided. NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants,"
NUREG-0123, "Standard Technical Specifications for General Electric Boiling Water Reactors," and NUREG-0626, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications" also recommend reducing the number of challenges to the SRVs.
4.0 TECHNICAL ANALYSIS
The manual actuation test currently prescribed in TSSRs 3.4.3.2 and 3.5.1.13 provides demonstration of the mechanical operation of the SRVs, and overlaps with other testing to demonstrate that the functions of the SRVs can be performed. The manual actuation test is performed once per operating cycle (two years), on a staggered test basis for the two SRV solenoids, so that each solenoid valve is tested every two cycles (four years).
The proposed testing uses a series of overlapping tests to demonstrate these functions.
Specifically:
- The' simulated automatic actuation test specified in TSSR 3.5.1.11, and additional surveillances associated with LCOs 3.3.5.1 and 3.3.3.2 and TRM 3.3.1, demonstrate the ability of various logics and controls to actuate the SRVs up to the point of energizing the solenoids. These tests are performed once per operating cycle (two years).
" A solenoid valve functional test will be performed in situ for each SRV solenoid valve once per operating cycle. This test demonstrates that when the solenoid is energized, it applies pneumatic pressure to the SRV actuator.
" An SRV actuator functional test will be performed at an offsite test facility as part of certification testing for each SRV pilot. Certification test intervals are determined in accordance with the Inservice Testing Program, which limits the maximum interval of service to six years under conditions described in Code Case OMN-17; typically, two-stage pilots are in service for a maximum of one operating cycle (two years). The actuator test demonstrates that the actuator moves the pilot stem when pneumatic ,pressure is applied.
- Setpoint testing is performed at the offsite test facility as part of certification testing for each SRV pilot, at intervals determined in accordance with the Inservice Testing Program. This test is the existing test required by Page 2 of 6
JAFP-09-0087 Attachment 1 TSSR 3.4.3.1. In addition to demonstrating that the SRV pilot stage will actuate on high steam pressure in the safety mode, this test overlaps with the actuator functional test to demonstrate that the pilot stage will actuate in the relief mode.
Main stage certification testing will be performed at the offsite test facility at intervals determined in accordance with the Inservice Testing Program, at least every six years per OMN-17. Main stage certification testing demonstrates that the main stage will open and port steam when actuated by the installed pilot stage.
TSSR 3.0.1 Bases states in part, "Surveillances may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified frequency." Whereas the above steps demonstrate the required safety functions, and whereas the testing frequency is once per operating cycle unless a longer frequency is specified by the Inservice Testing Program, the proposed testing satisfies this Bases statement.
In addition to the mechanical functional testing, the current manual actuation test demonstrates full steam flow capability of the SRV main stage and associated piping.
Following the initial demonstration during plant startup testing, limitations on steam flow would arise only through assembly errors or the introduction of foreign material into the piping system. Specific SRV maintenance procedures and plant Foreign Material Exclusion procedures and practices are sufficient to ensure full steam flow capability without periodic actuation testing. FitzPatrick has had no instance of test failure due to inadequate steam flow.
Failures of the manual actuation test have been uncommon. In April 2002, Hatch Unit 1 experienced an SRV failure to fully open and failure to reclose due to deterioration of the main stage piston-to-stem joint (Information Notice 03-001 and General Electric Service Information Letter 646). The deterioration involved a loss of joint preload followed by vibration induced wear that created a groove in the piston guide in which the piston hung up. This wear had occurred over a period of several cycles. FitzPatrick maintenance practice is to disassemble, refurbish, and retest each main stage following any removal from service; this practice would detect time-based degradation such as that involved in the Hatch event prior to main stage failure. Use of Code Case OMN-17 would require that each SRV be disassembled and refurbished every six years.
As discussed in the attached Relief Request, ASME OM Code requirements for testing of main steam pressure relief valves are satisfied by the above testing. The requirement of Section 1-3410(d) for manual actuation testing following reinstallation is to be exempted, and relief has been requested from that requirement based on the overlapping tests described above and the maintenance controls involved in reinstallation. The Relief Request provides additional discussion of Code Case OMN-17.
Page 3 of 6
JAFP-09-0087 Attachment 1 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration
- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the method of demonstrating the operability of the Safety/Relief Valves (SRVs) in both the safety and relief modes of operation. the SRVs are required to function in the safety mode to prevent overpressurization of the reactor vessel and reactor coolant system pressure boundary during various analyzed transients, including Main Steam Isolation Valve closure. SRVs associated with the Automatic Depressurization System are also required to function in the relief mode to reduce reactor pressure to permit injection by low pressure Emergency Core Cooling System (ECCS) pumps during certain reactor coolant pipe break accidents. The current testing method demonstrates the operability of the SRVs in both modes through manual actuation of the SRVs. The proposed new testing method demonstrates operability using a series of overlapping tests that show proper functioning of the SRV stages and supporting control components. This proposed testing method results in acceptable demonstration of the SRV functions in both the safety and relief modes, and therefore provides assurance that the probability of SRV failure will not increase. None of the accident safety analyses is affected by the requested Technical Specifications (TS) changes. Therefore, the consequences of accidents mitigated by the SRVs will not increase:
Certain SRV malfunctions are included in the FSAR safety analyses. Specifically, the plant safety analyses include the inadvertent opening of an SRV and a stuck open SRV.
By not actuating the SRVs during plant operation for testing and by reducing the incidence of pilot stage leakage of the SRVs, the proposed testing method will reduce the probability of these events.
Based on these considerations, the proposed test method does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change modifies the method of testing of the SRVs, but does not alter the functions or functional capabilities of the SRVs. Testing under the proposed method is performed in offsite test facilities or in the plant during outage periods when the SRV functions are not required. Existing analyses address events involving an SRV inadvertently opening or failing to reclose. Analyses also address the likelihood and consequences of failure of one or more SRVs to open. The proposed change does not Page 4 of 6
JAFP-09-0087 Attachment 1 introduce any new failure mode, and therefore, does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No.
Overpressure protection of the reactor coolant pressure boundary is based on the Safety/Relief Valves (SRVs) setpoint and total relief capacity. Setpoint is verified at an offsite testing facility; this requirement is not altered by the proposed change. Relief capacity of each SRV is determined by valve geometry, which is also not altered by the test methods. The margin of safety in the Loss of Coolant Accident analysis due to functioning of the Automatic Depressurization System is also based on total relief capacity of the associated SRVs. The proposed change in surveillance test methods demonstrates the operability of the SRVs, but does not alter the critical parameters that affect the margin of safety in analyses involving the SRV functions. Therefore, the proposed change does not involve a significant reduction in any margin of safety.
5.2 Applicable Regulatory Requirements / Criteria 10 CFR 50.36 requires in part that the operating license of a nuclear production facility include technical specifications. Paragraph (c)(2)(ii) of that part requires that a limiting condition for operation (LCO) of a nuclear reactor must be established for each item meeting one or more. of four criteria. The SRV functions identified in LCOs 3.4.3 and 3.5.1 both meet Criterion 3, "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." Paragraph (c)(3) further requires the establishment of surveillance requirements, "relating to test, calibration, or inspection to assure.. .that the limiting conditions for operation will be met." As discussed above, the proposed changes in the surveillance requirements for the SRVs are sufficient to demonstrate the safety and relief modes operation of the SRVs, and therefore, are sufficient to ensure that the limiting conditions for operation are met.
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.
6.0 ENVIRONMENTAL ASSESSMENT A review has determined that the proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed changes do not involve: (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet Page 5 of 6
JAFP-09-0087 Attachment 1 the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes.
7.0 PRECEDENT NUREG-1482 Paragraph 4.3.2.1 states, "In recent years, the NRC staff has received numerous requests for relief and/or TS changes related to the stroke testing requirements for BWR dual-function main steam safety/relief valves (SRVs). Both Appendix I to the ASME OM Code and the plant-specific TS require stroke testing of SRVs after they are reinstalled following maintenance activities. Several licensees have determined that in situ testing of the SRVs can contribute to undesirable seat leakage of the valves during subsequent plant operation and have received approval to perform testing at a laboratory facility coupled with in situ tests and other verifications of actuation systems as an alternative to the testing required by the ASME OM Code and TS."
The NRC has approved similar testing methods for Hatch, Hope Creek, and Limerick (prior to Limerick's conversion to three-stage SRVs).
Similar testing has also been approved for Dresden, Quad Cities, and Peach Bottom, which use three-stage Target Rock SRVs rather than two-stage SRVs. Testing approved for these plants included an in situ actuator test without steam (dry lift test).
The dry lift test is not suitable for two-stage SRVs because it has a high probability of causing unseating or leakage of the pilot stage, which can lead to spurious actuation or failure to reclose of the SRV.
Page 6 of 6
JAFP-09-0087 Attachment 2 Proposed Technical Specification Changes (Marked up)
Pages 3.4.3-2 3.5.1-7
S/RVs
.3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 +/- 34.3 psig. Following with the Inservice testing, lift settings shall be within +/- 1%. Testing Program SR 3.4.3.2 NOTE Not required to be pe .OF.mcd until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and floW arc adcquatct pefoFRm the test.
Verify each required S/RV opens when manua.l. 24 months on a actuated is capable of being opened. STAGGERED TEST BASlS for each valve solenoid In accordance with the Inservice Testing Program.
JAFNPP 3.4.3-2 Amendment 2-74
ECCS- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.13 NOTE Not reqUired to be pFr,-med until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> afte reactor steam pressure and flow are adequate-to peF orm the test.
24 moenths on a Verify each required ADS valve opens when STAGGERED manually actuated is capable of being opened. TEST BASIS for eaeh vave solenoid In accordance with the Inservice Testing Program JAFNPP 3.5.1-7 Amendment 2-7-4
JAFP-09-0087 Attachment 3 Proposed Technical Specification Changes (Final Typed)
Pages 3.4.3-2 3.5.1-7
S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 +/- 34.3 psig. Following with the Inservice testing, lift settings shall be within +/- 1%. Testing Program SR 3.4.3.2 Verify each required S/RV is capable of being In accordance opened. with the Inservice Testing Program.
JAFNPP 3.4.3-2 Amendment
ECCS- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.13 Verify each required ADS valve is capable of being In accordance opened. with the Inservice Testing Program JAFNPP 3.5.1-7 Amendment
JAFP-09-0087 Attachment 4 Proposed Technical Specification Bases Changes (Marked up)
(Information Only)
Pages B 3.4.3-4 B 3.5.1-16
S/RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 REQUIREMENTS (continued) Actuation of each required S/RV is'performed to verify that mechanically the valve is functioning properly. For the two-stage S/RV, this requires that the pilot stage be tested to show that it actuates when required and opens the associated main stage.
Likewise, the main stage must be tested to show that it opens and passes steam when the associated pilot stage actuates. The two stages are bench tested, together or separately, as part of the certification process, at intervals determined in accordance with the Inservice Testing Program. Maintenance procedures ensure that the S/RV stages are correctly installed in the plant, and that the S/RV and associated piping remain clear of foreign material that might obstruct valve operation or full steam flow. This approach provides adequate assurance that the required S/RVs will operate as required, while minimizing the challenges to the S/RVs and the likelihood of leakageor spurious operation.
For the purpose of this test, pilot actuation in the safety mode or relief mode is acceptable to satisfy the test requirements. Testing of the related solenoid valves is not required because they do not affect the safety mode operation of the S/RV. However, the solenoid valves are also tested in the IST program to support relief mode operation of the S/RVs for other functions.A manual actuation of each required
./RV is performed while bypassing main steam flow to the condense.
and observing Ž 10% closure of the tr. bine bypass valve- to vcri-,
that, mechanically, the valve is fundtioning properly andnobcag exists in the valve discharge line. This can also be demontatdb the response of the trbine control valves, by a change in the measured steam flow, Or by any other mnethod suitable to verif steam flow. Adequate reactor steamn dome pressure mustb t prfom tistest to avoid damaging the valve. Also, avalabe turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time i therefore allowed after the required pressure and fiow are achiee adequate...st.......owreust.
to performn this test. Adequate thepasi-,-g pressure+^~g at whichhemi this testu~eo is to be recommenations). Adequate steam flow is represented by two or more turbine bypass valves open, or total steam flow Ž!1-06 4b-Ibfh These conditions Will req-uire the plant to be in MODE 1, which has been shown to be an acceptable condition to perform this test. This test causes a small neutron flux transient which may cause a scram JAFNPP B 3.4.3-4 Revision 0-
S/RVs B 3.4.3 BASES in MOE 2 hile operating Giese to the Average Power Range Monitors Neut*on Flux- High (Sta-up) Allowable Value. Plant stahup is allowed prior to perform~ing this test because valve OPERABILI~
and the setpoints for overprcssurce protection are verified, per7 ASME-Code requirements, prior to valve instaIlation. Therefore, this SR is modified by a Note that states the Su~veillance is not required to be perform~ed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam prcssurc and flow are adequate to performA t he test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the Fequifed steam pressure and flow are reached is s ufficient to acoh ieve st-a bIem Oo-ndaliti ons for testi ng and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening-on overpressure, the safety function of the S/RV is considere OPERABLE; The 24 month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 24 month Frequncyvas developed based on the S/RV tests required by the ASMEh.- Boni~lper and Pressure Vessel Code,Section XI (Ref. 7). Operating expriecehas s hown that these components usually pass the S- meillanpce when performned at the 24 month Frequency. Therefore, the Frequenc isacpable from a Feliabilit' standpoint.
(continued)
JAFNPP B 3.4.3-4 Revision O_
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 (continued)
REQUIREMENTS The Frequency of 24 months is acceptable, given plant conditions required to perform the test and the other requirements existing to ensure adequate LPCI inverter performance during the 24 month interval. In addition, the Frequency is intended to be consistent with expected fuel cycle lengths.
SR 3.5.1.13 Actuation of each required ADS valve is performed to verify that mechanically the valve is functioning properly. For the two-stage S/RV, tests are required to demonstrate:
" That each ADS S/RV solenoid valve ports pneumatic pressure to the associated S/RV actuator when energized:
" That each ADS S/RV pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized: and
- That each ADS S/RV main stage opens and passes steam when the associated pilot stage actuates.
The solenoid valves are functionally tested once per cycle as part of the Inservice Testing Program. The pilot and main stages are bench tested, together or separately, as part of the certification process, at intervals determined in accordance with the Inservice Testing Program. Maintenance procedures ensure that the S/RV stages are correctly installed in the plant, and that the S/RV and associated piping remain clear of foreign material that might obstruct valve operation or full steam flow. This approach provides adequate assurance that the required ADS valves will operate when actuated, while minimizing the challenges to the valves and the likelihood of leakage or spurious operation. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
A manual actuation of each required ADS valve is performed while bypassing main steam flow to the condenser and obse~ving Žt 10% closure of the turbine bypass valves to Yerif' that the valve and solenoid are functioning properly and that no blockage exists in the S/RV discharge lines.
This can also be demonstrated by the response of the turbine control JAFNPP B 3.5.1-16 Revision 0
ECCS - Operating B 3.5.1 BASES or bypass valve Or by a cahange in the measured floW Or by any othcr
.ethodsuitab, e to ve.i. steamiflow. Adcquate reactor steam. dome pressure m.ust be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbinc Or turbinc bypass valves to continue to control reactor pressure when the ADS valves dive~t steam flo upn pning.-
Sufficient time is thercforc allowed after the reuie and Wress flow arc a.hic.ved to peForm this SR....qua...preR suea which this SR is to be performned isŽ!9700 psig (the prcssure consistcnt with vendor recommendations). Adequate steam flow is represented by at least two Or more turbine bypass valves opcn Or total steam flow
-- l'hr. These conditions Will require the plant to be in MODE 1, which has been shown to be an acceptable condition to perform this test. This test causes a small neutron flux transient whichmy as Range Monitors Neutron Flu - High (Startup) Allowable Value.
Reactor startup is allowed prior to performning this SR- because valve OPERABILITY and the setpoints for overpressure protec-tion are verified, per ASME reqUirements, prior to valve installation.
Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed forF manual actuation after the required pressure and flow are reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTiONAL TEST performed in LCO 3.3.5.1 oerldap this Surveillance to provide complete testing of the assumed safe~
funetien.
(continued)
JAFNPP B 3.5.1-16 Revision 0
JAFP-09-0087 Attachment 5 James A. FitzPatrick Nuclear Power Plant Fourth Inservice Test Interval Relief Request VRR-06 (Information Only)
I
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
System:
NUCLEAR BOILER AUTOMATIC DEPRESSURIZATION SYSTEM ASME Code Components Affected:
02RV-71A, B, C, D, E, F, G, H, J, K, L Component /System Function:
Eleven Safety/Relief Valves (SRVs) are installed on the main steam lines between the reactor vessel and the inboard main steam isolation valves. Each SRV discharges via a separate tailpipe to a point below the water level in the suppression pool. SRVs open:
- In the safety mode on high reactor pressure, to provide primary overpressure protection to the reactor coolant pressure boundary.
" In the relief mode when actuated by the SRV Electric Lift logic on high reactor pressure, as a backup to the safety mode actuation.
- In the relief mode when manually actuated by individual control switches in the Control Room, or by individual control switches in the Remote Shutdown system.
" For seven of the eleven SRVs, in the relief mode when actuated by the Automatic Depressurization System (ADS) logic of the Emergency Core Cooling Systems (ECCS).
The ADS function is to rapidly reduce reactor pressure to within the capacity of low pressure ECCS pumps in the event of a small or intermediate break Loss of Coolant Accident with the High Pressure Coolant Injection System (HPCI) unable to maintain level due to equipment failure or break size.
Applicable Code Edition and Addenda
ASME OM CODE 2001 Edition to 2003 Addenda Applicable Code Requirements:
Appendix I, paragraph 1-1320(a), "Test Frequencies, Class 1 Pressure Relief Valves" requires that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.
Appendix I, paragraph 1-3410(d),of the OM Code requires that valves that have been maintained or refurbished in place, removed for maintenance and testing, or both, and reinstalled shall be remotely actuated at reduced or normal system pressure to verify open and close capability of the valve before resumption of electric power generation.
Page 1 of 6
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Reason For Request:
This 4th Interval request for relief is based on Appendix I of the ASME OM Code-2001 Edition to 2003 Addenda. Exercising of the SRV after reinstallation can only be performed during reactor startup when there is sufficient steam pressure to actuate the main disk. Past history indicates that the main disks may not re-seat properly after being exercised during reactor startup resulting in steam leakage into the suppression pool. This leakage results in a decrease in plant performance and the potential for increased suppression pool temperatures which could force a plant shutdown to repair a leaking SRV. Past operating history indicates that the exercising performed during reactor startup is of no significant benefit in ensuring the proper operation of the individual SRV subassemblies.
This relief request also proposes to implement Code Case OMN-17 "Alternate Rules for Testing ASME Class 1 Pressure Relief/Safety Valves." OMN-17 states in Section (a) that safety valves shall be tested at least once every 72 months (6 years) with a minimum of 20% of the SRV group being tested within any 24-month interval. This 20% shall consist of valves thathave not been tested during the current 72-month interval, if they exist. The test interval for any individual valve that is in service shall not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods.
System
Description:
The SRVs are Target Rock Two-Stage, Model 7567F design. The SRVs are dual-function valves capable of being independently opened in either the safety or relief mode of operation.
A total of 11 SRVs are installed at the James A. FitzPatrick Nuclear Power Plant (JAF). In the safety mode (or spring mode of operation) the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. Each of the 11 SRVs can be opened manually in the relief mode from the control room by its associated two position switch. If one of these switches is placed in the open position the logic output will energize the associated SRV solenoid control valve directing the pneumatic supply to open the valve. Seven of these SRVs solenoid control valves can also be energized by the relay logic associated with the automatic depressurization system (ADS).
In addition each SRV can be manually operated from another control switch located at the ADS auxiliary panel located outside the control room.
Current Testing at JAF:
Testing of JAF SRVs is performed to satisfy Technical Specifications Surveillance Requirements (TSSRs) and the ASME OM Code-2001, "Code for Operation and Maintenance of Nuclear Power Plants with addenda." Certain tests are performed with the SRVs installed (in situ), while others are performed as "bench tests" after the valve is removed and transported to a maintenance and testing facility. Current requirements are as follows:
Page 2 of 6
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Current Testing at JAF (continued):
- 1. TS B 3.4.3-2 the safety function of the SRVs are required to be operable to satisfy the assumptions of the safety analysis. The requirements of this LCO are applicable only to the capability of the SRVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (Safety Function). This test is performed during the bench test at the vendor's facility.
- 2. TSSRs 3.4.3.2 and 3.5.1.13 verify each SRV opens when manually actuated.
- 3. Remote manual actuation is also required by ASME OM Code, Appendix I, paragraph I-3410(d), to verify open and close capability of the valve before resumption of electric power generation. This applies to valves that have been either maintained or refurbished in place, or removed for maintenance and testing and reinstalled.
Current Testing at Outside Facilities:
During each refueling outage, thel 1 SRV pilot assemblies and at least one main stage are removed and shipped to an offsite vendor for "as-found" testing, which includes visual inspection, leakage testing, and as-found set pressure testing. The tests are performed on a valve prior to maintenance on the valve. The pre-test leakage is measured at 1050 psig meeting the requirements of ASME OM Code-2001, 1-3310 (a), (b), and (c). Following the "as-found" testing, the SRVs are given a dimensional inspection followed by refurbishment, if required. This work is performed by the valve supplier. Post maintenance testing includes initial valve leakage testing, safety mode valve actuation to satisfy requirements for set pressure, reseat pressure, main disc stroke time, and final leakage testing. Final seat leakage tests are performed at approximately 1070 psig. Upon successful test completion, each valve receives written certification from the vendor and is returned to JAF for reinstallation. To receive certification, the valve must have zero seat leakage and meet the acceptance criteria of
+1- 1% for set pressure. These tests meet the requirements of ASME OM Code-2001, 1-3310 and Technical Specifications SR 3.4.3.1.
General Change Justification:
Leaking SRVs create operational problems associated with the suppression pool. SRV leakage increases both pool temperature and level, requiring more frequent use of the suppression pool cooling mode of the Residual Heat Removal (RHR) system.
As described previously, each SRV pilot assembly and at least one main stage removed during the refuel outage is tested at an offsite facility. The as found testing is performed within 12 months of removal, meeting the OM code requirements. The valves are refurbished the following year just prior to the refuel outage as necessary to meet the acceptance criteria of zero leakage, and are certified in writing as being leak free. The valves are then reinstalled in the plant and proper pilot operation is confirmed through leak rate testing of the pilot air operators and associated accumulator piping followed by manual lift at reactor power.
Page 3 of 6
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
General Change Justification (continued):
Several aspects of SRV design and operation can contribute to valve leakage. As mentioned earlier, these include test pressure, pilot valve disc and rod configuration, and system and valve cleanliness. Actuation of the SRVs after laboratory testing by any means allows these contributors to impact the ability of the valve to re-close completely. JAF has made significant efforts to minimize the effects of these contributors.
JAF currently uses ASME OM Code 2001 section 1-1320 "Test Frequencies, Class I Pressure Relief Valves." This establishes the five year frequency for SRV testing. JAF proposes to use Code Case OMN-17 "Alternate Rules for Testing ASME Class 1 Pressure Relief/Safety Valves."
This Code Case changes the frequency to six years to coincide with the 24-month refueling cycle at JAF.
Additionally, reducing challenges to the SRVs is a recommendation of NUREG-0737; "TMI Action Plan Requirements" item I1.K.3.(16). This recommendation is based on a stuck open SRV being a possible cause of a Loss of Coolant Accident (LOCA). This relief request is consistent with that NRC recommendation.
Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(ii):
As an alternate to the testing required by ASME OM Code-2001, Appendix I, paragraph 1-3410(d), JAF proposes to actuate the SRVs in the relief mode at the test facility (i.e., Wyle Laboratory). A test solenoid valve will be energized, the actuator will stroke, and the pilot rod lift will be verified. This test will verify that, given a signal to energize the solenoid, the pilot disc rod will lift. The pilot function will be recorded in the test documentation package for future reference, as needed. Alternate testing is justified since the remaining segments of the SRV relief mode of operation are proven by other tests. The ability of the pilot disc to open is shown
'in the safety mode actuation bench test. The integrity of the pneumatic and solenoid system for the SRVs is verified by performance of post maintenance leak rate testing, continuity testing, and a functional test of the solenoids while detached from the SRV. The joint between the pilot and solenoid / manifold assembly are visually inspected under the maintenance procedure with a second verification to ensure that the "0"ring is in place during reassembly. Automatic valve actuation is proven operable by logic system functional tests which include verification that the solenoid is energized by the automatic signal. The actuator to main body joint is inspected during ISI VT-2 exam performed prior to startup. The above proposed surveillance and testing of the SRVs and associated components provides reasonable assurance of adequate valve operation and readiness.
JAF proposes to implement Code Case OMN-17 that requires in section (a) 72-month test interval for Class 1 pressure relief valves with a minimum of 20% of the SRV group being tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve that is in service shall not exceed 72 months except that a six month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods.
Page 4 of 6
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(ii) (continued:
ASME Code Case OMN-17 section (d) "maintenance", requires the owner to disassemble and inspect each valve after as found set pressure testing to verify that parts are free of defects resulting from time related degradation or service induced wear. Section (e) requires that each valve shall have been disassembled and inspected in accordance with section (d) prior to the start of the 72-month (6 year) test interval.
Each refueling outage, 100% of the pilot assemblies and approximately one third of the main disc assemblies (4, 4, and 3 main assemblies over a three cycle period) will be sent to an offsite vendor and tested with steam pressure. As a result, even though actual valve movement is not performed after the SRV is re-installed in the plant, all pilot assemblies are tested with steam pressure once per cycle and all the main discs are tested with steam pressure at least once every three cycles. This proposed testing meets the testing requirements of Code Case OMN-
- 17. On the basis that compliance with ASME OM Code 2001 Edition 2003 Addenda testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, this proposed alternative should be granted pursuant to 10 CFR 50.55a(a)(3)(ii).
Duration:
The proposed alternative identified in this 10 CFR 50.55a Request shall be utilized during the Fourth Ten year IST Interval that began on October 1, 2007.
Precedents:
NUREG-1482 Paragraph 4.3.2.1 states, "In recent years, the NRC staff has received numerous requests for relief and/or TS changes related to the stroke testing requirements for BWR dual-function main steam SRVs. Both Appendix I to the ASME OM Code and the plant-specific TS require stroke testing of SRVs after they are reinstalled following maintenance activities.
Several licensees have determined that in situ testing of the SRVs can contribute to undesirable seat leakage of the valves during subsequent plant operation and have received approval to perform testing at a laboratory facility coupled with in situ tests and other verifications of actuation systems as an alternative to the testing required by the ASME OM Code and TS."
The NRC has approved similar testing methods at Nine Mile Point Station: reference MSS-VR-01 (Unit 2).
Similar testing has also been approved for Dresden, Quad Cities, and Peach Bottom, which use three-stage Target Rock SRVs rather than two-stage SRVs. Testing approved for these plants included an in situ actuator test without steam (dry lift test). The dry lift test is not suitable for two-stage SRVs because it has a high probability of causing unseating or leakage of the pilot stage, which can lead to spurious actuation or failure to reclose of the SRV.
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT 10 CFR 50.55a Request VRR-06 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
References:
- 1. ASME OM Code, 2001 Edition through 2003 Addenda
- 2. Code Case OMN-17, "Alternative Rules for Testing ASME Class 1 Pressure Relief Safety Valves
- 3. NUREG-1482, "Guidelines for Inservice Testing at Nuclear power Plants", April 1995 Page 6 of 6