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Category:Letter
MONTHYEARML24022A1172024-01-23023 January 2024 Acceptance of Requested Licensing Action Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations ML24012A2422024-01-16016 January 2024 Acceptance of Requested Licensing Action License Transfer Request ML23236A0042023-12-27027 December 2023 Issuance of Amendment 274 Re Changes to Perm Defueled Emergency Plan and Perm Defueled Emergency Action Level Scheme ML23355A1242023-12-26026 December 2023 Withdrawal of an Amendment Request Re License Amendment Request to Revise License Condition to Eliminate Cyber Security Plan Requirements ML23192A0772023-12-26026 December 2023 Letter Exemption from the Requirements of 10 CFR 140.11(a)(4) Concerning Offsite Primary and Secondary Liability Insurance ML23191A5222023-12-22022 December 2023 Exemption Letter from the Requirements of 10 CFR 50.54(W)(1) Concerning Onsite Property Damage Insurance (EPID - L-2022-LLE-0032) ML23263A9772023-12-22022 December 2023 Exemption from Certain Emergency Planning Requirements and Related Safety Evaluation ML23354A2602023-12-21021 December 2023 Reference Simulator Inspection Request for Information L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 PNP 2023-030, License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations2023-12-14014 December 2023 License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations PNP 2023-035, Withdrawal of License Amendment Request - Revise License Condition to Eliminate Cyber Security Plan Requirements2023-12-12012 December 2023 Withdrawal of License Amendment Request - Revise License Condition to Eliminate Cyber Security Plan Requirements PNP 2023-028, Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments2023-12-0606 December 2023 Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments L-23-012, Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point2023-11-13013 November 2023 Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point ML23291A4402023-11-0303 November 2023 Acceptance of Requested Licensing Action Request for Exemption from 10 CFR 50.82(a)(2) to Support Reauthorization of Power Operations IR 05000255/20230032023-10-0404 October 2023 NRC Inspection Report No. 05000255/2023003(DRSS)-Holtec Decommissioning International, LLC, Palisades Nuclear Plant ML23275A0012023-10-0202 October 2023 Request for Withholding Information from Public Disclosure for Palisades Nuclear Plant PNP 2023-025, Request for Exemption from Certain Termination of License Requirements of 10 CFR 50.822023-09-28028 September 2023 Request for Exemption from Certain Termination of License Requirements of 10 CFR 50.82 PNP 2023-026, Pre-Submittal Meeting Presentation - Palisades Nuclear Plant License Transfer Application to Support Resumption of Power Operations2023-09-28028 September 2023 Pre-Submittal Meeting Presentation - Palisades Nuclear Plant License Transfer Application to Support Resumption of Power Operations PNP 2023-023, Special Report High Range Noble Gas Monitor Inoperable2023-08-0909 August 2023 Special Report High Range Noble Gas Monitor Inoperable ML23215A2302023-08-0303 August 2023 Notice of Organization Change - Chief Nuclear Officer IR 05000255/20230022023-07-19019 July 2023 NRC Inspection Report 05000255/2023002 DRSS-Holtec Decommissioning International, LLC, Palisades Nuclear Plant ML23087A0362023-05-0202 May 2023 PSDAR Review Letter ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment PNP 2023-018, 2022 Annual Non-Radiological Environmental Operating Report2023-04-25025 April 2023 2022 Annual Non-Radiological Environmental Operating Report L-23-004, HDI Annual Occupational Radiation Exposure Data Reports - 20222023-04-24024 April 2023 HDI Annual Occupational Radiation Exposure Data Reports - 2022 PNP 2023-007, and Big Rock Point, 2022 Annual Radioactive Effluent Release and Waste Disposal Reports2023-04-19019 April 2023 and Big Rock Point, 2022 Annual Radioactive Effluent Release and Waste Disposal Reports PNP 2023-008, 2022 Radiological Environmental Operating Report2023-04-18018 April 2023 2022 Radiological Environmental Operating Report L-23-003, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-31031 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations PNP 2023-002, 6 to Updated Final Safety Analysis Report2023-03-31031 March 2023 6 to Updated Final Safety Analysis Report ML23088A0382023-03-29029 March 2023 Stations 1, 2, & 3, Palisades Nuclear Plant, and Big Rock Point - Nuclear Onsite Property Damage Insurance PNP 2023-006, Report of Changes to Security Plan, Revision 202023-03-29029 March 2023 Report of Changes to Security Plan, Revision 20 PNP 2023-012, Presentation on Regulatory Path to Reauthorize Power Operations2023-03-16016 March 2023 Presentation on Regulatory Path to Reauthorize Power Operations ML23038A0982023-03-15015 March 2023 Request for Withholding Information from Public Disclosure ML23095A0642023-03-14014 March 2023 American Nuclear Insurers, Notice of Cancellation Rescinded PNP 2023-001, Regulatory Path to Reauthorize Power Operations2023-03-13013 March 2023 Regulatory Path to Reauthorize Power Operations PNP 2023-004, Report of Changes to Palisades Nuclear Plant Technical Specification Bases2023-03-0808 March 2023 Report of Changes to Palisades Nuclear Plant Technical Specification Bases PNP 2023-005, Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report2023-03-0101 March 2023 Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report ML22361A1022023-02-24024 February 2023 Reactor Decommissioning Branch Project Management Changes for Some Decommissioning Facilities and Establishment of Backup Project Manager for All Decommissioning Facilities ML23052A1092023-02-17017 February 2023 FEMA Letter to NRC, Proposed Commission Paper Language for Palisades Nuclear Plant Emergency Plan Decommissioning Exemption Request ML23032A3992023-02-0101 February 2023 Regulatory Path to Reauthorize Power Operations IR 05000255/20220032022-12-28028 December 2022 NRC Inspection Report No. 05000255/2022003(DRSS); 07200007/2022001 (Drss) Holtec Decommissioning International, LLC, Palisades Nuclear Plant L-22-042, Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.152022-12-14014 December 2022 Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.15 PNP 2022-037, Report of Changes to Security Plan, Revision 192022-12-14014 December 2022 Report of Changes to Security Plan, Revision 19 ML22321A2852022-11-17017 November 2022 LLC Master Decommissioning Trust Agreement for Palisades Nuclear Plant IR 05000255/20224012022-11-0909 November 2022 Decommissioning Palisades Nuclear Plant - Decommissioning Security Inspection Report 05000255/2022401 PNP 2022-036, Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme2022-11-0808 November 2022 Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme PNP 2022-035, International - Notification of Commitment Cancellations for Remaining Activities Related to Beyond-Design-Basis Seismic Hazard Reevaluations2022-11-0202 November 2022 International - Notification of Commitment Cancellations for Remaining Activities Related to Beyond-Design-Basis Seismic Hazard Reevaluations PNP 2022-024, Request for Exemption from 10 CFR 140.11(a)(4) Concerning Primary and Secondary Liability Insurance2022-10-26026 October 2022 Request for Exemption from 10 CFR 140.11(a)(4) Concerning Primary and Secondary Liability Insurance PNP 2022-026, Request for Exemption from 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance2022-10-26026 October 2022 Request for Exemption from 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance ML22292A2572022-10-25025 October 2022 Permanently Defueled Emergency Plan License Amendment RAI Letter 2024-01-23
[Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50
MONTHYEARPNP 2023-030, License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations2023-12-14014 December 2023 License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations PNP 2022-031, License Amendment Request, Revise License Condition to Eliminate Cyber Security Plan Requirements2022-09-14014 September 2022 License Amendment Request, Revise License Condition to Eliminate Cyber Security Plan Requirements PNP 2022-016, License Amendment Request: Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme2022-07-12012 July 2022 License Amendment Request: Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme ML21152A1112021-06-0101 June 2021 Proposed Changes (mark-up) to Palisades Plant - Appendix a Technical Specifications, and Appendix B Environmental Protection Plan Pages ML21152A1122021-06-0101 June 2021 Page Change Instructions and Retyped Pages for the Palisades Plant Appendix a Permanently Defueled Technical Specifications and Appendix B Environmental Protection Plan ML21152A1132021-06-0101 June 2021 Palisade Nuclear Plant - Proposed Technical Specifications Bases Changes ML21152A1102021-06-0101 June 2021 Evaluation of the Proposed Changes PNP 2020-042, and Big Rock Point Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments2020-12-23023 December 2020 and Big Rock Point Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments 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 ML19149A3032019-05-28028 May 2019 Enclosure Attachment 1 to Pnp 2019-028: Renewed Facility Operating License Page Markups ML19149A3042019-05-28028 May 2019 Enclosure Attachment 1 (Continued) to Pnp 2019-028: Operating License Page Change Instructions and Retyped Renewed Facility Operating License Pages ML19149A3022019-05-28028 May 2019 Enclosure to Pnp 2019-028: Response to Request for Additional Information - License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications PNP 2019-007, License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications2019-03-0808 March 2019 License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications CNRO-2019-00003, Application to Revise Technical Specifications to Adopt TSTF-529 11 Clarify Use and Application Rules, 11 Revision 42019-01-31031 January 2019 Application to Revise Technical Specifications to Adopt TSTF-529 11 Clarify Use and Application Rules, 11 Revision 4 ML18305B3212018-11-0101 November 2018 License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications ML18305B3222018-11-0101 November 2018 License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications - Proposed Operating License Change PNP 2018-040, License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications - Proposed Renewed Facility Operating License Page Changes (Markup)2018-11-0101 November 2018 License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications - Proposed Renewed Facility Operating License Page Changes (Markup) ML18270A3252018-09-27027 September 2018 Attachment 4: Page Change Instructions and Revised Pages, and Appendix a Technical Specifications ML18270A3232018-09-27027 September 2018 Attachment 2: Probabilistic Risk Assessment Technical Adequacy ML18270A3242018-09-27027 September 2018 Attachment 3: Proposed Changes and Appendix a Technical Specifications Pages ML18270A3262018-09-27027 September 2018 Attachment 5: Proposed Technical Specification Bases Changes ML18270A3222018-09-27027 September 2018 Attachment 1: Description and Evaluation of Technical Specifications Changes ML18270A3282018-09-27027 September 2018 Attachment 7: Traveler TSTF-425 Versus Pnp RFOL SR Applicability Cross Reference Table ML18270A3272018-09-27027 September 2018 Attachment 6: No Significant Hazards Consideration PNP 2018-025, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b.2018-09-27027 September 2018 License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b. ML18152A9292018-05-30030 May 2018 License Amendment Request to Revise Emergency Diesel Generator Degraded Voltage Surveillance Requirement PNP 2018-004, Supplement to License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications2018-01-24024 January 2018 Supplement to License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications PNP 2017-066, License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications2017-11-0101 November 2017 License Amendment Request to Change the Full Compliance Implementation Date for the Fire Protection Program Transition License Condition for Required Modifications PNP 2017-059, License Amendment Request Supplement - Cyber Security Plan Implementation Schedule2017-10-17017 October 2017 License Amendment Request Supplement - Cyber Security Plan Implementation Schedule PNP 2017-035, License Amendment Request - Administrative Controls for a Permanently Defueled Condition2017-07-27027 July 2017 License Amendment Request - Administrative Controls for a Permanently Defueled Condition PNP 2017-007, License Amendment Request, Cyber Security Plan Implementation Schedule2017-03-30030 March 2017 License Amendment Request, Cyber Security Plan Implementation Schedule PNP 2016-046, License Amendment Request for Editorial and Administrative Changes to Technical Specifications2016-11-0909 November 2016 License Amendment Request for Editorial and Administrative Changes to Technical Specifications PNP 2016-049, License Amendment Request: Control Rod Drive Exercise Surveillance2016-08-22022 August 2016 License Amendment Request: Control Rod Drive Exercise Surveillance PNP 2016-005, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify SR Usage Rule Application to Section 5.5 Testing.2016-07-11011 July 2016 License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify SR Usage Rule Application to Section 5.5 Testing. PNP 2016-001, License Amendment Request - Revision to the Requirements for Steam Generator Tube Inspections and Repair Criteria in the Cold Leg Tube Sheet Region2016-03-0303 March 2016 License Amendment Request - Revision to the Requirements for Steam Generator Tube Inspections and Repair Criteria in the Cold Leg Tube Sheet Region PNP 2015-035, License Amendment Request - Cyber Security Plan Implementation Schedule2015-06-11011 June 2015 License Amendment Request - Cyber Security Plan Implementation Schedule PNP 2014-111, Revised License Condition - License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors2014-12-18018 December 2014 Revised License Condition - License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors PNP 2014-091, Supplement to License Amendment Request to Revise Emergency Response Organization Staff Augmentation Response Times2014-12-0909 December 2014 Supplement to License Amendment Request to Revise Emergency Response Organization Staff Augmentation Response Times ML14237A1442014-12-0808 December 2014 Issuance of Amendment Cyber Security Plan Implementation Schedule ML14211A5252014-07-29029 July 2014 WCAP-17628-NP, Revision 1, Palisades, Alternate Pressurized Thermal Shock (PTS) Rule Evaluation. PNP 2014-049, License Amendment Request to Implement 10 CFR 50.61 a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.2014-07-29029 July 2014 License Amendment Request to Implement 10 CFR 50.61 a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. PNP 2014-042, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-426, Revise or Add Actions to Preclude Into LCO 3.0.3 - RITSTF Initiatives 6b & 6c, Using the Consolidated Line Item Improvement Process2014-06-11011 June 2014 License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-426, Revise or Add Actions to Preclude Into LCO 3.0.3 - RITSTF Initiatives 6b & 6c, Using the Consolidated Line Item Improvement Process PNP 2013-086, License Amendment Request - Cyber Security Plan Implementation Schedule2013-12-30030 December 2013 License Amendment Request - Cyber Security Plan Implementation Schedule PNP 2013-044, License Amendment Request to Revise Emergency Response Organization Staff Augmentation Response Times2013-06-25025 June 2013 License Amendment Request to Revise Emergency Response Organization Staff Augmentation Response Times PNP 2012-074, License Amendment Request - Technical Specifications Change to Add LCO 3.0.8 on the Lnoperability of Snubbers2012-09-0606 September 2012 License Amendment Request - Technical Specifications Change to Add LCO 3.0.8 on the Lnoperability of Snubbers PNP 2012-046, License Amendment Request - Cyber Security Plan Implementation Schedule Milestones2012-06-20020 June 2012 License Amendment Request - Cyber Security Plan Implementation Schedule Milestones PNP 2012-005, License Amendment Request - Replacement of Spent Fuel Pool Region 1 Storage Racks2012-02-28028 February 2012 License Amendment Request - Replacement of Spent Fuel Pool Region 1 Storage Racks PNP 2011-058, License Amendment Request to Revise Calculated Peak Containment Internal Pressure2011-08-16016 August 2011 License Amendment Request to Revise Calculated Peak Containment Internal Pressure PNP 2011-018, Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years2011-04-0606 April 2011 Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years PNP 2011-002, License Amendment Request for Spent Fuel Pool Region I Criticality2011-01-31031 January 2011 License Amendment Request for Spent Fuel Pool Region I Criticality 2023-12-14
[Table view] Category:Technical Specification
MONTHYEARPNP 2023-030, License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations2023-12-14014 December 2023 License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations ML22102A2482022-04-14014 April 2022 Request for Additional Information Regarding License Amendment Request for Permanently Defueled Technical Specifications ML21152A1112021-06-0101 June 2021 Proposed Changes (mark-up) to Palisades Plant - Appendix a Technical Specifications, and Appendix B Environmental Protection Plan Pages ML21152A1132021-06-0101 June 2021 Palisade Nuclear Plant - Proposed Technical Specifications Bases Changes 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 PNP 2019-017, Submittal of Changes to Technical Specifications Bases2019-03-21021 March 2019 Submittal of Changes to Technical Specifications Bases PNP 2018-052, Supplement to License Amendment Request -Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force.2018-11-16016 November 2018 Supplement to License Amendment Request -Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force. ML18270A3222018-09-27027 September 2018 Attachment 1: Description and Evaluation of Technical Specifications Changes ML18270A3242018-09-27027 September 2018 Attachment 3: Proposed Changes and Appendix a Technical Specifications Pages ML18270A3252018-09-27027 September 2018 Attachment 4: Page Change Instructions and Revised Pages, and Appendix a Technical Specifications ML18270A3262018-09-27027 September 2018 Attachment 5: Proposed Technical Specification Bases Changes PNP 2017-067, Submittal of Changes to Palisades Nuclear Plant Technical Specifications Bases2017-11-0808 November 2017 Submittal of Changes to Palisades Nuclear Plant Technical Specifications Bases PNP 2016-046, License Amendment Request for Editorial and Administrative Changes to Technical Specifications2016-11-0909 November 2016 License Amendment Request for Editorial and Administrative Changes to Technical Specifications PNP 2014-088, Report to NRC of Changes to Technical Specifications Bases2014-09-15015 September 2014 Report to NRC of Changes to Technical Specifications Bases PNP 2014-042, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-426, Revise or Add Actions to Preclude Into LCO 3.0.3 - RITSTF Initiatives 6b & 6c, Using the Consolidated Line Item Improvement Process2014-06-11011 June 2014 License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-426, Revise or Add Actions to Preclude Into LCO 3.0.3 - RITSTF Initiatives 6b & 6c, Using the Consolidated Line Item Improvement Process PNP 2012-097, Report to NRC of Changes to Technical Specifications Bases2012-11-0909 November 2012 Report to NRC of Changes to Technical Specifications Bases PNP 2011-032, Report to NRC of Changes of Technical Specifications Bases2011-04-25025 April 2011 Report to NRC of Changes of Technical Specifications Bases PNP 2011-018, Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years2011-04-0606 April 2011 Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years ML1020105792010-07-20020 July 2010 License Amendment Request for Emergency Diesel Generator Fuel Oil and Lube Oil Systems ML0930604142009-11-0202 November 2009 Report of Changes to Technical Specifications Bases ML0913501812009-05-14014 May 2009 Supplement to License Amendment Request: Changes to Technical Specification Administrative Controls Section 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 ML0901602382009-02-0606 February 2009 Issuance of Amendment Spent Fuel Pool Region I Storage Requirements ML0827404452008-11-20020 November 2008 Issuance of Amendment Eliminate License Condition 2.F ML0822703122008-08-14014 August 2008 Report of Changes to Technical Specifications Bases ML0723203952007-08-16016 August 2007 Report of Changes to Technical Specifications Bases ML0722201032007-07-25025 July 2007 Submittal of Technical Specifications Bases Changes ML0718002162007-06-28028 June 2007 License Amendment Request: Replacement of Containment Sump Buffer ML0714501052007-05-31031 May 2007 Technical Specifications Pages Re Programs and Manuals for Steam Generator ML0631803592006-11-0606 November 2006 License Amendment Request: Realistic Large Break Loss-of-Coolant Accident ML0628303852006-09-25025 September 2006 Alternative Source Term- Proposed License Amendment ML0623000332006-08-17017 August 2006 Report of Changes to Technical Specifications Bases ML0524102132005-08-22022 August 2005 Report of Changes to Technical Specifications Bases ML0425904452004-09-0202 September 2004 Report of Changes to Technical Specifications Bases ML0421604082004-08-0202 August 2004 Tech Spec Pages for Amendment No. 217 Regarding the Relocating of the Primary Coolant System Pressure, Cold-Leg Temperature and Flow Departure from Nucleate Boiling Limits to the Core Operating Limits Report ML0404204242004-01-30030 January 2004 Application for Technical Specification Improvement to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors Using the Consolidated Line Item Improvement Process ML0330001112003-10-17017 October 2003 Report of Changes to Technical Specifications Bases ML0328900502003-10-0606 October 2003 License Amendment Request: Increase Rated Thermal Power, Response to Request for Additional Information ML0316110532003-06-0303 June 2003 License Amendment Request: Increase Rated Thermal Power ML0303805792003-01-22022 January 2003 Technical Specifications and Bases ML0231701012002-10-31031 October 2002 Report of Changes to Technical Specifications Bases ML0207201012002-03-0101 March 2002 License Amendment Request, Surveillance Requirement Frequency for Containment Spray Nozzle Inspections ML0205901512002-02-26026 February 2002 Amendment No. 207 Retechnical Specification on Spent Fuel Pool Boron Concentration ML18344A3791978-06-15015 June 1978 Proposed Technical Specification Changes on Power Distribution Limits ML18353B1461978-03-0909 March 1978 Iodine Removal System - Proposed Technical Specification Changes ML18353A8671978-02-22022 February 1978 Letter Proposed Technical Specifications Change on Heatup and Cooldown Curves for Cycle 3 Operation ML18353A8741978-02-21021 February 1978 Letter Proposed Technical Specifications Change on Steam Generator Operating Allowance ML18347A1881978-01-0303 January 1978 Proposed Technical Specifications Changes Relating to Overpressure Protection System ML18353B1451977-11-11011 November 1977 Transmittal of Technical Specification - Deletion of Annual Operating Report ML18353B1471977-11-10010 November 1977 Proposed Technical Specification Changes for Steam Generator Inservice Inspection 2023-12-14
[Table view] Category:Amendment
MONTHYEARPNP 2023-030, License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations2023-12-14014 December 2023 License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations ML22102A2482022-04-14014 April 2022 Request for Additional Information Regarding License Amendment Request for Permanently Defueled Technical Specifications ML21152A1112021-06-0101 June 2021 Proposed Changes (mark-up) to Palisades Plant - Appendix a Technical Specifications, and Appendix B Environmental Protection Plan Pages 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 PNP 2018-052, Supplement to License Amendment Request -Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force.2018-11-16016 November 2018 Supplement to License Amendment Request -Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force. PNP 2011-018, Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years2011-04-0606 April 2011 Regarding License Amendment Request to Extend the Containment Type a Leak Rate Test Frequency to 15 Years 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 ML0901602382009-02-0606 February 2009 Issuance of Amendment Spent Fuel Pool Region I Storage Requirements ML0827404452008-11-20020 November 2008 Issuance of Amendment Eliminate License Condition 2.F ML0718002162007-06-28028 June 2007 License Amendment Request: Replacement of Containment Sump Buffer ML0714501052007-05-31031 May 2007 Technical Specifications Pages Re Programs and Manuals for Steam Generator ML0631803592006-11-0606 November 2006 License Amendment Request: Realistic Large Break Loss-of-Coolant Accident ML0628303852006-09-25025 September 2006 Alternative Source Term- Proposed License Amendment ML0421604082004-08-0202 August 2004 Tech Spec Pages for Amendment No. 217 Regarding the Relocating of the Primary Coolant System Pressure, Cold-Leg Temperature and Flow Departure from Nucleate Boiling Limits to the Core Operating Limits Report ML0404204242004-01-30030 January 2004 Application for Technical Specification Improvement to Eliminate Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors Using the Consolidated Line Item Improvement Process ML0328900502003-10-0606 October 2003 License Amendment Request: Increase Rated Thermal Power, Response to Request for Additional Information ML0316110532003-06-0303 June 2003 License Amendment Request: Increase Rated Thermal Power ML0205901512002-02-26026 February 2002 Amendment No. 207 Retechnical Specification on Spent Fuel Pool Boron Concentration ML18344A3791978-06-15015 June 1978 Proposed Technical Specification Changes on Power Distribution Limits ML18347A1881978-01-0303 January 1978 Proposed Technical Specifications Changes Relating to Overpressure Protection System ML18353B1471977-11-10010 November 1977 Proposed Technical Specification Changes for Steam Generator Inservice Inspection ML18347A1711977-09-26026 September 1977 Additional Information Relating to Power Increase Request ML18348A7651976-11-16016 November 1976 Proposed Change to Technical Specification to Modify Spent Fuel Storage Facility ML18347A5511976-10-0808 October 1976 Transmitting Proposed Change to Tech Spec Concerning Operability & Surveillance Requirements for Hydraulic Shock Suppressors ML18347A6601976-06-0909 June 1976 Re Cpc'S 5/25/1976 Event Report, Requesting Change to Tech Spec to Operation of Diesel Generator with One High-Pressure Safety Injection Pump Out of Service ML18348A7701976-03-12012 March 1976 Proposed Change to Technical Specification to Modify Requirements for Testing of Primary System Pump Combinations & for Load Testing of Diesel Generators ML18348A7751976-02-25025 February 1976 Proposed Change to Technical Specification to Delete Requirement for Operation of Fuel Pool Charcoal Adsorber After I-131 Level in Irradiated Fuel Has Decayed by Factor of 10 ML18348A7761976-01-30030 January 1976 Proposed Change to Technical Specification to Generalize Tech Spec to Allow for Reactor Refueling Without Requiring Additional Tech Spec Changes ML18348A7801976-01-30030 January 1976 Proposed Change to Technical Specification to Revise Steam Generator Tube Plugging Criteria ML18348A7861975-03-10010 March 1975 Proposed Change to Technical Specification Based on Eddy-Current Inspection of Steam Generator 2023-12-14
[Table view] |
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Director, Nuclear Reactor Regulation Att: Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Reguiatory Commission Washington, DC 20555 \
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.' '*' .'i i *'
TO POWER DISTRIBUTION LIMITS ii Transmitted. herewith are three {3) originai\and. thirty-seven (37) conformed copies of a proposed. change to the.,Technical_Specifications for the Palisades' Plant, Docket 50-255, License DPR-.20 *. ;,
The purpose of this change is to modify allowable power distribution limits as determined by Exxon Nuclear Company, *rnc and reported in XN-NF-78-16 entitled, "Analysis of Axial Power. ~istrib1it;ion Limits for the Palisades Nuclear Reactor at 2530 MWt" date~ June *1, }978. For"ty (40) copies of t:tiis report were transmitted directly from Exxon to. the NRC by a letter dated June 15, 1978. 'L
Consumers Power requests that ~he propose'd changes be reviewed. as rapid,ly es possible so that current operating. restrict;ions related to power -'distribution limits may be eliminated by September l; 19f.18. To. maintain the currerit **
Technical Specification power distribution iimits the plant must operate with group 4 rods partially inserted. at P?wer levels above 85% full power.
Pursuant* to 10 CFR 170.22 the requested c.hanges involve a single issue *
. As such, it is concluded that the request. should-be--classified as Class III.
A check for the apprOJ;>riate amount is at~acheQ..
~~Ju~
David P Hoffman
- ------_. ~, r 7817::::06::::1 I
Assistant Nuclear Licensing A~~:q,i_strator CC: JGKeppler, USNRC
CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications
.License DPR-20 For the reasons hereinafter set forth., it is requested that the Technical Specifications contained ;n Provisional Operating License-DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 for the Palisades Plant be changed as described in Section I below:
I. Change A. Add new definitions for Assembl~ Radial Peaking Factor -FrA and Total Radial Peaking Factor -Fr to Section 1.1 as indicated on attached Technical Specification page changes.
B. Delete references to Hot Channel Factors in Section 2.1 as indicated on attached Technical Specification page changes.
C. Change Section 3.10.3.a as indicated on attached Technical Specifica-tion page changes.
D. Add new Section 3.10.3.g as indicated on attached Technical Specifica-tion page changes.
E. Change Basis for Section 3.10 as indicated on attached Technical Specification page changes".
F. Add new Section 3.lLg as indicated on attached.Technical Specifica-tion page changes.
G. Delete Figures 3-7 and 3-8 and Replace Figure 3-9 with new 3-9 included in attached Technical Specification page changes.
II. Discussion The primary purpose of these Technical Specification changes are to modif'y the axial limitations on linear heat rate. The previous limit was based on a single analyzed axial shape, while the new limit is based on a sensi-tivity study involving both LOCA and DNBR analyses performed by Exxon Nuclear (see XN-NF-78-16, "Analysis of Axial Power Distribution Limits for the Palisades Nuclear Reactor at 2530 MWt").
A. The definitions added to Section 1.1 to be consistent with assumptions used in the Exxon sensitivity study (XN.;;NF-78-16).
B. The references in Se.ction 2 .1 have been deleted and the appropriate information included in the new version of Section 3.10.3.
1
C.
The changes to Section 3.10.3.a incorporate the results of the axial power distribution sensitivity study as presented in XN-NF-78-16. There are different limits for ENC type and D type fuel because of the different number of fuel rods and the different engineering factors assumed in the analysis. The maximum LHGR
~or D fuel is the same as the current Technical Specification.
D. The new Section 3.10.3.g provides limits for core radial peaking factors consistent with those used in the LOCA and DNB analysis in XN-NF-78-16.
E. Changes to the Basis for Section 3.10 have been made to be consistent with the Exxon analysis (XN-NF-78-16).
F. The new Section 3.11.g has been added to when radial peaking factors are to be determined and what action is necessary if the limits of 3.10.3.g are exceeded.
G. With the new Exxon analysis (XN-NF-78-16) and the inclusion in 3.10.3.a of various iincertainty factors Figures 3-7 and 3-8 are no longer required. The new Figure 3-9 is identical to Figure 2-1 of the Exxon report XN-NF-78-16.
III. Conclusion(s)
Based on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed these changes and recommend their approval.
CONSUMERS POWER COMPANY By ---~~---~~---....-__;.7'9.__-
C R Bilby, Vice Pre Production & Transmi Sworn and subscribed to before me this 15 day of June 1978.
L'Ild8RThayer )NC>ta:Public Jackson County, Michigan May commission expires July 9, 1979.
1.1 REACTOR OPERATING CONDITIONS (Contd)
Low Power Physics Testing Testing performed under approved written procedures to determine control rod worths and other core nuclear properties. Reactor power during these tests shall not exceed lo-2% of rated power, not including decay heat and primary system temperature and pressure shall be in the range of 260°F to 538°F and 415 psia to 2150 psia, respectively. Certain devia-tions from normal operating practice which are necessary to enable per-forming some of these tests are permitted in accordance with the specific provisions.therefore in these Technical Specifications.
Shutdown Boron Concentrations Boron concentration sufficient to provide keff S 0.98 with all control rods in the core and the highest worth control rod fully withdrawn.
Refueling Boron Concentration Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 5% 6p with all control rods withdrawn).
Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.
Assembly Radial Peaking Factor ~ FrA The assembly radial peaking factor is the maximum ratio of individual fuel assembly power to core aver.age assembly power integrated over the total core height, including tilt.
Total Radial Peaking Factor - FrT The total radial peaking factor is the maximum product of the ~atio of individual assembly power to core average assembly power times the local peaking factor for that assembly*integrated over the total core height, including tilt. Local peaking factor is defined as the maximum ratio of the power in an individual fuel rod to assembly average rod power.
1.2 PROTECTIVE SYSTEMS Instrument Channels One of four independent measurement channels, complete with the sensors, sensor power supply units, amplifiers and bistable modules provided for each safety parameter.
Reactor Trip The de-energizing of the control rod drive mechanism (CRDM) magnetic clutch holding coils which releases the control rods and allows them to drop into the core.
1-2
1.2 PROTECTIVE SYSTEMS (Contd)
Reactor Protective System Logic This system utilizes relay contact outputs from individual instrument channels to provide the reactor trip signal for de-energizing the mag-netic clutch power supplies. The logic system is wired to provide a reactor trip on a 2-of-4 or 2-of-3 basis for any given input parameter.
Degree of Redundancy The difference between the number of operable channels and the number of channels. which when tripped will cause an automatic system trip.
Engineered Safety Features System Logic This system utilizes relay contact outputs from.individual instrument channels to provide a dual channel (right and left) signal to initiate independently the actuation of engineered safety feature equipment con-nected to diesel generator* 1-2 (right channel) and. diesel generator 1-1 (left channel). The logic system is wired to provide an appropriate sig-nal. for the actuation of the engineered safety feature equipment on a 2~of-4 basis for any given input parameter.
1.3 INSTRUMENTATION SURVEILLANCE Channel Check A qualitative determination of acceptable operability by observation of channel behavior during normal plant operation. This *determination shall, where feasible, include comparison of the channel with other independent channels measuring the same variable.
Channel Functional Test Injection of a simulated signal into the channel to verify that it is operable, including any alarm and/or trip initiating action.
Channel Calibration Adjustment of channel. output such that it responds, with acceptable range and.accuracy, to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment action, alarm, interlocks or trip and shall be deemed to include the channel func-tional test.
1.4 MISCELLANEOUS DEFINITIONS Operable A system or _component is operable if it is capable of fulfilling its de-sign functions.
Operating A system or component is operating if it is performing its design functions.
1-3
\.
1.4 MISCELLANEOUS DEFINITIONS (Contd)
Control Rods All full-length shutdown and regulating rods.
Containment. Integrity Containment integrity is defined to exist when all of the following are true:
- a. All nonautomatic containment isolation valves and blind flanges are closed.
- b. The equipment door is properly closed and sealed.
- c. At least one door in each personnel air lock is properly closed and sealed.
- d. All automatic containment isolation valves are operable or are locked closed.
- e. The uncontrolled containment leakage satisfies Specification 4.5.1.
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration.of I-131 (µC/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."
E - AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant ..
Safety Safety as used in these Technical Specifications refers to those safety issues related to the nuclear process and for example does not encompass OSHA considerations.
1-4
2.1 SAFETY LIMITS - REACTOR CORE (Contd) probability at a 95% confidence level than DNB will not occur which is considered an appropriate margin to DNB for all operating conditions. (l)
The curves of Figures 2-1, 2-2 and 2-3 represent the loci of points of thermal power, primary coolant system pressure and average temperature of various pump combinations for which the DNBR is ~ 1.3. The area of safe operation is below these lines. For 3- and 2-pump operation, the limiting condition is void fraction rather than DNBR. The void fraction limits assure stable flow and maintenance of DNBR greater than 1.3.
Flow maldistribution effects of operation under less than full primary 2
coolant flow have been evaluated via model tests. ( ) The flow model data established the maldistribution factors and hot channel inlet tem-peratures for the thermal analyses that were used to establish the safe operating envelopes presented in Figures 2-1 and 2-2. These figures were established on the basis that the thermal margin for part-loop operation should.be equal to or greater than the thermal margin for normal opera-tion.
The reactor protective system is designed to prevent any anticipated combination of transient conditions for primary coolant system tempera-ture, pressure and thermal power level that would result in a DNBR of less than 1.3.(3)
References (l)FSAR, Section 3,3,3,5.
( 2 )FSAR, Section 3,3,3,3, Appendix C.
( 3 )FSAR, Section 14.1.
2-2
3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to operation of control rods and hot channel factors during op-eration.
Objective To specify limits of control rod movement to assure an acceptable power distribution during power operation, limit worth of individual rods to values analyzed for accident conditions, maintain adequate shutdown margin after a reactor trip and to specify acceptable power limits for power tilt conditions.
Specifications 3.10.1 Shutdown Margin Requirements
- a. With four primary coolant pumps in operation at hot shutdown and above, the shutdown margin shall be 2%.
- b. With les.s than four primary coolant pumps in operation at hot shut-down and above, the shutdown margin shall be 3.75%.
- c. At less than the hot shutdown. condition, boron concentration shall be shutdown boron concentration.
- d. If a control rod cannot be tripped, shutdown margin shall be increased by boration as necessary to compensate for the worth of the withdrawn inoperable rod.
- e. The drop time of each control rod shall be no greater than 2.5 sec-onds from the beginning of rod motion to 90% insertion.
3.10.2 Individual Rod Worth.*
- a. The maximum worth of any one rod in the core at rated power shall be equal to or less than 0.6% in reactivity.
- b. The maximum worth of any one rod in the core at zero power shall be equal to or less than 1.2% in reactivity.
3.10.3 Power Distribution Limits
- a. The linear heat generation rate. at the peak power elevation z shall not exceed:
15.28 kW/ft x FA(Z) for ENC fuel types 14.12 kW/ft x FA(Z) for D type fuel where the function FA(Z) is shown in Figure 3.9. If the power dis-tribution is double peaked, both peaks shall satisfy the criterion.
Appropriate consideration shall be given to the fo.llowing factors:
(1) A flux peaking augmentation factor of 1.0, (2) A measurement calculational uncertainty factor of 1.10, -
3-58
3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd) 3.10.3 Power Distribution Limits (Contd)
(3) An engineering uncertainty factor (which includes fuel column shortening due to densification and thermal expansion) of:
1.03 for ENC fuel types and 1.05 for D-type fuel (4) A thermal power measurement uncertainty factor of 1.02.
- b. If the quadrant to core average power tilt exceeds 15%, except for physics tests, then:
(1) The linear heat generation rate shall promptly be demonstrated to be less than that specified in Part a, or (2) Immediate action shall be initiated to reduce reactor power to 75% or less of rated power.
- c. If the power in a quadrant exceeds core average by 10% for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the power in a quadrant exceeds. core average by 20% at any time, immediate action shall be initiated to reduce reactor power below 50% until the situation is remedied.
- d. If the power in. a.quadrant. exceeds the core average by 15% and if the linear heat generation rate cannot be demonstrated promptly to be within limits, then the overpower trip set point shall be reduced to 80% and the thermal margin low-pressure trip set point (P )
Trip shall be increased by 400 psi.
- e. If the power in a quadrant exceeds core average-by 5% for a period of 30 days, immediate action shall be.initiated to reduce reactor power to 75% or less of rated power.
- f. The part-length control rods will be completely withdrawn from the core (except for rod exercises and physics tests).
- g. The calculated value of FA shall be limited to < 1.45 + P and the r -
calculated value of FT shall be. limited to < 1.77 + P, where Pis r -
- the core thermal power in fraction of core rated thermal power
( 2 ' 530 Mwt ) .
3-59
3.10 CONTROL ROD .AND POWER DISTRIBUTION LIMITS (Contd)
Basis (Contd) rotor will not exceed acceptable limits. ( 5 ) The axial power distribution term ensures that the operating power distribution is enveloped by the design power distributions. Appropriate factors for measurement-calculational uncertainty, engineering factor and shortening of the fuel pellet* stack are specified to ensure that the linear heat gener-ation rate limit is not exceeded.
When a flux tilt exists for a sustained time period (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and can-not be corrected or if a .. flux tilt reaches 20%, reactor power will be*
reduced until the tilt can be corrected. A quadrant to core average power tilt may be indicated by two methods: Comparison of the output of the upper or lower sections of the ion chamber with the average value .and in-core detectors. ( 3 ) These values will form the basis for the calculat.ion of peaking factors. Calibration of the out-of-core detectors will take into account the local and total power distribution.
The insertion of part-length rods into the core, except for rod exer-cises or physics tests, is not permitted since it has been demonstrated on other CE plants that design power distribution envelopes can, under some circumstances, be violated by using part-length rods. Further in-formation may justify their use. Part-length rod insertion is permitted for physics tests, since resulting power distributions are closely mon-itored under test conditions. Part-length rod insertion for rod exer-cises (approximately 6 inches) is permitted since this amount of insertion has an insignificant effect on power distribution.
For a control rod misaligned up to 8 inches from the remainder of the banks, ~ot channel factors will be well within design limits. If a control rod is misaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met. If in-core detectors are not avail-able to measure power distribution and rod misalignments > 8 inches exist, then reactor power must not exceed 75% of rated power to insure that hot channel conditions are met.
The limitations on ~ and FT are provided to ensure that the assumptions r r .
used in the analysis for establishing the DNB margin, linear heat rate, thermal margin/low pressure and high power trip set points remain valid during operation at the various allowable control rod group insertion limits.
3-63
.~
3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd)
Continued operation with that rod fully inserted will only be permitted if the hot channel factors, shutdown margin and.ejected rod worth limits are satisfied.
In the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be maintained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations permitted by Specification 3.10.7 are required in order that the control rod worth values used in the reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations will only be in effect for the* time period required for the test being performed. The testing interval during which these deviations will be in effect will be kept to a minimum and special operating precaµ-
tions will be in effect during these deviations.in accordance with approved written testing procedures.
Violation of the power dependent insertion limits, when it is necessary to rapidly reduce power to avoid or minimize a situation harmful to plant personnel or equipment, is acceptable due to the brief period of time that such a violation would be expected to exist, and due to the fact that it is unlikely that core operating limits such as thermal margin and shutdown mar-gin would be violated as a result of the rapid rod insertion. Core thermal margin will actually increase as a result of the rapid rod insertion .. In addition, the required shutdown margin will most likely not be violated as a result of the rapid rod insertion because present power dependent inser-tion limits result in shutdown margin in excess of that required by the safety analysis. ( 5 )
References (1) FSAR, Section 14.
(2) FSAR, Section 3.3.3.
(3) FSAR, Sectiqn 7.4.2.2.
(4) FSAR, SEction 7.3.3.6.
(5) XN-NF-77-18.
(6) XN-NF-77-24.
3-64
3.11 IN-CORE INSTRUMENTATION (Contd)
Specification (Contd) a 10-hour period) at least each two hours thereafter or the reactor power level shall be reduced to less than 50% of rated power (65% of rated power if no dropped or misaligned rods are present). If read-
- ings indicate a local power level equal to or greater than the alarm set point, the action specified in 3.11.b shall be taken.
g . . FA and FT shall be determined whenever the core power distribution r r is evaluated. If either F A or F T is found to be in excess of the r r
-j limit specified in Section 3 .10. 3 .g; within \-sixhour-s--fheraaT*p-ower~-~~=--~-~= __ _
shall be reduced to less than [(1.77
- FrT) -~--253~-~J-~~-T(~.45-~
A Fr ) x 2530 MWt], whichever is lower.
Basis A system of 45 in-core flux detector and thermocouple assemblies and a data display, alarm and record functions has been provided. (l) The out-of ... core nuclear instrumentation calibration includes:
- a. Calibration (axial and azimuthal) of.the s'plit detectors at initial re-actor start-up and during the power escalation program.
- b. A comparison check with the in-core instrumentation in the event abnormal readings are observed on the out-of-core detectors during operation.
- c. Calibration check during subsequent reactor start-ups.
- d. Confirm that readi_ngs from the out-of-core split detectors are as expected.
Core power distribution verification includes:
- a. Measurement at initial reactor start-up to check that power distribu-tion is consistent with calculations.
- b. Subsequent checks during operation to insure that power distribution is consistent with calculations.
- c. Indication of power distribution in the event that abnormal situa-tions occur duri_ng reactor operation.
If the data logger for the in-core readout is not in operation for more than two hours, power will be reduced to provide margin between the actual __
peak linear heat generation rates and the limit and the in 7 cor~ readings will be manually collected at the terminal blocks in the control room utilizing a suitable signal detector. If this is not feasible with the 3-66
3.11 IN-CORE INSTRUMENTATION (Contd)
Basis (Contd) manpower available, the reactor power will be reduced further to minimize the probability of exceeding the peaking factors. The time interval of two hours and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distri-bution to detect significant changes until the data logger is returned to service.
Reference (1) FSAR, Section 7.4.2.4.
3-66a
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Break Points:
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- 3. (1.0, .67) 0.65 O. l 0.2 0.3 0.4 0.5 0'.6 0.7 0.8 0.9 1.0 Location of Axial Power Peak (Fraction of Active Fuel Height)
Allo~able LHGR as a function Palisades Figure of peak power location. fochuical Specifications 3-9