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MONTHYEARML16327A0682016-11-18018 November 2016 Issuance of the Core Operating Limits Report for Reload 21, Cycle 22, Revision 11 Project stage: Request ML17139D3572017-05-19019 May 2017 License Amendment Request Revise Technical Specifications Section 3.4.3 for the Remainder of the Current Operating Cycle Project stage: Request ML17158B2792017-06-0707 June 2017 Acceptance Review - Peach Bottom Unit 2 - License Amendment Request Regarding Safety Relief Valve and Safety Valve Operability for Cycle 22 Project stage: Acceptance Review ML17222A0962017-08-10010 August 2017 Request for Additional Information - Amendment Request Regarding Safety Relief Valve and Safety Valce Operability for Cycle 22 Project stage: RAI ML17241A1942017-08-29029 August 2017 License Amendment Request - Revise Technical Specifications Section 3.4.3 (Srvs/Svs) for the Remainder of the Current Operating Cycle for Unit 2 - Supplement 1 Response to Request for Additional Information Project stage: Supplement ML17249A1512017-10-25025 October 2017 Issuance of Amendment Safety Relief Valve and Safety Valve Operability for Cycle 22 (CAC No. MF9705; EPID L-2017-LLA-0229) Project stage: Approval 2017-05-19
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Category:Letter
MONTHYEARML24022A2262024-01-22022 January 2024 (PBAPS) Unit 3 - Submittal of the Inservice Inspection (ISI) Owner'S Activity Report (OAR) for the 24th Refueling Outage for Unit 3 ML23346A2412024-01-17017 January 2024 Staff Assessment of Updated Seismic Hazards Following the NRC Process for the Ongoing Assessment of Natural Hazards Information (EPID L-2019-PMP-0061) - Letter IR 05000277/20234022024-01-11011 January 2024 Security Baseline Inspection Report 05000277/2023402 and 05000278/2023402 and Independent Spent Fuel Storage Installation Security Inspection 07200029/2023401 ML23354A1922024-01-0505 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) ML23355A0622023-12-18018 December 2023 Independent Spent Fuel Storage Installation (Isfsi), Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 ML23334A2502023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23270C0072023-11-29029 November 2023 Notice of Proposed Amendment to Decommissioning Trust Agreement ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums ML23305A3472023-11-0101 November 2023 Revision of One Form That Implements the Radiological Emergency Response Plan (RERP) IR 05000277/20230032023-10-30030 October 2023 Integrated Inspection Report 05000277/2023003 and 05000278/2023003 and Exercise of Enforcement Discretion ML23286A0012023-10-13013 October 2023 Reply to a Notice of Violation (NOV) - NOV 05000277, 05000278/2023010-01 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans ML23271A1342023-10-11011 October 2023 Dtf Analysis Letter IR 05000277/20244012023-10-0505 October 2023 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000277/2024401 and 05000278/2024401 ML23265A1502023-09-22022 September 2023 License Amendment Request Revise a License Condition and Certain Technical Specifications to Remove Restrictions in Support of Decommissioning Activities ML23264A8682023-09-21021 September 2023 Independent Spent Fuel Storage Installation Registration of Dry Storage Canister WC-EOS-37PTH-008-B4L, WC-EOS-37PTH-007-B4L, and WC-EOS-37PTH-002-B4L IR 05000277/20230102023-09-15015 September 2023 Biennial Problem Identification and Resolution Inspection Report 05000277/2023010 and 05000278/2023010 and Notice of Violation ML23251A2022023-09-11011 September 2023 Cover Letter W/Report LR-N23-0045, and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement2023-09-0808 September 2023 and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement IR 05000277/20230052023-08-31031 August 2023 Updated Inspection Plan for Peach Bottom Atomic Power Station, Units 2 and 3 (Report 05000277/2023005 and 05000278/2023005) ML23158A1952023-08-30030 August 2023 R. E. Ginna - Issuance of Amendments to Adopt TSTF-273-A, Revision 2, Safety Function Determination Program Clarifications ML23232A0012023-08-19019 August 2023 Revision of One Procedure That Implements the Radiological Emergency Response Plan (Rerp). Includes EPP 06-009, Rev. 13 IR 05000277/20230112023-08-15015 August 2023 Comprehensive Engineering Team Inspection Report 05000277/2023011 and 05000278/2023011 ML23226A1112023-08-14014 August 2023 10 CFR 50.46 Annual Report IR 05000277/20230022023-08-0808 August 2023 Integrated Inspection Report 05000277/2023002 and 05000278/2023002 RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor IR 05000277/20230402023-07-17017 July 2023 95001 Inspection Supplemental Report 05000277/2023040 and Follow-Up Assessment Letter ML23188A0532023-07-0707 July 2023 Registration of Use of Independent Spent Fuel Storage Installation (ISFSI) Casks ML23186A1272023-07-0606 July 2023 Response to Disputed Non-Cited Violation Documented in Peach Bottom Atomic Power Station, Units 2 and 3, Inspection Report 05000277/2022004 and 05000278/2022004 ML23178A0352023-06-27027 June 2023 Senior Reactor and Reactor Operator Initial License Examinations RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23151A4162023-05-31031 May 2023 Revision of One Form That Implements the Radiological Emergency Response Plan (RERP) for Wolf Creek Generating Station (WCGS) ML23150A2052023-05-30030 May 2023 Independent Spent Fuel Storage Installation (Isfsi), Annual Radiological Environmental Operating Report 80 January 1, 2022 Through December 31, 2022 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 ML23142A1602023-05-23023 May 2023 Summary of April 11, 2023, Public Observation Meeting with Constellation Energy Generation, LLC Re Proposed Exemption Request from Work Hour Requirements ML23132A1582023-05-18018 May 2023 Correction to Technical Specification 3.4.7 (Page 3.4-16) Issued in Amendment No. 337 ML23132A1532023-05-12012 May 2023 Transmittal of 2022 Annual Financial Reports WM 23-0009, Financial Assurance Requirements for Decommissioning Nuclear Power Reactors 10 CFR 50.75(f)(1)2023-05-10010 May 2023 Financial Assurance Requirements for Decommissioning Nuclear Power Reactors 10 CFR 50.75(f)(1) IR 05000277/20230012023-05-0808 May 2023 Integrated Inspection Report 05000277/2023001 and 05000278/2023001 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI 2024-01-05
[Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50
MONTHYEARML23265A1502023-09-22022 September 2023 License Amendment Request Revise a License Condition and Certain Technical Specifications to Remove Restrictions in Support of Decommissioning Activities LR-N23-0045, and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement2023-09-0808 September 2023 and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement 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 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 ML21119A1412021-04-29029 April 2021 License Amendment Request, Revise Technical Specifications Administrative Controls Section 5.5.7 Regarding the Frequency for Performing Certain Ventilation Filter Testing 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 RS-21-016, Supplement to Application to Revise Technical Specifications to Adopt TSTF-566, Revise Actions for Inoperable RHR Shutdown Cooling Systems2021-01-29029 January 2021 Supplement to Application to Revise Technical Specifications to Adopt TSTF-566, Revise Actions for Inoperable RHR Shutdown Cooling Systems ML20337A3012020-12-0202 December 2020 Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf. ML20150A0072020-05-29029 May 2020 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 RS-20-031, License Amendment Request - Application to Adopt Technical Specification Task Force (TSTF) 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 Item Improvement Process2020-04-30030 April 2020 License Amendment Request - Application to Adopt Technical Specification Task Force (TSTF) 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 Item Improvement Process ML20024G4292020-03-0404 March 2020 Record of Decision - Issuance of Subsequent Renewed Facility Operation License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station, Units 2 and 3 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 ML19193A0062019-07-12012 July 2019 Supplement No. 7 - Changes to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application ML19182A1122019-07-0101 July 2019 First 10 CFR 54.21 (B) Annual Amendment to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application 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 ML19158A3122019-06-0707 June 2019 License Amendment Request to Revise Technical Specifications 3.8.4, DC Sources - Operating ML19116A1962019-04-26026 April 2019 License Amendment Request to Revise Technical Specifications 3.8.1, Required Action A.3, for Temporary One-Time Extension of Completion Time ML19042A1312019-02-11011 February 2019 Supplement No. 3 - Changes to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application RS-19-003, Application to Adopt TSTF-564, Safety Limit Mcpr.2019-02-0101 February 2019 Application to Adopt TSTF-564, Safety Limit Mcpr. ML19023A0152019-01-23023 January 2019 Supplement No. 2 - Changes to the Subsequent License Renewal Application ML18305B2702018-11-0101 November 2018 Response to Request for Additional Information - License Amendment Request for Approval of Changes to Emergency Plan Staffing Requirements ML18275A0232018-09-28028 September 2018 License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of ... ML18271A0092018-09-27027 September 2018 License Amendment Request -Technical Specifications Section 3.3.6.2 Functions 3 and 4 Applicability Changes Pertaining to Reactor Building and Refueling Floor Ventilation ML18257A1432018-09-14014 September 2018 Changes to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application ML18239A3552018-08-27027 August 2018 License Amendment Request - Expanded Actions for LEFM Conditions ML18213A2712018-08-0101 August 2018 Application to Revise Technical Specifications to Adopt TSTF-500, Revision 2, DC Electrical Rewrite Update to TSTF-360, - Supplement 1, Revision of Implementation Dates. ML18193A6972018-07-10010 July 2018 Cover Letter Submitting Subsequent License Renewal Application ML18193A6992018-07-10010 July 2018 Cover Letter Submitting Paper Copies, Cds and Boundary Drawings Accompany Submittal of Subsequent License Renewal Application 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 ML18149A2992018-05-10010 May 2018 Enclosure 2 - Peach Bottom Staffing LAR - Evaluation of Proposed Changes and Attachments 2A & 2B, Emergency Plan Marked-Up Pages and Emergency Plan Clean Copy Pages ML18149A3022018-05-10010 May 2018 Enclosure 4 - Summary of Regulatory Commitments 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) ML17275A0692017-09-29029 September 2017 Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-500, Revision 2, DC Electrical Rewrite-Update to TSTF-360. ML17258A1792017-09-15015 September 2017 Submittal of Measurement Uncertainty Recapture Power Uprate License Amendment Request - Supplement 5, Supplemental Reload Licensing Report ML17243A0142017-08-30030 August 2017 Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors. ML17241A1942017-08-29029 August 2017 License Amendment Request - Revise Technical Specifications Section 3.4.3 (Srvs/Svs) for the Remainder of the Current Operating Cycle for Unit 2 - Supplement 1 Response to Request for Additional Information ML17139D3572017-05-19019 May 2017 License Amendment Request Revise Technical Specifications Section 3.4.3 for the Remainder of the Current Operating Cycle RS-17-049, Application to Revise Technical Specifications to Adopt TSTF-529 Clarify Use and Application Rules, Revision 4, Using the Consolidated Line Item Improvement Process2017-03-28028 March 2017 Application to Revise Technical Specifications to Adopt TSTF-529 Clarify Use and Application Rules, Revision 4, Using the Consolidated Line Item Improvement Process ML17080A0672017-03-20020 March 2017 Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate - Supplement 1 - Request for Non-Proprietary Version of Cameron Corporation Proprietary Documents ML17048A4442017-02-17017 February 2017 Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate ML16342C4552016-12-0707 December 2016 Supplement to License Amendment Request, Revise Turbine Condenser - Low Vacuum Scram Instrumentation Function Allowable Value ML16309A2982016-11-0404 November 2016 Request for License Amendment to Revise Turbine Condenser - Low Vacuum Scram Instrumentation Function Allowable Value ML16280A4022016-10-0606 October 2016 License Amendment Request - Supplement 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. ML16209A2182016-07-26026 July 2016 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing. ML16173A3712016-06-20020 June 2016 Application to Revise Technical Specifications to Adopt TSTF-501-A, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control, Revision 1, Using the Consolidated Line Item Improvement Process ML16132A4402016-05-11011 May 2016 Supplement to License Amendment Request, Revise Surveillance Requirement 3.5.1.5 Involving Recirculation Pump Discharge Valves 2023-09-08
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARCCN:23-09, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information ML22312A3642022-11-0808 November 2022 Response to NRC Inspection Report and Preliminary White Finding ML22145A0852022-05-25025 May 2022 Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel N-16A Nozzle Repair RS-22-027, Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated2022-02-23023 February 2022 Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated P NMP1L3447, Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs2022-02-0202 February 2022 Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs 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-21-0044, 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-21-0032, Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting2021-04-20020 April 2021 Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting ML21035A0332021-02-0404 February 2021 Response to Request for Additional Information - 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 ML21029A1662021-01-29029 January 2021 Response to Request for Additional Information - 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 RS-20-112, Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-5682020-09-0303 September 2020 Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-568 ML20224A5302020-08-11011 August 2020 Response to the Order for Implementation of Additional Security Measures and Fingerprinting for Unescorted Access for Wolf Creek Operating Corporation Independent Spent Fuel Storage Installation ML20188A2642020-07-0606 July 2020 Clinton Power Station, R.E. Ginna Station, Limerick Station, Nine Mile Point Station & Peach Bottom Station - Proposed Alternative to Utilize Code Case OMN-26 - Response to Request for Additional Information ML19276F2812019-10-0303 October 2019 Supplemental Response to Request for Additional Information (RAl-1) License Amendment Request to Revise Technical Specifications 3.8.4, DC Sources-Operating ML19241A4652019-08-29029 August 2019 Response to Request for Additional Information (RAl-1) License Amendment Request to Revise Technical Specifications 3.8.4, DC Sources-Operating ML19225B9762019-08-13013 August 2019 Response to NRC Request for Clarification of Information, Dated August 1, 2019 Related to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application ML19206A0562019-07-24024 July 2019 Stations, Units 2 and 3 - Response to Request for Additional Information - License Amendment Request to Revise Technical Specifications 3.8.1, Required Action A.3, for Temporary One-Time Extension of Completion Time ML19163A2232019-06-12012 June 2019 Revised Response to NRC Request for Additional Information, Core Shroud Support Fatigue Analysis Reevaluation, Related to the Subsequent License Renewal Application ML19157A0092019-06-0606 June 2019 Response to NRC Request for Additional Information, Dated May 15, 2019 Related to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application JAFP-19-0057, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-06-0404 June 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 ML19150A2972019-05-30030 May 2019 Revised Responses to NRC Requests for Additional Information, Fire Water System, Related to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application ML19122A2892019-05-0202 May 2019 Response to NRC Requests for Additional Information, Set 1, Dated April 10, 2019 Related to Plant Subsequent License Renewal Application NMP1L3279, Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants2019-05-0101 May 2019 Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants ML19085A3852019-03-26026 March 2019 Response to Second Request for Additional Information - License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications ... ML19065A0082019-03-0505 March 2019 Response to Nrg Audit Review Information Request - Application for Subsequent Renewed Operating Licenses Section 4.6 Primary Containment Fatigue Analyses ML19046A1292019-02-15015 February 2019 Response to RAI - License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times ML19028A2802019-01-28028 January 2019 Response to NRC Request for Additional Information, Dated December 13, 2018, Regarding Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Severe Accident Mitigation Alternatives Requests for Additional Information JAFP-19-0006, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-01-0808 January 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 ML19007A3262019-01-0707 January 2019 Response to Request Dated December 7, 2018 for Docketing of Additional Documents to Support Nrc'S Environmental Review of the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application ML18354B0612018-12-20020 December 2018 Response to NRC Request for Additional Information, Dated November 23, 2018, Regarding the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application, Environmental Requests for Additional Information ML18340A1852018-12-0606 December 2018 Response to Request for Additional Information License Amendment Request - Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to Be Out-of-Service with Increased Reactor Pressure Safety Limit RS-18-143, Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements2018-11-29029 November 2018 Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements ML18337A2402018-11-28028 November 2018 Response to Request for Additional Information - Relief Requests Associated with the Fifth Inservice Inspection Interval ML18305B2702018-11-0101 November 2018 Response to Request for Additional Information - License Amendment Request for Approval of Changes to Emergency Plan Staffing Requirements ML18257A0172018-09-14014 September 2018 Response to Request for Additional Information Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-500, Revision 2, DC Electrical Rewrite Update to ... ML18222A3872018-08-10010 August 2018 Response to Request for Additional Information and Supplemental Information - Application to Adopt 1 O CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components ... ML18128A0092018-05-0707 May 2018 Response to Request for Additional Information Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants. RS-18-061, Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis)2018-05-0202 May 2018 Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis) ML17354A3812017-12-20020 December 2017 Supplemental Response Concerning License Amendment Request to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 RS-17-149, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from The.2017-12-15015 December 2017 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from The. ML17241A1942017-08-29029 August 2017 License Amendment Request - Revise Technical Specifications Section 3.4.3 (Srvs/Svs) for the Remainder of the Current Operating Cycle for Unit 2 - Supplement 1 Response to Request for Additional Information ML17223A6262017-08-11011 August 2017 Response to Request for Additional Information Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 ML17220A2142017-08-0808 August 2017 Measurement Uncertainty Recapture License Amendment Request - Supplement 3 Response to Request for Additional Information ML17195A2852017-07-13013 July 2017 Measurement Uncertainty Recapture License Amendment Request - Supplement 2 Response to Request for Additional Information RS-17-044, Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography2017-03-13013 March 2017 Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML16224A3422016-08-11011 August 2016 Response to Request for Additional Information Regarding Proposed License Amendment Concerning Diesel Generator Lube Oil Inventory (TSTF-501-A) for Peach Bottom Atomic Power Station, Units 2 and 3 ML16216A0762016-08-0202 August 2016 Response to Draft Request for Additional Information Regarding the Proposed License Amendment Request Concerning Emergency Diesel Generator Fuel Oil Transfer Surveillance Requirements ML16162A1012016-06-0909 June 2016 Response to Request for Additional Information Regarding Proposed License Amendment Concerning Emergency Diesel Generator Fuel Oil Transfer Surveillance Requirements RA-16-049, Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure2016-05-26026 May 2016 Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure ML16083A3942016-03-23023 March 2016 Changes to High Pressure Coolant Injection and Reactor Core Isolation Cooling Surveillance Requirements - Supplement 1, Response to Request for Additional Information 2023-11-10
[Table view] |
Text
Exelon Generation ~
200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 August 29, 2017 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Unit 2 Renewed Facility Operating License No. DPR-44 NRC Docket No. 50-277
Subject:
License Amendment Request- Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2 -
Supplement 1 Response to Request for Additional Information
References:
- 1. Exelon letter to the NRC, "License Amendment Request -
Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2," dated May 19, 2017 (ADAMS Accession No. ML17139D357)
- 2. Email from R. Ennis (USNRC) to D. Neff (Exelon), "Peach Bottom Unit 2 - Request for Additional Information -
Amendment Request Regarding Safety Relief Valve and Safety Valve Operability for Cycle 22 (CAC MF9705)," dated August 10, 2017 (ADAMS Accession No. ML17222A096)
In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requested an amendment to Renewed Facility Operating License No. DPR-44 for Peach Bottom Atomic Power Station (PBAPS) Unit 2 (Reference 1). Specifically, the proposed change would revise the Technical Specifications (TSs) to decrease the number of safety relief valves (SRVs) and safety valves (SVs), required to be operable, when operating at a power level less than or equal to 3358 megawatts thermal (MWt). This change would be in effect for the current PBAPS Unit 2 operating cycle (Cycle 22) that is scheduled to end in October 2018.
During their technical review of the application, the NRC Staff identified the need for additional information. Reference 2 provided the Requests for Additional Information (RAls) from the NRC Reactor Systems Branch (SRXB). The attachment to this letter provides the responses to the RAls.
Exelon has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the U.S. Nuclear Regulatory Commission in Reference 1. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Further, the additional information provided in this
SRVOOS LAR Supplement 1 Response to Request for Additional Information August 29, 2017 Page2 submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania and the State of Maryland of this response by transmitting a copy of this letter to the designated State Officials.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact Mr. David Neff at (610) 765-5631.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of August 2017.
Respectfully,
~~~
Jam~arstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC
Attachment:
Response to Requests for Information from NRC Review Branch SRXB cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Pennsylvania Bureau of Radiation Protection S. T. Gray, State of Maryland
Attachment Peach Bottom Atomic Power Station, Unit 2 NRC Docket No. 50-277 Response to Request for Additional Information from NRC Review Branch SRXB
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 1of8 Responses to NRC Staff's Request for Additional Information By application dated May 19, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17139D357), Exelon Generation Company, LLC (Exelon) submitted a license amendment request (LAR) for Peach Bottom Atomic Power Station (PBAPS), Unit 2.
The amendment would revise the Technical Specifications (TSs) to decrease the number of safety relief valves (SRVs) and safety valves (SVs}, required to be operable, when operating at a power level less than or equal to 3358 megawatts thermal (MWt). This change would be in effect for the current PBAPS Unit 2 operating cycle (Cycle 22) that is scheduled to end in October 2018.
In an email dated August 10, 2017, from the NRC (Rick Ennis) to Exelon (David Neff) (ADAMS Accession No. ML17222A096}, the NRC provided Requests for Additional Information (RAls) seeking clarification of certain issues. In a phone call on August 10, 2017, Mr. Neff stated that Exelon would provide a response to the RAI questions by August 31, 2017.
SRXB-RAl-1: ASME Overpressure Analysis for New Operating Condition Overpressure protection for the reactor coolant pressure boundary (RCPB) during power operation is provided by SRVs and SVs and the reactor protection system. The NRC's acceptance criteria are based on: (1) draft GDC-9 1 , insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; and (2) final GDC-31, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a non-brittle manner and that the probability of rapidly propagating fracture is minimized.
The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level (two SRVs out*of-service (2SRVOOS), reactor powers 3358 MWt). In order to facilitate the NRC staff in performing an independent confirmatory calculation (simulating the limiting ASME overpressure event of all main steam isolation valve closures with scram on high neutron flux (MSIVF)), please provide the following information based on the current PBAPS Unit 2 reload licensing analysis (Cycle
- 22) and plant operating history:
- 1. Justification that the limiting ASME overpressure event is MSIVF for the new operating condition;
- 2. The main steam isolation valves' flow area (fraction of full open flow area) as a function of time during MSIVF event (LAR Figure 1);
1 As discussed in Appendix H to the PBAPS Updated Final Safety Analysis Report (UFSAR), during the construction/licensing process, PBAPS, Units 2 and 3, were evaluated against the then-current Atomic Energy Commission (AEC) draft of the 27 General Design Criteria (GDC) issued in November 1965. On July 11, 1967, the AEC published for public comment, in the Federal Register (32 FR 10213), a revised and expanded set of 70 draft GDC (hereinafter referred to as the "draft GDC"). Appendix H of the PBAPS UFSAR contains an evaluation of the design basis of PBAPS, Units 2 and 3, against the draft GDC. The licensee concluded that PBAPS, Units 2 and 3, conforms to the intent of the draft GDC. The licensees for PBAPS, Units 2 and 3, have made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other plant-specific design and licensing basis documentation.
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 2 of 8
- 3. The control rod insertion as a function of time during MSIVF event (LAR Figure 1); and
- 4. The test results for the last two refueling cycles' on SRVs/SVs lift setpoint tolerance.
RESPONSE
- 1. Two potentially limiting overpressure protection events are typically analyzed for BWRs: (1)
Main Steam Isolation Valve Closure with Scram on High Flux (MSIVF), and (2) Turbine Trip with Bypass Failure and Scram on High Flux. However, based on both plant initial core analyses and subsequent power uprate evaluations, the MSIVF is more limiting than the turbine trip (TI) event with respect to reactor overpressure. Closure of the MS IVs, which are closer to the reactor vessel than the turbine stop valves, reduces the amount of steam volume available to accommodate the pressurization (i.e., less steam line volume) compared to closure of the turbine stop valves and results in a higher pressurization rate and magnitude. PBAPS Extended Power Uprate (EPU) evaluations show a 40 to 50 psi difference between these two events. Operation at a reduced power level at PBAPS Unit 2 does not change the relative behavior/results of these two events and the MSIVF event will remain limiting for the new operating condition.
- 2. The Main Steam Isolation Valve (MSIV) flow area (fraction of full open flow) as a function of time assumed for the analysis of the MSIVF event is provided in the following table.
Closure Flow Area Time (sec) (fraction) 0.0 1.0 0.6 1.0 1.7 0.01 3.0 0.0
>3.0 0.0
- 3. The control rod (CR) insertion time assumed for the MSIVF event analysis is specified in PBAPS TS Table 3.1.4-1 (as a function of control rod notch position) and is provided in the table below (as a function of% insertion).
Scram Time CR Insertion (sec) (%)
0.0 0 0.2 0 0.49 5 0.9 20 2.0 50 3.5 90 3.875 100
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 3 of 8
- 4. The test results for the last two refueling cycles' on SRVs/SVs lift setpoint tolerance are provided in the tables below. The SRV/SV lift setpoints and setpoint tolerances are defined in TS Section 3.4.3, "Safety Relief Valves (SRVs) and Safety Valves (SVs)."
Fall 2016 PBAPS Unit 2 Refueling Outage Valve S/N Pressure As-Found Pressure Pressure Lift Result Set point Lift (psig) Setpoint Setpoint (PASS I (psig) Tolerance Deviation (psig FAIL)
(psig I%) I%)
20 1135 1135 +/- 34.1/+/-3% 010 PASS 21 1145 1124 +/- 34.4 / +/- 3% -21/-1 .8 PASS 22 1155 1150 +/-34.7/+/-3% -5 / -0.4 PASS 25 1155 1140 +/-34.7/+/-3% -15/-1.3 PASS SV14005 1260 1255 +/- 37.8 / +/- 3% -5 / -0.4 PASS Fall 2014 PBAPS Unit 2 Refueling Outage Valve S/N Pressure As-Found Pressure Pressure Lift Result Set point Lift Pressure Setpoint Setpoint (PASS I (psig) (psig) Tolerance Deviation (psig FAIL)
(psig I%) I%)
193 1135 1122 +/- 34.1/+/-3% -13/-1.1 PASS 83 1135 1141 +/- 34.1/+/-3% +61 +0.5 PASS 73 1145 1123 +/- 34.4 / +/- 3% -22/-1.9 PASS 88 1155 1137 +/-34.7/+/-3% -18/-1.6 PASS BL1095 1260 1212 +/- 37.8 / +/- 3% -48 / -3.8 FAIL BL 1104 1260 1272 +/-37.8 / +/-3% +12 / +1 .0 PASS SRXB-RAl-2: ATWS Overpressure Analysis for New Operating Condition An anticipated transient without scram (ATWS) is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs 14 and 15. Requirements related to ATWS events are specified in 10 CFR 50.62. The NRC staff review includes confirming that the peak reactor vessel bottom pressure will be less than the ASME Service Level C limit of 1500 psig during an ATWS overpressure event as protected by the SRVs/SVs and systems required in accordance with 10 CFR 50.62 (e.g., alternate rod injection system, standby liquid control system) .
The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level. In order to facilitate the NRC staff in performing an independent confirmatory calculation (simulating the limiting ATWS overpressure event), please provide the following information:
- 1. What is the limiting ATWS overpressure event for the proposed new operating condition (2SRVOOS, reactor powers 3358 MWt)? Please provide justification.
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 4 of 8
- 2. Clarify the differences among the following calculated peak reactor (vessel bottom) pressures:
(a) 1430 psig value discussed on pages 5 and 6 of Attachment 1 of the LAR; (b) 1419 psig value discussed on page 4 of Attachment 1 of the LAR and in Section 9.3.1.1 of NEDC-33720P 2 ; and (c) Value specified in the first sentence of Note 2 in Table 9-3 of NEDC-33720P.
Please confirm which one of the above is the licensing basis for current plant operation (i.e., 1SRVOOS, reactor powers 3951 MWt)?
- 3. The main steam isolation valves' flow area (fraction of full open flow area) as a function of time during MSIVC.
- 4. The sequence of events for the current ATWS overpressure licensing analysis (1SRVOOS, reactor powers 3951 MWt).
RESPONSE
- 1. The ATWS overpressure analysis is performed for two events - Main Steam Isolation Valve Closure (MSIVC) and Pressure Regulator Failure Open (PRFO). The PRFO event results in an MSIV closure after an initial reactor depressurization and the results are similar to the MSIVC event. The MSIVC event results in a higher peak reactor pressure by a small amount (12 psi). The Turbine Trip with Bypass Failure and Failure to Scram is not analyzed because, as noted in the response to SRXB-RAl-1 above, closure of the MSIVs, which are closer to the reactor vessel than the turbine stop valves, reduces the amount of steam volume available to accommodate the pressurization (i.e., less steam line volume) compared to closure of the turbine stop valves and results in a higher pressurization rate and magnitude. Operation at a reduced power level does not change the relative behavior/results of these two events and the MSIVC event will remain limiting. Therefore, the limiting ATWS overpressure event for the proposed new operating condition will remain the same (MSIVC).
- 2. The peak reactor pressure of 1430 psig value discussed on pages 5 and 6 of Attachment 1 of the LAR is the calculated result for the PBAPS Measurement Uncertainty Recapture (MUR) uprated condition (Reference 1) currently under NRC review. The peak reactor 2
GEH report NEDC-33720P, Revision 0, "Safety Analysis Report for Peach Bottom Atomic Power Station Units 2 & 3 Maximum Extended Load Line Limit Analysis Plus." was submitted to the NRC as Attachment 4 to the licensee's MELLLA+ application dated September 4, 2014 (ADAMS Accession No. ML14247A503). Attachment 4 is a proprietary (i.e., non-public) version of the report. A non-proprietary (i.e., public) version of the report is contained in to the MELLLA+ application.
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 5 of 8 pressure of 1419 psig value discussed on page 4 of Attachment 1 of the LAR and in Section 9.3.1 .1 of NEDC-33720P is the current analysis of record (licensing basis) result for the PBAPS CLTP conditions (EPU/MELLLA+).
Note 2 in Table 9-3 of NEDC-33720P refers to previous ATWS analyses performed to support implementation of PBAPS EPU prior to MELLLA+. The EPU ATWS overpressure analysis was performed utilizing the NRC approved ODYN code resulting in a peak reactor vessel pressure of 1461 psig. The ATWS overpressure analysis reported in NEDC-33720P was performed utilizing the NRC approved TRACG code. Note 2 in Table 9-3 of NEDC-33720P provides additional PBAPS EPU ATWS overpressure results utilizing the TRACG code to allow a direct comparison to the MELLLA+ results.
- 3. The MSIV flow area (fraction of full open flow) as a function of time assumed for the analysis of the MSIVC A TWS event using nominal values is provided in the following table. Since the ATWS event is an Anticipated Operational Occurrence (AOO), nominal initial conditions are used consistent with the NRC approved TRACG methodology.
Closure Flow Area Time (sec) (fraction) 0.0 1.0 0.8 1.0 3.8 0.01 4.0 0.0
>4.0 0.0 4 . The sequence of events for the current ATWS overpressure licensing analysis (MSIVC) is identified in the table below.
Seq. No. Event Response Time (sec) 1 MSIV Isolation Initiated 0.0 2 MSIVs Fully Closed 4.0 3 Recirculation Pump Trip 4.3 4 Qpenini:i of First Relief Valve 4.5 5 Peak Vessel Pressure 11.9 SRXB-RAl-3: Thermal Limits Assessment for New Operating Condition Draft GDCs 6, 14, 15, and 29 provide requirements related to core design and protection systems in order to assure that acceptable fuel damage limits are not exceeded. The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at reduced reactor power level (2SRVOOS, reactor powers 3358 MWt). To facilitate the staff's review, provide the following information based on the current PBAPS Unit 2 reload licensing analysis (Cycle 22):
- 1. What is the limiting AOO transient to determine the operating limit minimum critical power ratio (MCPR) multiplier, Kp, and power dependent linear heat generation rate (LHGR) multiplier (LHGRFAC(P)) for powers 85% of rated (reference Tables 4-2 and 5-3 of the PBAPS Unit 2, Cycle 22, core operating limits report (ADAMS Accession No. ML16327A068))?
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 6of8
- 2. What is the timing of the minimum MCPR and maximum LHGR and are there any SRVs required to lift prior to minimum MCPR and maximum LHGR for the limiting AOO transient identified above?
RESPONSE
- 1. The operating limit minimum critical power ratio (MCPR) multiplier, Kp, and power dependent linear heat generation rate (LHGR) multiplier (LHGRFAC(P)) for powers 85% of rated are determined by evaluating the limiting AOO transients (e.g., Load Rejection No Bypass [LRNBP], Feedwater Controller Failure (FWCF)) for each Equipment Out-of-Service (EOOS) combination. No one particular AOO transient establishes the Kp or LHGRFAC(P) values for all combinations or for all power levels within a combination. For instance, in the Base combination of PB2C22 Core Operating Limits Report (COLR) Table 4-2, the LRNBP is limiting at some power levels, but the FWCF is limiting at other power levels. In the Turbine Bypass System Out-of-Service (TBSOOS) combination, the FWCF is typically limiting for most, but not all, power levels. For the COLR Table 5-3, the FWCF is typically, but not always, limiting for LHGRFAC(P). The Kp and LHGRFAC(P) multipliers are developed in a manner that insures all AOO transients are bounded since no one AOO transient event is limiting at all off-rated conditions.
- 2. The most limiting MCPR and maximum LHGR values during pressurization events are calculated to occur prior to the opening of any SRVs or SVs. The limiting MCPR occurs approximately 0. 7-0.8 seconds prior to SRV opening and the maximum LHGR occurs approximately 1.1 seconds prior to SRV opening. Therefore, the most limiting MCPR and maximum LHGR values are not impacted by SRVs or SVs that may be out-of-service.
SRXB-RAl-4: Assessments of ECCS LOCA Performance and High Pressure System Performance for New Operating Condition The NRC's acceptance criteria related to emergency core cooling system (ECCS) and loss-of-coolant accident (LOCA) performance is based, in part, on: (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) draft GDCs 40 and 42, insofar as they require that protection be provided for engineered safety features against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; and (3) draft GDCs 37, 41, and 44, insofar as they require that a system to provide abundant emergency core cooling be provided so that fuel and clad damage that would interfere with the emergency core cooling function will be prevented.
The LAR proposed a new operating condition with decreasing the required number of SRVs/SVs while operating at a reduced reactor power level (2SRVOOS, reactor power s 3358 MWt). It is stated on page 8 of Attachment 1 to the LAR that the ECCS/LOCA performance and high pressure performance systems have been assessed for the proposed new operating condition. However, there are no further details provided regarding what the assessment involved and how the assessment had been performed. Please provide these details.
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 7 of 8
RESPONSE
ECCS/LOCA Performance The impact of SRV/SV setpoint tolerance relaxation and SRV/SV OOS on the ECCS/LOCA performance was addressed in Reference 2. Reference 2 included evaluations for Containment Pressure and Temperature for Design Basis Accident (DBA) Loss of Coolant Accident (LOCA),
Small Steam Line Breaks, Intermediate and Small Line Break Accidents, Suppression Pool Temperature, and DBA LOCA Hydrodynamic Loads. Those evaluations concluded that the ECCS/LOCA results were not affected by SRV/SV setpoint tolerance relaxation and/or an SRV/SV OOS. A review of the evaluation bases and conclusions in Reference 2 determined that the conclusions would not change for an additional SRV/SV OOS. In addition, as noted in Reference 2, PBAPS has been previously analyzed and licensed to operate with 2 SRV/SV OOS.
Reference 2 addressed the impact from operation with an SRVOOS on SRV dynamic loads.
The report states that having an SRVOOS does not have an effect on SRV dynamic loads.
SRV loads are driven by SRV opening pressure and the SRV discharge line water level at the time of the second SRV actuation. The SRV discharge line water level is in turn a function of the time between the closure of the valves at the end of the first actuation and the time of the second actuation. Having an SRVOOS will not affect the operable valves' setpoints or the time between the initial and second valve actuations. As such, having an SRVOOS has no effect on SRV dynamic loads. Safety Valves (SV) have no impact on suppression pool dynamic loads.
These conclusions do not change for an additional SRV/SV OOS.
Therefore, it is concluded that operation with an additional (up to 2) SRV/SV OOS would have no adverse effect on the ECCS/LOCA performance, including SRV dynamic loads.
High Pressure System Performance The high pressure systems include the High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC) and Standby Liquid Control (SLC) systems. The Reference 2 report states that the most potentially significant effect of an SRV/SV OOS and/or relaxing the SRV/SV setpoint tolerance on the HPCI and RCIC systems' operations is the maximum reactor pressure at which they are required to deliver water to the reactor. Both systems are required to provide injection into the reactor pressure vessel at the pressure corresponding to the lowest group of SRV lift setpoints (including drift) - PBAPS includes four SRVs in the lowest setpoint group.
The condition of an SRV/SV OOS does not change the SRV lift setpoints. This statement applies equally to either one or two SRV/SV OOS. Therefore, it is concluded that the maximum reactor pressure at which the HPCI and RCIC systems are required to deliver water to the reactor would not be adversely impacted by operation with an additional (up to two) SRV/SV OOS. In addition, as noted in Reference 2, PBAPS has been previously analyzed and licensed to operate with two SRV/SV OOS.
The limiting ATWS case for long term results, including peak reactor lower plenum pressure during SLC system injection, has been evaluated for PBAPS CLTP conditions and the more limiting MUR condition (Reference 1) currently under NRC review. For the MUR uprated condition, at the time of SLC system injection at 124 seconds for the limiting MSIVC case, the reactor power is in the range of approximately 25% power. This power level is well within the available steam relief system capacity, even with an additional (up to two) SRV/SV OOS. In
SRVOOS LAR Supplement 1 Attachment Responses to Requests for Information Page 8 of 8 addition, the MUR analysis determines that there is significant SLC system relief valve margin of 183 psi. This margin is significantly greater than the 70 psi minimum requirement. Therefore, it is concluded that peak reactor lower plenum pressure during SLC system injection would not increase significantly and the SLC system performance would not be adversely impacted by operation with an additional (up to two) SRV/SV OOS.
References
- 1. Exelon Letter to NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated February 17, 2017 (ADAMS Accession No. ML17048A444)
- 2. "PBAPS Units 2 and 3 Safety Valve Setpoint Tolerance Increase Safety Analysis Report,"
NEDC-33533P, Rev. 1, dated May 2013 (ADAMS Accession No. ML13162A619)