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Category:Letter
MONTHYEARIR 05000333/20230042024-02-0707 February 2024 Integrated Inspection Report 05000333/2023004 and Independent Spent Fuel Storage Installation Inspection Report 07200012/2023001 ML24037A0102024-02-0606 February 2024 Requalification Program Inspection ML24018A0012024-01-18018 January 2024 Notification of Commercial Grade Dedication Inspection (05000333/2024010) and Request for Information ML24004A2302024-01-0808 January 2024 Project Manager Reassignment ML23356A0832024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0058 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) 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 JAFP-23-0065, License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications2023-12-14014 December 2023 License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications IR 05000333/20234012023-12-0808 December 2023 Cybersecurity Inspection Report 05000333/2023401 (Cover Letter Only) RS-23-126, Request for Exemption from 10 CFR 2.109(b)2023-12-0707 December 2023 Request for Exemption from 10 CFR 2.109(b) JAFP-23-0069, Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements JAFP-23-0057, and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation JAFP-23-0064, Emergency Plan Document Revision2023-11-15015 November 2023 Emergency Plan Document Revision JAFP-23-0063, Registration of Spent Fuel Cask Use2023-11-13013 November 2023 Registration of Spent Fuel Cask Use IR 05000333/20230032023-11-13013 November 2023 Integrated Inspection Report 05000333/2023003 ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums IR 05000333/20230102023-10-26026 October 2023 Biennial Problem Identification and Resolution Inspection Report 05000333/2023010 JAFP-23-0059, Registration of Spent Fuel Cask Use2023-10-24024 October 2023 Registration of Spent Fuel Cask Use IR 05000333/20233012023-10-19019 October 2023 Initial Operator Licensing Examination Report 05000333/2023301 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 JAFP-23-0050, Physical Security Plan, Revision 242023-08-31031 August 2023 Physical Security Plan, Revision 24 JAFP-23-0048, Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-31031 August 2023 Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis IR 05000333/20230052023-08-31031 August 2023 Updated Inspection Plan for James A. FitzPatrick Nuclear Power Plant (Report 05000333/2023005) JAFP-23-0047, Correction to the 2022 Annual Radioactive Effluent Release Report2023-08-30030 August 2023 Correction to the 2022 Annual Radioactive Effluent Release Report ML23228A1342023-08-16016 August 2023 Licensed Operator Positive Fitness-For-Duty Test IR 05000333/20230022023-08-0707 August 2023 Integrated Inspection Report 05000333/2023002 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 JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0043, 10 CFR 50.46 Annual Report2023-07-31031 July 2023 10 CFR 50.46 Annual Report JAFP-23-0038, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance2023-07-28028 July 2023 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance ML23208A1622023-07-27027 July 2023 Operator Licensing Examination Approval IR 05000333/20234022023-07-26026 July 2023 Security Baseline Inspection Report 05000333/2023402 IR 05000333/20230112023-07-25025 July 2023 Post-Approval Site Inspection for License Renewal - Phase 4 Inspection Report 05000333/2023011 IR 05000333/20235012023-07-20020 July 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000333/2023501 JAFP-23-0033, License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation2023-06-28028 June 2023 License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation IR 05000333/20234202023-06-26026 June 2023 Security Baseline Inspection Report 05000333 2023420 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 ML23164A0322023-06-13013 June 2023 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000333/2023010 ML23152A0042023-06-0101 June 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection 05000333/2023401 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 JAFP-23-0025, 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023 2022 Annual Radiological Environmental Operating Report IR 05000333/20230012023-05-0303 May 2023 Integrated Inspection Report 05000333/2023001 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI JAFP-23-0023, 2022 Annual Radioactive Effluent Release Report2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report IR 05000333/20230122023-04-13013 April 2023 Quadrennial Fire Protection Inspection Report 05000333/2023012 ML23095A3722023-04-0505 April 2023 2023 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes Transmittal RS-23-049, Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-23023 March 2023 Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations IR 05000333/20220042023-03-20020 March 2023 Integrated Inspection Report 05000333/2022004 JAFP-23-0010, 2022 REIRS Transmittal of NRC Form 52023-03-20020 March 2023 2022 REIRS Transmittal of NRC Form 5 ML23061A1632023-03-0303 March 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch 3 2024-02-07
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARJAFP-22-0032, Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b..2022-06-16016 June 2022 Response to Request for Additional Information for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.. 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-22-0008, Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary2022-01-14014 January 2022 Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Eliminate Selected Response Time Testing for Reactor Protection System and Primary JAFP-21-0087, 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-21-0073, Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance.2021-08-0909 August 2021 Response to Request for Additional Information Regarding Application to Revise the James A. FitzPatrick Nuclear Power Plant Limiting Condition for Operation (LCO) 3.5.1, ECCS - Operating Surveillance. JAFP-21-0044, 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 ML21138A0132021-05-18018 May 2021 NRC Response to Request for Additional Time to Respond to James A. Fitzpatrick Nuclear Power Plant: Final Significance Determination of a White Finding with Assessment Follow Up and Notice of Violation-Inspection No. 05000333/2021090 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 JAFP-21-0008, Supplement to Application to Revise Surveillance Requirement (SR) 3.5.1.62021-02-18018 February 2021 Supplement to Application to Revise Surveillance Requirement (SR) 3.5.1.6 JAFP-21-0005, Response to Request for Additional Information to Support Review of License Amendment Request - Application to Modify Technical Specifications 3.6.1.3 Primary Containment Isolation Valves2021-01-26026 January 2021 Response to Request for Additional Information to Support Review of License Amendment Request - Application to Modify Technical Specifications 3.6.1.3 Primary Containment Isolation Valves ML20315A2452020-11-10010 November 2020 Response to Request for Additional Information to Support Review of License Amendment Request - Application to Modify Technical Specifications 3.6.1.3 Primary Containment Isolation Valves JAFP-20-0030, Response to Request for Additional Information in Support of License Amendment Request - Proposed Changes to the Technical Specifications to Primary Containment Hydrodynamic Loads2020-04-0606 April 2020 Response to Request for Additional Information in Support of License Amendment Request - Proposed Changes to the Technical Specifications to Primary Containment Hydrodynamic Loads JAFP-20-0026, Response to Request for Additional Information to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation2020-03-31031 March 2020 Response to Request for Additional Information to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation JAFP-20-0020, Response to Request for Additional Information to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation2020-02-21021 February 2020 Response to Request for Additional Information to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation JAFP-20-0008, Response to Request for Additional Information Associated with the License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences2020-01-16016 January 2020 Response to Request for Additional Information Associated with the License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences JAFP-19-0105, Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (Rwcu)...2019-11-0606 November 2019 Response to Request for Supplemental Information by the Office of Nuclear Reactor Regulation to Support Review of a License Amendment Request to Revise the Allowable Value for Reactor Water Cleanup (Rwcu)... ML19279A0012019-10-0606 October 2019 Response to Request for Additional Information by the Office of Nuclear Reactor Regulation to Support Review of an Emergency License Amendment Request for One Time Extension to the TS 3.8.1 Action ... 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 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 JAFP-19-0009, End of Interval Relief Request Associated with the Fourth Ten-Year Inservice Inspection (ISI) Interval2019-01-15015 January 2019 End of Interval Relief Request Associated with the Fourth Ten-Year Inservice Inspection (ISI) Interval 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 JAFP-18-0068, Response to Request for Additional Information: License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors2018-07-12012 July 2018 Response to Request for Additional Information: License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors RS-18-067, Response to Request for Additional Information Regarding Generic Letter 2016-012018-05-29029 May 2018 Response to Request for Additional Information Regarding Generic Letter 2016-01 JAFP-18-0038, Response to Request for Additional Information Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22018-04-19019 April 2018 Response to Request for Additional Information Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 JAFP-18-0029, Response to Request for Additional Information by the Office of Nuclear Reactor Regulation to Support Review of James A. FitzPatrick Nuclear Power Plant to Address Secondary Containment Personnel Access Door Openings2018-03-15015 March 2018 Response to Request for Additional Information by the Office of Nuclear Reactor Regulation to Support Review of James A. FitzPatrick Nuclear Power Plant to Address Secondary Containment Personnel Access Door Openings JAFP-18-0014, Response to Request for Additional Information on Submittal of Relief Request JAF-15R-05 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)2018-03-0202 March 2018 Response to Request for Additional Information on Submittal of Relief Request JAF-15R-05 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702) ML18025A7992018-01-25025 January 2018 Response to Request for Additional Information Regarding Generic Letter 2016-01 JAFP-18-0012, Supplemental Response Concerning License Amendment Request to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22018-01-22022 January 2018 Supplemental Response Concerning License Amendment Request to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 ML17346A1532017-12-11011 December 2017 Submittal of Response to Request for Additional Information on Relief Requests Associated with the Fifth Inservice Inspection (ISI) Interval JAFP-17-0029, Revision to Entergy'S Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of Near-Term Task Force Review of Insights...2017-04-27027 April 2017 Revision to Entergy'S Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of Near-Term Task Force Review of Insights... JAFP-17-0005, Response to 03/12/2012 Request for Information Regarding Recommendations of the Near-Term Task Force Review of Insights from the Fukushima Daiichi Accident, Recommendation 9.3 for Emergency Preparedness2017-01-23023 January 2017 Response to 03/12/2012 Request for Information Regarding Recommendations of the Near-Term Task Force Review of Insights from the Fukushima Daiichi Accident, Recommendation 9.3 for Emergency Preparedness JAFP-16-0182, Spent Fuel Pool Evaluation, 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 Fukushima Dai-ichi Accident2016-12-22022 December 2016 Spent Fuel Pool Evaluation, 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 Fukushima Dai-ichi Accident ML16335A1042016-11-29029 November 2016 Independent Spent Fuel Storage Installation, Response to Request for Additional Information Regarding Application for Order Approving Transfer F Renewed Facility Operating License and Proposed Conforming License Amendment JAFP-16-0170, Response to Request for Additional Information (RAI) Regarding License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type a and Type C Leak Rate Test Frequencies - Supplement 12016-11-21021 November 2016 Response to Request for Additional Information (RAI) Regarding License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type a and Type C Leak Rate Test Frequencies - Supplement 1 JAFP-16-0165, Response to Generic Letter 2016-01 - Monitoring of Neutron Absorbing Materials in Spent Fuel Pools2016-11-0303 November 2016 Response to Generic Letter 2016-01 - Monitoring of Neutron Absorbing Materials in Spent Fuel Pools JAFP-16-0142, Revision to Technical Specification (TS) Administrative Controls for Staffing and Training Upon Permanent Cessation of Operation (CAC No. MF7280) - Supplement 22016-09-19019 September 2016 Revision to Technical Specification (TS) Administrative Controls for Staffing and Training Upon Permanent Cessation of Operation (CAC No. MF7280) - Supplement 2 JAFP-16-0131, Response to Request for Additional Information Regarding Request for Approval of a Certified Fuel Handler Training and Retraining Program2016-08-18018 August 2016 Response to Request for Additional Information Regarding Request for Approval of a Certified Fuel Handler Training and Retraining Program JAFP-16-0130, License Amendment Request to Revise the Emergency Plan to Address the Permanently Defueled Condition - Supplement 22016-08-10010 August 2016 License Amendment Request to Revise the Emergency Plan to Address the Permanently Defueled Condition - Supplement 2 JAFP-16-0113, License Amendment Request to Revise the Emergency Plan to Address the Permanently Defueled Condition - Supplement 12016-07-0606 July 2016 License Amendment Request to Revise the Emergency Plan to Address the Permanently Defueled Condition - Supplement 1 JAFP-16-0114, Response to Request for Additional Information for Proposed Inservice Inspection Program Alternative in Accordance with 10 CFR 50.55a(z)(2), RR-202016-06-27027 June 2016 Response to Request for Additional Information for Proposed Inservice Inspection Program Alternative in Accordance with 10 CFR 50.55a(z)(2), RR-20 JAFP-16-0077, Response to Request for Additional Information (RAI) Regarding Revision to Technical Specification (TS) Administrative Controls for Staffing and Training Upon Permanent Cessation of Operation (CAC No. MF7280) - Supplement 12016-06-0303 June 2016 Response to Request for Additional Information (RAI) Regarding Revision to Technical Specification (TS) Administrative Controls for Staffing and Training Upon Permanent Cessation of Operation (CAC No. MF7280) - Supplement 1 JAFP-16-0078, Response to Request for Additional Information (RAI) Regarding Request for Approval of Certified Fuel Handler Training and Retaining Program2016-06-0303 June 2016 Response to Request for Additional Information (RAI) Regarding Request for Approval of Certified Fuel Handler Training and Retaining Program CNRO-2016-00016, Entergy - Response to RAI Questions and Submittal of Relief Request EN-ISI-15-1 (Updated)2016-06-0202 June 2016 Entergy - Response to RAI Questions and Submittal of Relief Request EN-ISI-15-1 (Updated) ML16071A4412016-03-0707 March 2016 Entergy Fleet Relief Request No. RR-EN-15-1-Proposed Alternative to Use ASME Code Case N-789-1 - E-mail from G.Davant to R.Guzman - Response to Second RAI (MF6340 - MF6349) CNRO-2016-00005, Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM)2016-02-25025 February 2016 Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM) CNRO-2016-00002, Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Secti2016-01-29029 January 2016 Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Section Xl, CNRO-2015-00002, Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 12015-12-0404 December 2015 Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 1 JAFP-15-0094, Response to Request for Additional Information for Expedited Seismic Evaluation Process Report2015-08-0404 August 2015 Response to Request for Additional Information for Expedited Seismic Evaluation Process Report 2022-06-16
[Table view] |
Text
JAFP-16-0114 June 27, 2016
U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555
Subject:
Response to Request for Additional Information for Proposed Inservice Inspection Program Alternative in Accordance with 10 CFR 50.55a(z)(2), RR-20 at James A. FitzPatrick Nuclear Power Plant James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59
References:
- 1. ENOI letter, Proposed Inservice Inspection Program Alternative in Accordance with 10 CFR 50.55a(z)(2), RR-20 at James A. FitzPatrick Nuclear Power Plant, JAFP-16-0108, dated June 21, 2016 2. NRC memo, James A. FitzPatrick Nuclear Power Plant Inservice Inspection Program Alternative Relief Request RR Draft Request for Additional Information, dated June 24, 2016
Dear Sir or Madam:
James A. FitzPatrick Nuclear Power Plant (JAF) submitted a request on June 21, 2016, for an alternative to specific portions of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," pursuant to 10 Code of Federal Regulations (CFR) 50.55a(z)(2) [Reference 1]. The NRC made a request for additional information on June 24, 2016 [Reference 2]. The attachment to this letter contains the response to this request. JAF requested approval of RR-20 [Reference 1] on or before June 28, 2016, to accommodate application of the request during the next potential outage. On June 24, 2016, due to an unrelated equipment issue, JAF has entered a shutdown. The Safety Relief Valve replacement activities have been planned for this shutdown and this will require a system leakage test and associated VT-2 visual examination of the mechanical joint connections during startup. Based on the current plant shutdown schedule, approval is requested by June 30, 2016.
Entergy Nuclear NortheastEntergy Nuclear Operations, Inc. James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-342-3840
William C. Drews Regulatory Assurance Manager - JAF JAFP-16-0114 Page 2 of 2 There are no new regulatory commitments made in this letter. Should you have any questions, please contact the Regulatory Assurance Manager, Mr. William C. Drews, at (315) 349-6562.
Very truly yours, c_f)_ William C. Drews Regulatory Assurance Manager WCD:dc
Attachment:
Response to Request for Additional Information for RR-20 cc: USNRC, Regional Administrator, Region I USNRC, Project Directorate USNRC, Resident Inspector
JAFP-16-0114 Attachment Response to Request for Addi tional Information for RR-20 (3 Pages)
JAFP-16-0114 Attachment Response to Request for Additional Information for RR-20 Page 1 of 3 RAI-1 Page 2 of the relief request states that "-JAF may perform the required VT-2 leakage examination for any repair/replacement activities of mechanical joint connections performed in a future shutdown at a reactor pressure of 905 psig (consistent with 87% of the pressure required by IWB-5221(a) from ASME Code Case N-795)- Please explain the statement "JAF may perform-" rather than "JAF will perform-
Response The "JAF may perform-" statement on Page 2 of the Attachment for Relief Request RR-20 was referring to the time of a future shutdown. Upon approval, the relief request was meant to be an option during outage planning. During the emergent plant shutdown on June 24, 2016, James A.
FitzPatrick (JAF) will perform replacement of 3-stage Safety Relief Valves (SRVs) with 2-stage SRVs. JAF will implement the Relief Request RR-20, if approved, for the current shutdown. NRC approval of the Relief Request is sought prior to startup of the current outage in order to allow the performance of the VT-2 examinations on Class 1 SRV mechanical joint components at a pressure less than the Code-required pressure. RAI-2 Page 3 of the relief request states that "-The proposed alternative provides an acceptable level of quality and safety and is consistent with requirements of ASME Code Case N-795 for a BWR Class 1 system leakage test, following repair/replacement activities. JAF proposes to perform the system leakage test at a pressure at least 5% (at least 55 psi) above the minimum pressure required by Code Case N-795." On page 2, the licensee stated that it may use a pressure of greater than 905 psig to perform the system leakage test. The operating pressure at 100% rated power is 1040 psig. Therefore, 905 psig is 87% of 100% rated power (905/1040 = 87%). The above statement appears to indicate that JAF will perform the pressure test at 960 psig (905 psig + 55 psi). This discussion could be confusing when one area of the relief request proposes 905 psig and another area of the relief request proposes 960 psig. Please confirm that the system leakage test will be performed at a pressure of at least 960 psig and clarify as necessary.
Response On the bottom of Page 2 at the start of the 'Proposed Alternative and Basis for Use' section, JAF was stating the minimum pressure of at least 905 psig to be consistent with Code Case N-795. The maximum reactor pressure is associated with procedural limitations on Drywell entry due to plant conditions (15% power), at approximately 970 psig. The 960 psig test pressure is a pressure that is greater than 5% above the minimum test pressure of greater than or equal to 905 psig (905 psig being the minimum pressure required by Code Case N-795). The Page 3 reference to at least 55 psig above the minimum pressure required by the Code case is stating that JAF intends to perform the test at a higher pressure than minimum test pressure of 905 psig (namely at least 960 psig for added margin).
If RR-20 is approved, JAF would perform the system leakage test at a test pressure of at least 960 psig. RAI-JAF-3 Page 2 of the relief request states in part that "Disposition of any observed leakage will consider the marginal increase in leakage rates that might occur at the nominal operating pressure associated with 100% rated reactor power (i.e. 1040 psig) and the actual reactor pressure when the examination was performed-" Please provide further explanation of this statement.
JAFP-16-0114 Attachment Response to Request for Additional Information for RR-20 Page 2 of 3 Response The statement attempted to recognize that Class 1 pressure boundary leakage rates detected at a lower approved Relief Request RR-20 test pressure would likely be greater if detected at the higher nominal operating pressure associated with 100% rated power (1040 psig). The recognition of higher leakage rates at higher pressures would be taken into consideration during evaluations for corrective action should leakage be detected. JAF will follow IWA-5250(a) and any deficiencies will be evaluated in the Corrective Action Program. In the case of identified leakage, a ratio comparing the lower test pressure to the higher operating pressure would be used during engineering evaluations of identified leakage. RAI-JAF-4 If there are unplanned shutdowns with drywell entries before permanent cessation of operation, what additional action would the licensee take to inspect the affected mechanical joint connections to look for leakage?
Response During startup from the current outage, after repl acement of the SRVs, VT-2 examinations will be used to detect leakage from the affected mechanical joint connections. During plant operation, drywell monitoring systems will be utilized to identify any leakage. If there are additional unplanned shutdowns with drywell entries before permanent cessation of operation, JAF will perform an initial walk dow n upon entry to identify any evidence of leakage from SRV-related repair and replacement activities. RAI-JAF-5 The NRC understands the licensee's desire to test at lower pressure. However, It is not clear to the NRC staff from the licensee's request how the needed pressure will be obtained. Based on other, similar requests, the NRC believes that the licensee may intend to use nuclear heat to obtain the necessary pressure. a. Does the licensee propose to use nuclear heat to obtain the necessary pressure for testing? b. If so, please discuss the risk to the plant associated with the proposed test method as compared with other available options. This discussion should include risks associated with the use of nuclear heat, as well as changes in plant alignment (including safety systems which must be removed from service) associated with the proposed and alternate methods of obtaining the necessary pressure. This discussion may be presented in tabular form if desired.
Response a) Yes, if Relief Request RR-20 is approved, nuclear heat will be used to obtain the lower test pressure during the startup sequence. b) During a typical post refuel outage system leakage test, the reactor vessel and main steam lines are essentially water solid and safety systems are isolated or in standby. The alternate method of using nuclear steam (heat) to obtain the required test pressure will use the Electro-Hydraulic Control (EHC) System to control pressure. This is a normal pressure control mode for a Boiling Water Reactor (BWR) with Reactor Pressure Vessel (RPV) temperature and level in more normal operating bands. Safety systems will be operable during this leakage test (ECCS, PCIS and RPS) as identified in the table below and there is no change to leakage indication available to monitor unidentified drywell leakage.
JAFP-16-0114 Attachment Response to Request for Additional Information for RR-20 Page 3 of 3 Performance of a pressure test using nuclear steam is a safer method to pressure test a mechanical joint in a BWR than to perform a cold hydrostatic test for the following reasons: a) A cold hydrostatic pressure test is performed with the RPV and steam lines flooded, which results in a nearly water solid condition. Pressure is controlled by carefully balancing RPV water injection with RPV water rejection/letdown. Small changes in this balance can create a rapid rate of pressure change. A cold hydrostatic test is performed at low temperature / high pressure. This condition places the plant significantly closer to the RPV pressure-temperature limits curve. b) The nuclear steam pressure test is performed at high temperature/high pressure. This places the plant further from the RPV pressure-temperature (P-T) limits curve. c) Fuel decay heat at the time of the proposed pressure test is much higher in this case than after a normal refueling outage cold hydrostatic test. More fuel decay heat impacts RPV pressure control. Risk to the plant is minimized by performance of a leakage test using Relief Request RR-20 opposed to a cold hydrostatic leakage test because of significantly increased margin to RPV P-T limits curve with only a minimal reduction in test pressure; significantly more controllability of RPV pressure using the EHC system, and safety systems are available as opposed to steam-driven systems not being available during a cold hydrostatic test. System Leakage Test during startup with RR-20 "Hydro" Test (ST-39H) High Pressure Coolant Injection Operable Isolated Core Spray Operable Operable Low Pressure Coolant Injection Operable Operable Extensive valve manipulations No (only per plant start-up) Yes per ST-39H Reactor Vessel and Main Steam lines are virtually water solid No Yes RCIC Operable Isolated Extensive water moves No Yes Flood Risk Steam will be present at a lower than operating pressure Water at a higher pressure present therefore a greater
flood risk Energy input to the drywell if leakage is detected after
replacement More energy input due to higher steam temperature Less energy input due to lower temperature water