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MONTHYEARML0607300032006-03-29029 March 2006 General Electric Company, Request for Withholding Information from Public Disclosure for James A. FitzPatrick Nuclear Power Plant Project stage: Withholding Request Acceptance ML0630501222006-11-0707 November 2006 Request for Additional Information Regarding Amendment Application for Arts/Meod Modifications Project stage: RAI JAFP-06-0180, Core Operating Limits Report, Revision 222006-12-27027 December 2006 Core Operating Limits Report, Revision 22 Project stage: Request ML0703902602007-02-26026 February 2007 Request for Withholding Information from Public Disclosure Regarding Amendment Application for Arts/Meod Modifications Project stage: Withholding Request Acceptance ML0705904352007-03-0505 March 2007 Draft Safety Evaluation for Implementation of the Average Power Rate Monitor, Rod Block Monitor Technical Specification Improvements with the Maximum Extended Operating Domain Analysis Project stage: Draft Approval ML0714303022007-05-17017 May 2007 Technical Specifications, Allow Additional Startup and Operating Flexibility and an Expanded Operating Domain Project stage: Other ML0704300652007-05-17017 May 2007 License Amendment, Issuance of Amendment Implementation of Average Power Rate Monitor, Rod Block Monitor Technical Specification Improvements with Maximum Extended Operating Domain Analysis Project stage: Approval JAFP-07-0090, Core Operating Limits Report, Revision 232007-07-23023 July 2007 Core Operating Limits Report, Revision 23 Project stage: Other 2007-02-26
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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:Request for Additional Information (RAI)
MONTHYEARML24024A1372024-01-24024 January 2024 NRR E-mail Capture - Final Snsb RAI Regarding FitzPatrick Amendment to Modify Safety Relief Valves Setpoint Lower Tolerance ML24018A0012024-01-18018 January 2024 Notification of Commercial Grade Dedication Inspection (05000333/2024010) and Request for Information ML23264A7992023-09-21021 September 2023 NRR E-mail Capture - Final RAI - Constellation Energy Generation, LLC Fleet Request License Amendment Request to Adopt TSTF-580, Revision 1 ML23164A0322023-06-13013 June 2023 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000333/2023010 ML22321A0102022-11-17017 November 2022 Notification of Conduct of a Fire Protection Team Inspection ML22124A2672022-05-0404 May 2022 Request for Additional Information for James A. FitzPatrick Nuclear Power Plant TSTF-505 ML22041B5362022-02-10010 February 2022 NRR E-mail Capture - Constellation Energy Generation, LLC - Request for Additional Information Regarding Fleet License Amendment Request to Adopt TSTF-541 ML22020A0642022-01-13013 January 2022 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles ML21256A1902021-09-10010 September 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21187A0522021-07-0606 July 2021 Fitz RAI Regarding FitzPatrick Amendment Request to Modify SR 3.5.1.6 ML21144A2132021-05-24024 May 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21062A0652021-03-0101 March 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative to Documentation Requirements for Pressure Retaining Bolting ML21049A2572021-02-18018 February 2021 Request for Additional Information Byron/Dresden Proposed Changes to Site Emergency Plans to Support Post-Shutdown and Permanently Defueled Conditions (EPID-2020-LLA-0240 & EPID-2020-LLA-0237) ML21041A1932021-02-10010 February 2021 Request for Information for a Triennial Baseline Design-Basis Capability of Power-Operated Valves Under 10 CFR 50.55a Requirements Inspection; Inspection Report (05000333/2021011) ML20066L3682020-03-0505 March 2020 Request for Additional Information: License Amendment Request for Change to the Technical Specifications to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment ML20056E7992020-02-25025 February 2020 Request for Additional Information for LAR on Primary Containment Hydrodynamic Loads ML20035D5762020-02-0303 February 2020 Request for Additional Information: License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequence ML20027A0112020-01-23023 January 2020 Request for Additional Information: License Amendment Request for Change to the Technical Specifications to Revise the Allowable Value for Reactor Water Cleanup (RWCU) System Primary Containment Isolation ML19353A9452019-12-19019 December 2019 Request for Additional Information: License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences ML19322A0062019-11-15015 November 2019 NRR E-mail Capture - Fermi 2: Request for Additional Information- Relief Request VRR-006, Proposed Alternative for Preservice Testing of Butterfly Valves ML19280A0372019-10-0505 October 2019 NRR E-mail Capture - FitzPatrick Revised Request for Additional Information: Emergency Amendment to Extend Completion Time of Transformer to 21 Days ML19280A0392019-10-0505 October 2019 NRR E-mail Capture - FitzPatrick Additional Request for Additional Information: Emergency Amendment to Extend Completion Time of Transformer to 21 Days ML19280A0362019-10-0404 October 2019 NRR E-mail Capture - FitzPatrick Request for Additional Information: Emergency Amendment to Extend Completion Time of Transformer to 21 Days ML19099A2672019-04-16016 April 2019 Use of Encryption Software for Electronic Transmission of Safeguards Information ML19099A2852019-03-12012 March 2019 Draft RAI: FitzPatrick Request to Update Electronic Transmission of Safeguards Information ML19025A1202019-01-24024 January 2019 NRR E-mail Capture - Calvert Cliffs, Fitzpatrick, and Nine Mile Point - Request for Additional Information Regarding License Amendment Request to Revise Emergency Response Organization Staffing ML18353B5112018-12-20020 December 2018 Request for Additional Information Relief Request No. 14R-22, for Fourth 10-Year Inservice Inspection Interval ML18164A3652018-06-13013 June 2018 NRR E-mail Capture - FitzPatrick RAIs - LAR to Adopt EAL Schemes Pursuant to NEI 99-01, Revision 6 ML18094B0922018-04-0505 April 2018 Enclosurequest for Additional Information (Letter to B. S. Ford RAI Regarding Entergy Operations, Inc.'S Decommissioning Funding Plan Update for ISFSI Docket Nos.: 72-43, 72-51, 72-1044, 72-07, 72-12, and 72-59) ML18085A6922018-03-26026 March 2018 NRR E-mail Capture - James A. FitzPatrick Nuclear Power Plant, Unit 1 - Request for Information to Adopt Traveler TSTF-542, RPV Water Inventory Control ML18085A6912018-03-23023 March 2018 NRR E-mail Capture - Draft Request for Information (RAI) for JAFNPP Traveler TSTF-542, RPV Water Inventory Control ML18029A8422018-01-29029 January 2018 NRR E-mail Capture - Request for Additional Information: FitzPatrick License Amendment Request to Revise Technical Specifications to Address Secondary Containment Personnel Access Door Openings ML17335A1002017-11-29029 November 2017 NRR E-mail Capture - FitzPatrick - Request for Additional Information - Relief Request 15R-02 Regarding the Use of BWRVIP Guidelines Instead of ASME Code (CAC: MG0116; EPID: L-2017-LLR-0083) ML17285B1962017-10-27027 October 2017 Request for Additional Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools. ML17228A0052017-08-15015 August 2017 NRR E-mail Capture - Request for Additional Information - St. Lucie Relief Request #3 - Icw 30 Pipe Defect Removal (MF9288) ML16299A0192016-11-0202 November 2016 Request for Additional Information Regarding Direct License Transfer from Entergy to Exelon ML16287A6502016-10-21021 October 2016 Request for Additional Information Regarding: License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type a and Type C Leak Rate Test Frequencies ML16280A5732016-10-18018 October 2016 Request for Additional Information Relief Request for Proposed Alternative for the Implementation of BWRVIP-05 ML16117A1862016-04-26026 April 2016 Entergy Fleet Relief Request EN-ISI-15-1 Request for Additional Information - 04/26/16 Email from R. Guzman to G. Davant (CAC Nos. MF7133-MF7136) ML16013A0642016-01-13013 January 2016 NRR E-mail Capture - Request for Additional Information Entergy CNRO-2015-00023 - Revision to Entergy Quality Assurance Program Manual (Fleet Submittal CAC Nos. MF7086-MF7097) ML15341A1662015-12-0707 December 2015 NRR E-mail Capture - Entergy Fleet RR-EN-15-1, Request for Additional Information (CACs MF6341-MF6349) ML14240A6122014-09-24024 September 2014 Request for Additional Information Regarding Proposed Changes to the Technical Specification Low Pressure Safety Limit ML14184A6152014-07-25025 July 2014 Request for Additional Information Regarding Proposed Safety Limit Minimum Critical Power Ratio License Amendment ML14195A0972014-07-16016 July 2014 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1, Seismic Hazard and Screening Report ML14164A5382014-07-0202 July 2014 Request for Additional Information Regarding 10 CFR 50.55A Alternate Request PRR-05 (Tac No. MF3680) ML14093A6772014-05-0101 May 2014 SONGS - Request for Additional Information Concerning Pre-Emption Authority ML13338A6452013-12-12012 December 2013 Interim Staff Evaluation and Request for Additional Information, Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation ML13304B4182013-11-0101 November 2013 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns ML13226A5342013-08-29029 August 2013 Request for Additional Information Regarding Overall Integrated Plan for Reliable Spent Fuel Pool Instrumentation (Order EA-12-051) ML13179A4622013-07-0505 July 2013 Entergy Nuclear Operations, Inc., Decommissioning Funding Status Report Request for Additional Information Regarding the Decommissioning Funding Status Report 2024-01-24
[Table view] |
Text
November 7, 2006Mr. Michael R. KanslerPresident Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - REQUEST FORADDITIONAL INFORMATION REGARDING AMENDMENT APPLICATION FOR ARTS/MEOD MODIFICATIONS (TAC NO. MC9681)
Dear Mr. Kansler:
On January 26, 2006, Entergy Nuclear Operations, Inc. (Entergy), submitted an application fora proposed amendment for the James A. FitzPatrick Nuclear Power Plant which would modify Technical Specification (TS) requirements to support the implementation of Average Power Range Monitor, Rod Block Monitor, TSs/Maximum Extended Operating Domain (ARTS/MEOD) analyses.The Nuclear Regulatory Commission staff is reviewing the submittal and has determined thatadditional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). During a telephone call on October 25, 2006, the Entergy staff indicated that a response to the RAI would be provided within 45 days. Please contact me at (301) 415-2901 if you have any questions on this issue.Sincerely,/RA/John P. Boska, Senior Project ManagerPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-333
Enclosure:
RAIcc w/encl: See next page
ML063050122OFFICELPL1-1/PMLPL1-1/LASBWB/BCLPL1-1/BCNAMEJBoskaSLittleGCranstonRLauferDATE11/01/0611/06/0611/06/0611/07/06 FitzPatrick Nuclear Power Plant cc:
Mr. Gary J. TaylorChief Executive Officer Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213Mr. John T. HerronSr. VP and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Mr. Peter T. DietrichSite Vice President Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power PlantP.O. Box 110 Lycoming, NY 13093Mr. Kevin J. MulliganGeneral Manager, Plant Operations Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power PlantP.O. Box 110 Lycoming, NY 13093Mr. Oscar LimpiasVice President Engineering Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Mr. Christopher SchwarzVice President, Operations Support Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Mr. John F. McCannDirector, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Ms. Charlene D. FaisonManager, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Mr. Michael J. ColombDirector of Oversight Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Mr. David WallaceDirector, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093Mr. James CostedioManager, Regulatory Compliance Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093Assistant General CounselEntergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601Regional Administrator, Region IU.S. Nuclear Regulatory Commission
475 Allendale Road King of Prussia, PA 19406Resident Inspector's OfficeJames A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 FitzPatrick Nuclear Power Plant cc:
Mr. Charles Donaldson, EsquireAssistant Attorney General New York Department of Law 120 Broadway New York, NY 10271Mr. Peter R. Smith, PresidentNew York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399Mr. Paul EddyNew York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350Oswego County Administrator Mr. Steven Lyman 46 East Bridge Street Oswego, NY 13126SupervisorTown of Scriba Route 8, Box 382 Oswego, NY 13126Mr. James H. SniezekBWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490Mr. Michael D. LysterBWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306Mr. Garrett D. Edwards814 Waverly Road Kennett Square, PA 19348 EnclosureREQUEST FOR ADDITIONAL INFORMATIONREGARDING AMENDMENT APPLICATION FOR ARTS/MEOD MODIFICATIONS ENTERGY NUCLEAR OPERATIONS, INC.JAMES A. FITZPATRICK NUCLEAR POWER PLANTDOCKET NO. 50-333On January 26, 2006, Entergy Nuclear Operations, Inc. (Entergy), submitted an application fora proposed amendment for the James A. FitzPatrick Nuclear Power Plant which would modify Technical Specification (TS) requirements to support the implementation of Average Power Range Monitor (APRM), Rod Block Monitor (RBM), TSs/Maximum Extended Operating Domain (ARTS/MEOD) analyses. The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has the following questions:Section 1.0 Introduction1-1Allowable value (AV) and analytical limit (AL) are used in a few places in attachment 5 tothe application. For example, on page 1-5, the first paragraph states "In the low flow stability region, the scram AVs are based on the scram ALs given in terms of core flow using the JAF
[James A. FitzPatrick] core flow to drive flow relationship...". Provide a clear definition of AV and AL as used in the above statement. What is the difference between AV and AL and how they are related?1-2 The two tables given on page 1-5 provide the scram and rod block AVs for single-loopoperation. Provide the basis for the flow-biased scram and rod block setpoints. How were the specific slopes derived or established?1-3Pages 1-4 and 1-5 show the two-loop operation (TLO) and single-loop operation (SLO) AVs for ranges of drive flow. Explain if the plant can operate at SLO condition at different pump speeds? If not, explain the reasons for the range of flow-biased SLO rod block and scram lines. Are these flow-biased scram lines (which differ from the TLO scram lines) intended to initiate a scram if a transient occurs while the plant is operating at SLO condition? If yes, state what analyses are supporting operation at lower pump speeds for SLO (single pump at maximum capacity).1-4 During SLO what is the corresponding percent power and percent flow for maximumextended load line limit analysis (MELLLA) operation?Section 3 Fuel Thermal Limits3-1The last paragraph on Page 1-5 discussed RBM setpoint relaxation and that the RBM isnot credited in the rod withdrawal error (RWE) for the reload analysis. Did Entergy perform RWE analyses to determine that a thermal limits penalty will not be necessary? 3-2In Table 3-5, page 3-14, notes b, c, and d in the table were misplaced. Please correctthis.3-3On Page 3-1, the first paragraph has the statement "The minimum core flow at 100% ofrated thermal power (RTP) used in the analysis presented in this section is 81% of RCF [rated core flow ]." Why not use the exact point at 79.8% of RCF corresponding to Figure 1-1 in the analysis?3-4At the top of Page 3.2, it states "The other two events (IRLS [idle recirculation loop start-up] and FRFI [fast recirculation flow increase]) are by design most limiting at off-rated conditions. Even when originated from their most limiting off-rated condition, the IRLS and FRFI are less limiting than the fast pressurization events (TTNBP [turbine trip with no bypass],
LRNBP [load rejection with no bypass], or FWCF [feedwater controller failure]) at rated power conditions. Thus, the IRLS and FRFI events were not considered in the determination of the off-rated limits." (a)Provide additional information (for example, plant response) to support the abovestatements. Explain why those two events were not analyzed in MELLLA operation domain.(b)In addition, the table on Page 3-3 (lower left corner) states "The LFWH [loss offeedwater heating], FLE [fuel loading error], IRLS, and FRFI events are not limiting at off-rated conditions." This statement is not consistent with the above statement. Please provide clarification.3-5In Table 3-2, the table and footnote showed peak transient response values whichoccurred at end of cycle (EOC). Confirm if this analysis was performed at other exposures such as beginning of cycle (BOC) and middle of cycle (MOC). 3-6Section 3.2 has a table for analytical assumptions. Please document that the JAF TSminimum number of SRVs and Turbines Out of Service is consistent with the analyses assumptions.3-7Section 3.3.6 stated that "Only adjustment of the P < PBypass portion of the MCPR(P)[minimum critical power ratio, power dependent] curve is required because, at P PBypass, theK(P) applies the rated power OLMCPR [operating limit MCPR] adjustment to the MCPR(P)." Please reference the appropriate NRC-approved amendment to the General Electric StandardApplication for Reactor Fuel or other topical report that explains this off-rated calculation methodology.This section also provided an equation for the adjustment as follows when operating in SLO:
SLO OLMCPR = OLMCPR dual-loop + SLMCPR [safety limit MCPR] SLO - SLMCPR dual-loop.Provide additional information on this approach (e.g., how was the above equation obtained?)and references. Section 4 Reactor Recirculation (RR) System4-1Section 4.0 on the RR System has the following statements:"The effects of aging and degradation mechanisms (e.g., jet pump crudding) were not includedin the evaluation.""The results of the evaluation indicate that the capability of the [recirculation] system to supportoperation at 105% of RCF may be marginal during some of the fuel cycle. If so, full 105% core flow may not be available until the end of the fuel cycle when the core differential pressure decreases, which causes the jet pump flow to increase for a given [recirculation] pump flow.
Rotating equipment limitations are economic in nature and do not affect plant safety."a) Is the recirculation system operating at its nameplate rating in capacity? Explain how plantmeasured flow data compared to the previous reload analysis in terms of the measured core flow versus assumed AV. b) What is the potential impact of not accounting for "jet pump crudding," in the safetyanalyses? Explain why the analyses results are valid if the impact of aging and degradation of the system (such as jet pump crudding) are not accounted for. c) Discuss what impact jet pump crudding may have on SLO flow calculations.
Section 7 Instability7-1Figure 7-1 shows the scram AV and rod block AV lines, which do not appear to bestraight linear scram lines as expected from the equations. Specifically, the scram and rod block lines within the stability exclusion appear to be curved lines. Explain why.
7-2Designate the SLO operating state point in the power/flow map and identify the corresponding scram and rod block lines. Show the flow biased SLO scram and rod block setpoints on Figure 7-1.7-3Describe how you obtain core mass flow rates, Wc, for SLO. Specifically, consideringthe reverse flow in the inoperable loop, explain how the accuracy of the Wc values is determined. 7-4On page 7-2, it states "For JAF Cycle 16, the core average power-to-flow ratio isestimated to be 56.8 MWt/Mlbm/hr [megawatts thermal per million pounds mass per hour] and the generic DIVOM [delta critical power ratio over initial MCPR versus the oscillation magnitude]
slope is valid for Cycle 16 operation." Explain what core flow state point was used in determining the 56.8 Mwt/Mlbm/hr value? Was this value calculated based on the rated thermal power at 75% core flow? If not, state why the minimum core flow state point is not an appropriate value. 7-5In the middle of page 7-2, a new APRM flow-biased flux scram line for ARTS/MEODoperation was determined with the additional conservatism in the evaluation. The additional conservatism was listed in numerical order 1, 2, and 3. Please explain why these assumptions were considered as conservative and how those numbers used in the assumptions wereobtained.Section 8 Loss-of-Coolant Accident (LOCA)8-1Please provide additional discussion on what kind of axial power profiles were assumedin the LOCA analysis.8-2 At the end of the fourth paragraph on page 8-1, it states "These results show thatoperation in the MELLLA region affects the nominal PCT [peak cladding temperature] by +3F and the Appendix K PCT by +93F." Please explain why the effects on the Appendix K PCT aremuch more severe than the nominal PCT.8-3 What were your upper bound PCT results?
8-4 On Page 8-4, there are the following notes for Table 8-2: "(a) The effect on the ECCS[emergency core cooling systems]-LOCA analysis PCT of operation in the MEOD domain for GE14 is conservatively applicable to GE12 and (b) The effect on the ECCS-LOCA analysis PCT for operation at core flows greater than 100% (ICF [increased core flow]) is negligible. Thus the PCTs for the limiting large break cases at rated conditions are applicable to the ICF condition."
Please provide justification for these notes. 8-5Please provide all state points including SLO and ICF for the calculation in Table 8-2.8-6 On page 8-2, there is a statement "The current JAF Licensing Basis PCT for GE12 fuelis 1370F with a 170F adder for 10 CFR 50.46 reported errors applicable to the JAFECCS-LOCA analysis". Was the adder limited to 10 CFR 50.46 GE12 fuel or are additional adders applicable to theGE14 fuel? In addition, a "170F" PCT adder is a significant number. Justify why JAF did notperform LOCA reanalysis for the GE12 fuel.8-7 Did JAF perform full spectrum ECCS-LOCA analysis? If not, please justify.
Section 11 Anticipated Transient Without Scram (ATWS)GE topical Table 11-2 shows the Peak Vessel Bottom Pressure for ATWS analysis at 1493psig, which provides very little margin to the ATWS overpressure protection criterion of 1500 psig. The following questions pertain to the key assumptions, conservatism and valve tolerances assumed in the ATWS analysis.11-1Table 11-1 shows that the analysis assumed two of the safety-relief valves (SRVs) withthe lowest pressure setpoints were assumed to be out of service (OOS). This is conservative.
However, explain why two SRVs OOS was assumed in the analysis? Are there specific known reasons (e.g., tolerances outside TS values) that may lead to declaring SRVs OOS frequently? 11-2To demonstrate the SRV performance at FitzPatrick, provide the "as found" SRVtolerances data. If the SRV tolerances are outside the TS value, justify why the tolerances assumed in the safety analyses should not be increased. 11-3The application stated: "The MEOD analysis assumed that the SRVs opened at theupper Analytical Limit of the SRV Electric Lift Subsystem [(SRVELS)], and that the two lowest set SRVs were OOS-" The following questions address crediting the SRVELS in the safety analyses.11-3.1Table 11-1 specifies the initial conditions assumed in the ATWS analyses and showsthat the SRVELS was credited. However, the Updated Final Safety Analysis Report (UFSAR) 4.4.5 (4th paragraph, last sentence) states that "SRVELS is not credited in any accident analysis." Since the SRV electric lift system is not a TS specified safety-grade system, justify why it is acceptable to take credit for it. Most importantly, state why credit for the SRVELS is necessary at FitzPatrick?11-3.2Table 11-1 shows SRVELS opening analytical limits at the UFSAR setpoint values of+1.5%. Justify the basis for not using TS +3% tolerance value. 11-4Document the reasons why main steam isolation valve (MSIV) and pressure regulatorfailed open (PRFO) are the most limiting ATWS transients.11-5Table 11-2 shows a peak vessel bottom pressure for ATWS analysis of 1493 psig at31.4 seconds. Since the ATWS peak pressure margin is low, state what conservatisms were assumed in the plant-specific inputs and ATWS analysis methods that will provide some confidence that the small margin is acceptable.11-6 Confirm standby liquid control system (SLCS) success by providing the following eventsequence information, preferably in tabular format:a) Time of SLCS initiationb) Reactor pressure vessel (RPV) bottom head pressure at time of SLCS initiation c) RPV bottom head pressure at time of SLCS injection (SLCS initiation plus 30 sec SLCS liquid transport time - from Table 11-1) d) SLC pump discharge relief valve setpoint e) Delta psig SLCS pump margin available at time of SLCS initiation f) Time hot shutdown is achieved.Instrumentation Questions1.Attachment 5 of the January 26, 2006, submittal is the General Electric TechnicalReport NEDC-33087P, Revision 1, dated September 2005. This report, on pages 1-4 and 1-5, lists the AV for the Flow-Biased APRM neutron flux high trip setting as:Two Loop Operation:0.38
- Wd + 61.0% for0% < Wd 24.7%1.15
- Wd + 42.0% for24.7% < Wd 47.0%0.63
- Wd + 73.7% for47.0% < Wd 68.7%With a maximum of 117.0% power for Wd > 68.7% Single Loop Operation:0.38
- Wd + 57.9% for0% < Wd 32.7%1.15
- Wd + 32.8% for32.7% < Wd 50.1%0.58
- Wd + 61.3% for50.1% < Wd 95.9%With a maximum of 117.0% power for Wd > 95.9%By letter dated December 22, 2005, Entergy submitted Revisions 19 and 20 (cycle 17 update)to the Core Operating Limits Report which also listed the same equation. However, these equations are listed as trip settings. Based on this, the NRC staff is unable to determine whether these values have been calculated as AVs or trip settings. As a result, it is not clear as to the setpoint methodology used or how the instrument uncertainties have been addressed.
Therefore, in order for the staff to determine the adequacy of the setpoint determination, please provide the setpoint methodology and calculation performed to determine the trip setpoint and limiting safety system setting. Also provide the information on how the operability of these instruments are determined during the surveillance tests required by the TSs.2.On the bottom of page 5 of Attachment 1 to the December 22, 2005, application, itstates that the physical changes to the plant to accommodate the expanded operating region include Flow Control Trip Reference cards. GE Report NEDC- 33087 references GE licensing topical report NEDC -32339P-A, Supplement 2, Revision 1. The staff's acceptance of NEDC -
32339P - A is based on certain design and installation criteria. Please provide the information to show that FitzPatrick meets those criteria. If these criteria are not met then justify the acceptability of these cards to establish the expanded operating region.