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Category:Letter type:RS
MONTHYEARRS-24-070, Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions2024-07-12012 July 2024 Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations RS-24-053, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed2024-06-0606 June 2024 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed RS-24-055, 2023 Corporate Regulatory Commitment Change Summary Report2024-05-17017 May 2024 2023 Corporate Regulatory Commitment Change Summary Report RS-24-042, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion2024-05-10010 May 2024 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion RS-24-046, 10 CFR 50.46 Annual Report2024-05-0606 May 2024 10 CFR 50.46 Annual Report RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024 Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests RS-24-001, Response to Request for Additional Information Regarding Relief Request I5R-26, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval2024-01-0303 January 2024 Response to Request for Additional Information Regarding Relief Request I5R-26, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval RS-23-128, Response to Request for Additional Information for the Emergency License Amendment Request Increase Technical Specifications Completion Time in TS 3.8.1.B.4 from 7 Days to 30 Days2023-12-15015 December 2023 Response to Request for Additional Information for the Emergency License Amendment Request Increase Technical Specifications Completion Time in TS 3.8.1.B.4 from 7 Days to 30 Days RS-23-123, Emergency License Amendment Request - Increase Technical Specifications Completion Time in TS 3.8.1.B.4 from 7 Days to 30 Days2023-12-13013 December 2023 Emergency License Amendment Request - Increase Technical Specifications Completion Time in TS 3.8.1.B.4 from 7 Days to 30 Days RS-23-104, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-17017 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RS-23-113, Submittal of Updated Final Safety Analysis Report (Ufsar), Revision 17 and Fire Protection Report (Fpr), Revision 262023-10-20020 October 2023 Submittal of Updated Final Safety Analysis Report (Ufsar), Revision 17 and Fire Protection Report (Fpr), Revision 26 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 RS-23-089, Sixth Ten-Year Interval Inservice Testing Program2023-09-0505 September 2023 Sixth Ten-Year Interval Inservice Testing Program RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs RS-23-086, Relief Request I5R-26, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval2023-08-28028 August 2023 Relief Request I5R-26, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval 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 RS-23-059, 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 4b2023-06-0808 June 2023 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-23-060, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors2023-06-0808 June 2023 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors 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-22-067, 10 CFR 50.46 Annual Report2023-05-0404 May 2023 10 CFR 50.46 Annual Report RS-23-068, Response to Request for Additional Information for Quad Cities Relief Request I6R-11, Proposed Alternatives for a Temper Bead Weld Repair of the Mating Surfaces of the Reactor Pressure Vessel Head and Shell2023-04-28028 April 2023 Response to Request for Additional Information for Quad Cities Relief Request I6R-11, Proposed Alternatives for a Temper Bead Weld Repair of the Mating Surfaces of the Reactor Pressure Vessel Head and Shell RS-23-058, Submittal of Sixth Inservice Inspection Interval Relief Request I6R-11 Proposed Alternatives for a Temper Bead Weld Repair of the Mating Surfaces of the Reactor Pressure Vessel Head and Shell Flanges2023-04-24024 April 2023 Submittal of Sixth Inservice Inspection Interval Relief Request I6R-11 Proposed Alternatives for a Temper Bead Weld Repair of the Mating Surfaces of the Reactor Pressure Vessel Head and Shell Flanges 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 RS-23-007, Application to Adopt TSTF-564, Safety Limit MCPR2023-03-0303 March 2023 Application to Adopt TSTF-564, Safety Limit MCPR RS-23-045, Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.2032023-02-28028 February 2023 Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.203 RS-23-032, Application to Move SR 3.5.1.2 Note to LCO 3.5.1 in Accordance with TSTF-416, LPCI Valve Alignment Verification Note Location2023-02-0303 February 2023 Application to Move SR 3.5.1.2 Note to LCO 3.5.1 in Accordance with TSTF-416, LPCI Valve Alignment Verification Note Location RS-23-034, Notification of Extension to the Fifth Ten-Year Interval of the Inservice Testing Program2023-02-0202 February 2023 Notification of Extension to the Fifth Ten-Year Interval of the Inservice Testing Program RS-23-003, Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-102023-01-31031 January 2023 Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10 RS-23-033, Request for Exemption from 10 CFR 2.109(b)2023-01-27027 January 2023 Request for Exemption from 10 CFR 2.109(b) RS-23-005, Response to Request for Additional Information for Quad Cities Relief Request RV-04, Inservice Testing of High Pressure Coolant Injection Drain Pot Solenoid Valves2023-01-17017 January 2023 Response to Request for Additional Information for Quad Cities Relief Request RV-04, Inservice Testing of High Pressure Coolant Injection Drain Pot Solenoid Valves RS-22-127, Submittal of Sixth Inservice Inspection Interval Relief Request I6R-10 Reactor Pressure Vessel Penetration N-11B Repair2022-12-14014 December 2022 Submittal of Sixth Inservice Inspection Interval Relief Request I6R-10 Reactor Pressure Vessel Penetration N-11B Repair RS-22-126, Constellation Energy Generation, LLC - Request to Use Provisions of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2022-11-30030 November 2022 Constellation Energy Generation, LLC - Request to Use Provisions of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI RS-22-119, Withdrawal of Relief Request RV-11 Associated with the Sixth Inservice Testing Interval2022-10-31031 October 2022 Withdrawal of Relief Request RV-11 Associated with the Sixth Inservice Testing Interval RS-22-109, Response to Request for Additional Information License Amendments Related to Fuel Storage2022-10-12012 October 2022 Response to Request for Additional Information License Amendments Related to Fuel Storage RS-22-112, Submittal of RV-04 Relief Request Associated with the Sixth Inservice Testing Interval2022-10-0707 October 2022 Submittal of RV-04 Relief Request Associated with the Sixth Inservice Testing Interval RS-22-108, Response to Request for Additional Information LaSalle County Station, Units 1 and 2 and Quad Cities Nuclear Power Station, Units 1 and 2 License Amendments Related to Fuel Storage2022-10-0505 October 2022 Response to Request for Additional Information LaSalle County Station, Units 1 and 2 and Quad Cities Nuclear Power Station, Units 1 and 2 License Amendments Related to Fuel Storage RS-22-092, Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration2022-10-0303 October 2022 Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration RS-22-093, Advisement of Leadership Changes for Constellation Energy Generation, LLC and Submittal of Updated Standard Practice Procedures Plans2022-08-18018 August 2022 Advisement of Leadership Changes for Constellation Energy Generation, LLC and Submittal of Updated Standard Practice Procedures Plans RS-22-102, Supplement to Request to Revise Technical Specification 3.1.4, Control Rod Scam Times2022-08-18018 August 2022 Supplement to Request to Revise Technical Specification 3.1.4, Control Rod Scam Times RS-22-095, Response to Request for Additional Information Regarding Request to Expand Applicability of GNF Thermal Mechanical Analysis Methods to Framatome Fuel2022-08-10010 August 2022 Response to Request for Additional Information Regarding Request to Expand Applicability of GNF Thermal Mechanical Analysis Methods to Framatome Fuel RS-22-096, Response to Request for Additional Information Related to Relief Request I6R-01 Associated with the Sixth Inservice Inspection Interval2022-08-10010 August 2022 Response to Request for Additional Information Related to Relief Request I6R-01 Associated with the Sixth Inservice Inspection Interval RS-22-098, Response to Request for Additional Information for Quad Cities Relief Request RV-11, Code Case OMN-282022-08-0101 August 2022 Response to Request for Additional Information for Quad Cities Relief Request RV-11, Code Case OMN-28 RS-22-094, Submittal of Sixth Inservice Inspection Interval Relief Request I6R-09 to Permit Continued Application of Certain ASME Section XI 2013 Edition Non-Destructive Examination Requirements2022-07-25025 July 2022 Submittal of Sixth Inservice Inspection Interval Relief Request I6R-09 to Permit Continued Application of Certain ASME Section XI 2013 Edition Non-Destructive Examination Requirements RS-22-090, Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies2022-07-13013 July 2022 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies RS-22-077, Response to Request for Additional Information for Quad Cities Relief Requests RV-08, Safety Relief Valves2022-06-30030 June 2022 Response to Request for Additional Information for Quad Cities Relief Requests RV-08, Safety Relief Valves RS-22-078, Response to Request for Additional Information for Quad Cities Relief Request RV-09, MSSVs2022-06-30030 June 2022 Response to Request for Additional Information for Quad Cities Relief Request RV-09, MSSVs RS-22-080, Withdrawal and Proposed Alternative I6R-08 Associated with Code Case N-921 for Sixth Inservice Inspection Intervals2022-06-0909 June 2022 Withdrawal and Proposed Alternative I6R-08 Associated with Code Case N-921 for Sixth Inservice Inspection Intervals RS-22-069, Constellation Nuclear Radiological Emergency Plan Addendum Revision. Includes EP-AA-1006, Addendum 3, Revision 10, Emergency Action Levels for Quad Cities Station2022-05-25025 May 2022 Constellation Nuclear Radiological Emergency Plan Addendum Revision. Includes EP-AA-1006, Addendum 3, Revision 10, Emergency Action Levels for Quad Cities Station RS-22-059, CFR50.46 Annual Report2022-05-0404 May 2022 CFR50.46 Annual Report 2024-07-12
[Table view] Category:Report
MONTHYEARRS-24-046, 10 CFR 50.46 Annual Report2024-05-0606 May 2024 10 CFR 50.46 Annual Report RS-22-090, Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies2022-07-13013 July 2022 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies NMP1L3469, Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits2022-06-30030 June 2022 Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits RS-22-006, Request to Expand Applicability of Prime Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel2022-01-20020 January 2022 Request to Expand Applicability of Prime Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel SVP-21-041, Regulatory Commitment Change Summary Report2021-07-15015 July 2021 Regulatory Commitment Change Summary Report RS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021 Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting SVP-20-074, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report2020-10-0808 October 2020 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report RS-20-086, Renewed Facility Operating License Nos. DPR-29 and DPR-302020-09-24024 September 2020 Renewed Facility Operating License Nos. DPR-29 and DPR-30 ML19168A1962019-08-29029 August 2019 Staff Assessment of Flood Hazard Integrated Assessment RS-19-049, Submittal of 10 CFR 50.46 Annual Report2019-05-0202 May 2019 Submittal of 10 CFR 50.46 Annual Report ML19064B3702019-03-0505 March 2019 Enclosure a - QDC-7500-M-2341, Revision 0, Quad Cities Units 1 & 2 Secondary Containment Drawdown Analysis. ML19064B3712019-03-0505 March 2019 Enclosure B - QDC-0000-N-1481, Revision 3, Quad Cities Units 1 & 2 Post-LOCA Eab, LPZ, and CR Dose - AST Analysis. SVP-19-002, Submittal of 10 CFR 50.59, 10 CFR 72.48 Summary Report2019-01-0404 January 2019 Submittal of 10 CFR 50.59, 10 CFR 72.48 Summary Report RS-18-087, Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2018-08-14014 August 2018 Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) RS-18-089, Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (EA-13-109)2018-08-14014 August 2018 Report of Full Compliance with Phase 1 and Phase 2 of June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (EA-13-109) RS-18-048, Submittal of 10 CFR 50.46 Annual Report Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.2018-05-0202 May 2018 Submittal of 10 CFR 50.46 Annual Report Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors. SVP-17-046, Regulatory Commitment Change Summary Report2017-07-0606 July 2017 Regulatory Commitment Change Summary Report ML16271A3242016-10-31031 October 2016 Technical Review Documentation for Flooding Hazard Reevaluation of Quad Cities Nuclear Power Station, Units 1 and 2 RS-16-044, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima..2016-02-19019 February 2016 High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima.. ML15309A4932016-02-10010 February 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force ML15251A3842015-08-31031 August 2015 Attachment 9 - NP-3305NP, Revision 1, Mechanical Design Report for Quad Cities and Dresden Atrium 10XN Fuel Assemblies - Licensing Report. (Non-Proprietary) SVP-15-052, Submittal of Regulatory Commitment Change Summary Report2015-07-0808 July 2015 Submittal of Regulatory Commitment Change Summary Report RS-15-085, Provides Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentaiton (Order Number EA-12-051)2015-05-0707 May 2015 Provides Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentaiton (Order Number EA-12-051) RS-14-277, Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 12014-09-24024 September 2014 Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1 SVP-14-057, Owner'S Activity Report Submittal Fifth 10 Year Interval 2014 Refueling Outage Activities2014-08-0101 August 2014 Owner'S Activity Report Submittal Fifth 10 Year Interval 2014 Refueling Outage Activities ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML14101A2282014-04-0707 April 2014 Holtec International Report No. HI-2125245, Revision 5, Licensing Report for Quad Cities Criticality Analysis for Inserts. RS-14-072, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights..2014-03-31031 March 2014 Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights.. RS-13-288, Request for License Amendment to Technical Specifications Section 5.6.5. Core Operating Limits Report.2013-12-19019 December 2013 Request for License Amendment to Technical Specifications Section 5.6.5. Core Operating Limits Report. ML13317A9572013-11-20020 November 2013 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Quad Cities Nuclear Power Station, TAC Nos.: MF1048 and MF1049 ML13199A0382013-07-16016 July 2013 Attachment 5 - Holtec International Report No. HI-2125245, Revision 2, Licensing Report for Quad Cities Criticality Analysis for Inserts - Non Proprietary Version IR 05000456/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000289/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML13008A2192013-01-31031 January 2013 U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000454/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000461/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000237/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000373/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000352/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000219/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000277/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000254/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000272/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee SVP-13-001, Summary Report of Completed Changes, Tests and Experiments for January 2011 - December 20122013-01-0404 January 2013 Summary Report of Completed Changes, Tests and Experiments for January 2011 - December 2012 ML12361A0872012-10-31031 October 2012 Report 12Q0108.40-R-002, Rev. 3, Seismic Walkdown Report in Response to the 50.54(f) Information Request Re Fukushima Near-Term Force Recommendation 2.3: Seismic for Quad Cities Generating Station Unit 2. Part 5 of 15 ML12361A0862012-10-31031 October 2012 Report 12Q0108.40-R-002, Rev. 3, Seismic Walkdown Report in Response to the 50.54(f) Information Request Re Fukushima Near-Term Force Recommendation 2.3: Seismic for Quad Cities Generating Station Unit 2. Part 4 of 15 ML12361A0892012-10-31031 October 2012 Report 12Q0108.40-R-002, Rev. 3, Seismic Walkdown Report in Response to the 50.54(f) Information Request Re Fukushima Near-Term Force Recommendation 2.3: Seismic for Quad Cities Generating Station Unit 2. Part 6 of 15 ML12361A0902012-10-31031 October 2012 Report 12Q0108.40-R-002, Rev. 3, Seismic Walkdown Report in Response to the 50.54(f) Information Request Re Fukushima Near-Term Force Recommendation 2.3: Seismic for Quad Cities Generating Station Unit 2. Part 7 of 15 ML12361A0912012-10-31031 October 2012 Report 12Q0108.40-R-002, Rev. 3, Seismic Walkdown Report in Response to the 50.54(f) Information Request Re Fukushima Near-Term Force Recommendation 2.3: Seismic for Quad Cities Generating Station Unit 2. Part 8 of 15 2024-05-06
[Table view] Category:Miscellaneous
MONTHYEARSVP-21-041, Regulatory Commitment Change Summary Report2021-07-15015 July 2021 Regulatory Commitment Change Summary Report ML20303A1752020-10-23023 October 2020 Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting ML19168A1962019-08-29029 August 2019 Staff Assessment of Flood Hazard Integrated Assessment SVP-19-002, Submittal of 10 CFR 50.59, 10 CFR 72.48 Summary Report2019-01-0404 January 2019 Submittal of 10 CFR 50.59, 10 CFR 72.48 Summary Report RS-18-048, Submittal of 10 CFR 50.46 Annual Report Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.2018-05-0202 May 2018 Submittal of 10 CFR 50.46 Annual Report Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors. SVP-17-046, Regulatory Commitment Change Summary Report2017-07-0606 July 2017 Regulatory Commitment Change Summary Report RS-16-044, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima..2016-02-19019 February 2016 High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima.. ML15309A4932016-02-10010 February 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force SVP-15-052, Submittal of Regulatory Commitment Change Summary Report2015-07-0808 July 2015 Submittal of Regulatory Commitment Change Summary Report RS-15-085, Provides Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentaiton (Order Number EA-12-051)2015-05-0707 May 2015 Provides Report of Full Compliance with March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentaiton (Order Number EA-12-051) SVP-14-057, Owner'S Activity Report Submittal Fifth 10 Year Interval 2014 Refueling Outage Activities2014-08-0101 August 2014 Owner'S Activity Report Submittal Fifth 10 Year Interval 2014 Refueling Outage Activities RS-14-072, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights..2014-03-31031 March 2014 Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re Recommendation 2.1 of the Near-Term Task Force Review of Insights.. IR 05000456/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000272/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000254/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000277/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000219/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000352/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000373/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000289/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee ML13008A2192013-01-31031 January 2013 U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000454/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000461/20132012013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee IR 05000237/20132022013-01-31031 January 2013 Exelon Generation Co., LLC, U.S. Nuclear Regulatory Commission (NRC) Office of Investigations (01) Investigation; Summary of 01 Report No. 3-2010-034; NRC Inspection Report Conference Letter - Licensee SVP-13-001, Summary Report of Completed Changes, Tests and Experiments for January 2011 - December 20122013-01-0404 January 2013 Summary Report of Completed Changes, Tests and Experiments for January 2011 - December 2012 ML12178A2152012-08-0202 August 2012 Closeout of Bulletin 2011-01 Migrating Strategies. SVP-12-062, Post Outage Inservice Inspection (ISI) Summary Report2012-07-13013 July 2012 Post Outage Inservice Inspection (ISI) Summary Report ML1127000692011-09-26026 September 2011 Enclosure 2, Mfn 10-245 R4, Description of the Evaluation and Surveillance Recommendations for BWR/2-5 Plants SVP-11-050, Regulatory Commitment Change Summary Report2011-06-11011 June 2011 Regulatory Commitment Change Summary Report ML1108712502011-03-24024 March 2011 BWR Vessel and Internals Inspection Summaries for Spring 2010 Outages ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants ML0802800922008-01-15015 January 2008 U.S. Dept of Homeland Security/Quad Cities Nuclear Power Station Medical Services (MS-1) Drill Report/Report ML0731000992007-11-0606 November 2007 Federal Emergency Management Agency, Final Report for September 28, 2007 Medical Service Drill of the Offsite Radiological Emergency Response Plans Site-Specific to the Quad Cities Nuclear Power Station ML0711004532007-04-0404 April 2007 Westinghouse Application for Withholding, Affidavit, and Non-Proprietary Version of Attachment 2 (Attachment 3) SVP-06-081, Extension of the Provisional Variance of National Pollutant Discharge Elimination System Permit No. IL00050372006-08-10010 August 2006 Extension of the Provisional Variance of National Pollutant Discharge Elimination System Permit No. IL0005037 RS-06-113, Startup and Power Ascension Testing Following Installation of Acoustic Side Branch Modifications2006-08-0202 August 2006 Startup and Power Ascension Testing Following Installation of Acoustic Side Branch Modifications ML0600301402005-11-30030 November 2005 Enclosure 6 - Non-Proprietary GE Report GE-NE-0000-0048-1485-01-NP, Damping Value for Steam Dryer Hood Focused on a Single Frequency Regime. SVP-05-058, Regulatory Commitment Change Summary Report2005-07-28028 July 2005 Regulatory Commitment Change Summary Report SVP-05-043, Core Operating Limits Report for Plant Cycle 19, Rev 12005-05-31031 May 2005 Core Operating Limits Report for Plant Cycle 19, Rev 1 ML0611504632005-04-30030 April 2005 C.D.I. Technical Note No. 05-04, Quad Cities 2 New Dryer Smt Loads, Revision 5, Dated April 2005, Non-Proprietary SVP-05-031, Occupational Radiation Exposure Report for January Through December 20042005-04-22022 April 2005 Occupational Radiation Exposure Report for January Through December 2004 RS-05-040, Technical Documentation Related to Analysis and Design of New Quad Cities Steam Dryers, and Responses to Requests for Additional Information Related to EPU Operation at Dresden and Quad Cities Nuclear Power Stations2005-04-0101 April 2005 Technical Documentation Related to Analysis and Design of New Quad Cities Steam Dryers, and Responses to Requests for Additional Information Related to EPU Operation at Dresden and Quad Cities Nuclear Power Stations ML0611504572005-01-31031 January 2005 C.D.I. Report No. 04-09P, Methodology to Determine Unsteady Pressure Loading on Components in Reactor Steam Domes, Revision 5, Dated January 2005, Non-Proprietary ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement SVP-05-001, Summary Report of Changes, Tests, and Experiments Completed2005-01-0303 January 2005 Summary Report of Changes, Tests, and Experiments Completed SVP-04-067, Regulatory Commitment Change Summary Report2004-08-13013 August 2004 Regulatory Commitment Change Summary Report ML0309006662003-03-19019 March 2003 Additional Information Regarding Notice of Proposed Amendments to Trust Agreement to Implement Assignment of Decommissioning Trust Funds for Exelon Generation Company, LLC SVP-03-034, Day Post Outage ISI Letter2003-02-24024 February 2003 Day Post Outage ISI Letter SVP-03-002, 10 CFR 50.59 Summary Report2003-01-0303 January 2003 10 CFR 50.59 Summary Report ML0326713602002-08-22022 August 2002 Calculation QDC-0000-N-1117, Site Boundary and Control Doses Following a Loss of Coolant Accident Using Alternative Source Terms, Revision 0 2021-07-15
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4300 Winfield Road Warrenville, IL 60555 Exelon Generation 630 657 2000 Office RS-18-048 10 CFR 50.46 May 2, 2018 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265
Subject:
10 CFR 50.46 Annual Report
Reference:
Letter RS-17-052 from P.R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 2, 2017 This letter provides the annual report required by 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, " for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The attachments describe the changes in accumulated peak cladding temperature (PCT) since the previous annual report submitted in the referenced letter.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (630) 657-2831 .
P rick R. Simpso~
Manager - Licensing Attachments:
- 1. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5
- 2. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5
- 3. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6
- 4. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel)- USA6
- 5. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (AREVA Fuel)
- 6. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (AREVA Fuel)
- 7. Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Quad Cities Nuclear Power Station
ATTACHMENT 1 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
OPTIMA2-TR021QC-LOCA, Revision 5, September 2009
- 2. "Quad Cities Nuclear Power Station Unit 1 Cycle 23 MAPLHGR Report," NF-BEX-12-188-NP, Revision 0, January 2013 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 7, 2010 (See Note 1) 'PCT = 0°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) 'PCT = 0°F 10 CFR 50.46 Report dated May 4, 2012 (See Note 3) 'PCT = 0°F 10 CFR 50.46 Report dated May 3, 2013 (See Note 4) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2014 (See Note 5) 'PCT = 0°F 10 CFR 50.46 Report dated May 1, 2015 (See Note 6) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 2 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
OPTIMA2-TR021QC-LOCA, Revision 5, September 2009
- 2. "Quad Cities Nuclear Power Station Unit 2 Cycle 23 MAPLHGR Report," NF-BEX-13-168-NP, Revision 0, January 2014 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 7, 2010 (See Note 1) 'PCT = 0°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) 'PCT = 0°F 10 CFR 50.46 Report dated May 4, 2012 (See Note 3) 'PCT = 0°F 10 CFR 50.46 Report dated May 3, 2013 (See Note 4) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2014 (See Note 5) 'PCT = 0°F 10 CFR 50.46 Report dated May 1, 2015 (See Note 6) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 3 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA6 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model Updates:
Supplement 4 to Code Description, Qualification and Application,"
WCAP-16865-P-A, Revision 1, October 2011 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
NF-BEX-13-143-P, Revision 3, April 2016
- 2. "Quad Cities Nuclear Power Station Unit 1 Cycle 24 Supplemental MAPLHGR Report," NF-BEX-14-135-NP, Revision 1, October 2015 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2017 (Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 4 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA6 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model Updates:
Supplement 4 to Code Description, Qualification and Application,"
WCAP-16865-P-A, Revision 1, October 2011 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
NF-BEX-13-143-P, Revision 3, April 2016
- 2. "Quad Cities Nuclear Power Station Unit 2 Cycle 24 MAPLHGR Report," NF-BEX-15-174-NP, Revision 0, January 2016 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 5 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (AREVA Fuel)
PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: EXEM BWR-2000 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "EXEM BWR-2000 ECCS Evaluation Model," EMF-2361(P)(A),
Revision 0, May 2001 Calculations: 1. "Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," ANP-3328P, Revision 2, December 2016
- 2. "Quad Cities Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel," ANP-3356P, Revision 2, February 2017 Fuel Analyzed in Calculation: ATRIUM 10XM Limiting Fuel Type: ATRIUM 10XM Limiting Single Failure: High Pressure Coolant Injection system Limiting Break Size and Location: 0.13 ft2 split break recirculation discharge pipe Reference Peak Cladding Temperature (PCT): PCT = 2138°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2138°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2138°F
ATTACHMENT 6 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (AREVA Fuel)
PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: EXEM BWR-2000 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "EXEM BWR-2000 ECCS Evaluation Model," EMF-2361(P)(A),
Revision 0, May 2001 Calculations: 1. "Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," ANP-3328P, Revision 2, December 2016
- 2. "Quad Cities Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel," ANP-3356P, Revision 3, January 2018 Fuel Analyzed in Calculation: ATRIUM 10XM Limiting Fuel Type: ATRIUM 10XM Limiting Single Failure: High Pressure Coolant Injection system Limiting Break Size and Location: 0.13 ft2 split break recirculation discharge pipe Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS New LOCA Model (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes
- 1. Prior Loss-of-Coolant Accident (LOCA) Assessment The referenced letter reported a new Westinghouse LOCA analysis of record (AOR),
which incorporated modifications for the newly added recirculation pump adjustable speed drives (ASD). The new AOR (Revision 5) updated the LOCA analysis to apply approved updated Westinghouse methods, incorporate previous corrections, and update plant specific inputs. The new Westinghouse LOCA analysis demonstrated that the limiting peak cladding temperature (PCT) was 2150°F.
The referenced letter also reported the impact of subsequent corrections for an incorrect bypass hole flow coefficient, and for updated vessel leakage values. The PCT impact of these two corrections on the limiting PCT was 9°F and 2°F, respectively. These PCT impacts will remain in effect only until the MAPLHGR limits for all bundles in future QCNPS Units 1 and 2 cores are evaluated for this change. The MAPLHGR limits for all bundles in the QCNPS Unit 1 cores starting with Q1C25 and QCNPS Unit 2 cores starting with Q2C25 were evaluated for these changes and these penalties no longer apply.
[
Reference:
Letter from J. L. Hansen (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,' Annual Report," dated May 7, 2010]
- 2. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no emergency core cooling system (ECCS)-related changes or modifications occurred at Quad Cities Nuclear Power Station (QCNPS) that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from J. L. Hansen (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,' Annual Report," dated May 6, 2011]
- 3. Prior LOCA Assessment The referenced letter provided the annual 10 CFR 50.46 report for Units 1 and 2. The letter reported errors in the current Westinghouse QCNPS LOCA analysis associated with the use of incorrect R-factors. The impact due to this change was determined to be 18°F increase in PCT. This PCT will remain in effect only until the MAPLHGR limits for all bundles in future QCNPS Units 1 and 2 cores are evaluated for this change. The MAPLHGR limits for all bundles in the QCNPS Unit 1 cores starting with Q1C25 and QCNPS Unit 2 cores starting with Q2C25 were evaluated for this change and this penalty no longer apply.
[
Reference:
Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Annual Report, dated May 4, 2012]
Page 1
ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes
- 4. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 3, 2013]
- 5. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 2, 2014]
- 6. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 1, 2015]
- 7. Prior LOCA Assessment The new USA6 LOCA Evaluation Model and calculation was implemented for QCNPS Unit 2 for all fresh fuel starting with Cycle 24. This evaluation model and calculation supplement the existing Evaluation Model and calculation. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation Model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 2, 2016]
- 8. Prior LOCA Assessment The new USA6 LOCA Evaluation Model and calculation was implemented for QCNPS Unit 1 during Cycle 24 for the Optima2 fuel loaded fresh into the Cycle 24 core. This evaluation model and calculation supplement the existing Evaluation Model and calculation. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
Page 2
ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes The new AREVA EXEM BWR-2000 Evaluation Model and calculation has been implemented for QCNPS Unit 1 for the fresh fuel of ATRIUM 10XM loaded into the Cycle 25 core. This model reports a limiting PCT of 2138°F and has no additional PCT impacts or estimates.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation model, USA6 LOCA Evaluation model or EXEM BWR-2000 LOCA Evaluation model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 2, 2017]
- 9. Current Assessment The new AREVA EXEM BWR-2000 Evaluation Model and calculation has been implemented for QCNPS Unit 2 for the fresh fuel of ATRIUM 10XM loaded into the Cycle 25 core. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
Additionally, updated LOCA/MAPLHGR analyses were performed for QCNPS, Unit 1 and Unit 2 for the Westinghouse fuel first introduced in their respective Cycle 23 reload. These analyses for QCNPS, Unit 1 and Unit 2 maintained the calculated PCT at 2150°F and accounted for the incorrect vessel leakage and bypass hole coefficient outlined in Note 1 and the incorrect R-factors outlined in Note 3. All other Westinghouse fuel from before Cycle 23 that was impacted by these errors within Unit 1 and 2 has been discharged. Therefore, the previous penalty of 11°F in Note 1 and 18°F in Note 3 no longer apply for QCNPS, Unit 1 and Unit 2. The penalty in Attachment 1 and 2 listed for Note 1 and 3 has all been changed to 0°F.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation model, USA6 LOCA Evaluation model or EXEM BWR-2000 LOCA Evaluation model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
Page 3
4300 Winfield Road Warrenville, IL 60555 Exelon Generation 630 657 2000 Office RS-18-048 10 CFR 50.46 May 2, 2018 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265
Subject:
10 CFR 50.46 Annual Report
Reference:
Letter RS-17-052 from P.R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 2, 2017 This letter provides the annual report required by 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, " for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The attachments describe the changes in accumulated peak cladding temperature (PCT) since the previous annual report submitted in the referenced letter.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (630) 657-2831 .
P rick R. Simpso~
Manager - Licensing Attachments:
- 1. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5
- 2. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5
- 3. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6
- 4. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel)- USA6
- 5. Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (AREVA Fuel)
- 6. Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (AREVA Fuel)
- 7. Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Quad Cities Nuclear Power Station
ATTACHMENT 1 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
OPTIMA2-TR021QC-LOCA, Revision 5, September 2009
- 2. "Quad Cities Nuclear Power Station Unit 1 Cycle 23 MAPLHGR Report," NF-BEX-12-188-NP, Revision 0, January 2013 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 7, 2010 (See Note 1) 'PCT = 0°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) 'PCT = 0°F 10 CFR 50.46 Report dated May 4, 2012 (See Note 3) 'PCT = 0°F 10 CFR 50.46 Report dated May 3, 2013 (See Note 4) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2014 (See Note 5) 'PCT = 0°F 10 CFR 50.46 Report dated May 1, 2015 (See Note 6) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 2 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA5 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
OPTIMA2-TR021QC-LOCA, Revision 5, September 2009
- 2. "Quad Cities Nuclear Power Station Unit 2 Cycle 23 MAPLHGR Report," NF-BEX-13-168-NP, Revision 0, January 2014 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 7, 2010 (See Note 1) 'PCT = 0°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) 'PCT = 0°F 10 CFR 50.46 Report dated May 4, 2012 (See Note 3) 'PCT = 0°F 10 CFR 50.46 Report dated May 3, 2013 (See Note 4) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2014 (See Note 5) 'PCT = 0°F 10 CFR 50.46 Report dated May 1, 2015 (See Note 6) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 3 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA6 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model Updates:
Supplement 4 to Code Description, Qualification and Application,"
WCAP-16865-P-A, Revision 1, October 2011 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
NF-BEX-13-143-P, Revision 3, April 2016
- 2. "Quad Cities Nuclear Power Station Unit 1 Cycle 24 Supplemental MAPLHGR Report," NF-BEX-14-135-NP, Revision 1, October 2015 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2017 (Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 4 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) - USA6 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA6 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "Westinghouse BWR ECCS Evaluation Model Updates:
Supplement 4 to Code Description, Qualification and Application,"
WCAP-16865-P-A, Revision 1, October 2011 Calculations: 1. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"
NF-BEX-13-143-P, Revision 3, April 2016
- 2. "Quad Cities Nuclear Power Station Unit 2 Cycle 24 MAPLHGR Report," NF-BEX-15-174-NP, Revision 0, January 2016 Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location: 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2016 (See Note 7) 'PCT = 0°F 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 5 Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (AREVA Fuel)
PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: EXEM BWR-2000 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "EXEM BWR-2000 ECCS Evaluation Model," EMF-2361(P)(A),
Revision 0, May 2001 Calculations: 1. "Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," ANP-3328P, Revision 2, December 2016
- 2. "Quad Cities Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel," ANP-3356P, Revision 2, February 2017 Fuel Analyzed in Calculation: ATRIUM 10XM Limiting Fuel Type: ATRIUM 10XM Limiting Single Failure: High Pressure Coolant Injection system Limiting Break Size and Location: 0.13 ft2 split break recirculation discharge pipe Reference Peak Cladding Temperature (PCT): PCT = 2138°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated May 2, 2017 (See Note 8) 'PCT = 0°F Net PCT 2138°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2138°F
ATTACHMENT 6 Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (AREVA Fuel)
PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: EXEM BWR-2000 REPORT REVISION DATE: 05/02/2018 CURRENT OPERATING CYCLE: 25 ANALYSIS OF RECORD Evaluation Model: "EXEM BWR-2000 ECCS Evaluation Model," EMF-2361(P)(A),
Revision 0, May 2001 Calculations: 1. "Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," ANP-3328P, Revision 2, December 2016
- 2. "Quad Cities Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel," ANP-3356P, Revision 3, January 2018 Fuel Analyzed in Calculation: ATRIUM 10XM Limiting Fuel Type: ATRIUM 10XM Limiting Single Failure: High Pressure Coolant Injection system Limiting Break Size and Location: 0.13 ft2 split break recirculation discharge pipe Reference Peak Cladding Temperature (PCT): PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None 'PCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS New LOCA Model (See Note 9) 'PCT = 0°F Total PCT change from current assessments 'PCT = 0°F Cumulative PCT change from current assessments ~'PCT~ = 0°F Net PCT 2150°F
ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes
- 1. Prior Loss-of-Coolant Accident (LOCA) Assessment The referenced letter reported a new Westinghouse LOCA analysis of record (AOR),
which incorporated modifications for the newly added recirculation pump adjustable speed drives (ASD). The new AOR (Revision 5) updated the LOCA analysis to apply approved updated Westinghouse methods, incorporate previous corrections, and update plant specific inputs. The new Westinghouse LOCA analysis demonstrated that the limiting peak cladding temperature (PCT) was 2150°F.
The referenced letter also reported the impact of subsequent corrections for an incorrect bypass hole flow coefficient, and for updated vessel leakage values. The PCT impact of these two corrections on the limiting PCT was 9°F and 2°F, respectively. These PCT impacts will remain in effect only until the MAPLHGR limits for all bundles in future QCNPS Units 1 and 2 cores are evaluated for this change. The MAPLHGR limits for all bundles in the QCNPS Unit 1 cores starting with Q1C25 and QCNPS Unit 2 cores starting with Q2C25 were evaluated for these changes and these penalties no longer apply.
[
Reference:
Letter from J. L. Hansen (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,' Annual Report," dated May 7, 2010]
- 2. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no emergency core cooling system (ECCS)-related changes or modifications occurred at Quad Cities Nuclear Power Station (QCNPS) that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from J. L. Hansen (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,' Annual Report," dated May 6, 2011]
- 3. Prior LOCA Assessment The referenced letter provided the annual 10 CFR 50.46 report for Units 1 and 2. The letter reported errors in the current Westinghouse QCNPS LOCA analysis associated with the use of incorrect R-factors. The impact due to this change was determined to be 18°F increase in PCT. This PCT will remain in effect only until the MAPLHGR limits for all bundles in future QCNPS Units 1 and 2 cores are evaluated for this change. The MAPLHGR limits for all bundles in the QCNPS Unit 1 cores starting with Q1C25 and QCNPS Unit 2 cores starting with Q2C25 were evaluated for this change and this penalty no longer apply.
[
Reference:
Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Annual Report, dated May 4, 2012]
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ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes
- 4. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 3, 2013]
- 5. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 2, 2014]
- 6. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysis. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 1, 2015]
- 7. Prior LOCA Assessment The new USA6 LOCA Evaluation Model and calculation was implemented for QCNPS Unit 2 for all fresh fuel starting with Cycle 24. This evaluation model and calculation supplement the existing Evaluation Model and calculation. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation Model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S.
NRC, "10 CFR 50.46 Annual Report," dated May 2, 2016]
- 8. Prior LOCA Assessment The new USA6 LOCA Evaluation Model and calculation was implemented for QCNPS Unit 1 during Cycle 24 for the Optima2 fuel loaded fresh into the Cycle 24 core. This evaluation model and calculation supplement the existing Evaluation Model and calculation. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
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ATTACHMENT 7 Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes The new AREVA EXEM BWR-2000 Evaluation Model and calculation has been implemented for QCNPS Unit 1 for the fresh fuel of ATRIUM 10XM loaded into the Cycle 25 core. This model reports a limiting PCT of 2138°F and has no additional PCT impacts or estimates.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation model, USA6 LOCA Evaluation model or EXEM BWR-2000 LOCA Evaluation model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
[
Reference:
Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "10 CFR 50.46 Annual Report," dated May 2, 2017]
- 9. Current Assessment The new AREVA EXEM BWR-2000 Evaluation Model and calculation has been implemented for QCNPS Unit 2 for the fresh fuel of ATRIUM 10XM loaded into the Cycle 25 core. This model reports a limiting PCT of 2150°F and has no additional PCT impacts or estimates.
Additionally, updated LOCA/MAPLHGR analyses were performed for QCNPS, Unit 1 and Unit 2 for the Westinghouse fuel first introduced in their respective Cycle 23 reload. These analyses for QCNPS, Unit 1 and Unit 2 maintained the calculated PCT at 2150°F and accounted for the incorrect vessel leakage and bypass hole coefficient outlined in Note 1 and the incorrect R-factors outlined in Note 3. All other Westinghouse fuel from before Cycle 23 that was impacted by these errors within Unit 1 and 2 has been discharged. Therefore, the previous penalty of 11°F in Note 1 and 18°F in Note 3 no longer apply for QCNPS, Unit 1 and Unit 2. The penalty in Attachment 1 and 2 listed for Note 1 and 3 has all been changed to 0°F.
Additionally, there are no new changes, error corrections, or enhancements in the current QCNPS LOCA analysis based on the USA5 LOCA Evaluation model, USA6 LOCA Evaluation model or EXEM BWR-2000 LOCA Evaluation model. Also, no ECCS-related changes or modifications occurred at QCNPS that affected the assumptions in the ECCS analyses.
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