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| {{#Wiki_filter:Exelon Nuclear Security, LLC 4300 Winfield Road Warrenville, IL 60555 RS-12-0BO www.exeloncorp.com Nuclear 10 CFR 50.46(a)(3)(ii) | | {{#Wiki_filter:Exelon Nuclear Security, LLC www.exeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL 60555 RS-12-0BO 10 CFR 50.46(a)(3)(ii) |
| May 4,2012 u.s. Nuclear Regulatory Commission A TIN: 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 | | May 4,2012 u.s. Nuclear Regulatory Commission ATIN: 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 |
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| ==Subject:== | | ==Subject:== |
| 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for water nuclear power reactors," Annual Report | | 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Annual Report |
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| ==Reference:== | | ==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 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, Units 1 and 2. The attachments describe the changes in accumulated peak cladding temperature since the previous annual report submitted in the referenced letter. |
| | This report covers the period from May 6, 2011, through May 4, 2012. |
| | There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2B04. |
| | Respectfully, 7?fJ~ |
| | David M. Gullott Manager - Licensing Attachments: |
| | : 1. Quad Cities Nuclear Power Station, Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) |
| | : 2. Quad Cities Nuclear Power Station, Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) |
| | : 3. Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes |
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| 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,'
| | ATTACHMENT 1 Quad Cities Nuclear Power Station, Unit 1 10 CFR 50.46 Report (Westinghouse Fuel) |
| Annual Report," dated May 6, 2011 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, Units 1 and 2. The attachments describe the changes in accumulated peak cladding temperature since the previous annual report submitted in the referenced letter. This report covers the period from May 6, 2011, through May 4, 2012. There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2B04.
| | Page 1 of 1 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 04/13/2012 CURRENT OPERATING CYCLE: 22 ANALYSIS OF RECORD Evaluation Model: 'Westinghouse BWR ECCS Evaluation Model: |
| Respectfully, David M. Gullott Manager -Licensing Attachments:
| | Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004. |
| : 1. Quad Cities Nuclear Power Station, Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel) 2. Quad Cities Nuclear Power Station, Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) 3. Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes ATTACHMENT 1 Quad Cities Nuclear Power Station, Unit 1 10 CFR 50.46 Report (Westinghouse Fuel) Page 1 of 1 PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: ANALYSIS OF RECORD Quad Cities Unit 1 USA5 04/13/2012 22 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: "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009. Fuel Analyzed in Calculation:
| | Calculations: "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009. |
| SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and 1.0 double-ended guillotine break in the recirculation pump Location: | | Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and 1.0 double-ended guillotine break in the recirculation pump Location: 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 =11°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) ~PCT = OaF Net PCT 2161°F B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) ~PCT = 18°F Total PCT change from current assessments L~PCT = 18°F Cumulative PCT change from current assessments L I~PCT I = 18° F Net PCT 2179°F |
| 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) | | |
| =11°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) | | ATTACHMENT 2 Quad Cities Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) |
| = OaF Net PCT 2161°F B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) | | Page 1 of 1 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 04/13/2012 CURRENT OPERATING CYCLE: 22 ANALYSIS OF RECORD Evaluation Model: 'Westinghouse BWR ECCS Evaluation Model: |
| = 18°F Total PCT change from current assessments | | Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004. |
| = 18°F Cumulative PCT change from current assessments L I I = 18° F Net PCT 2179°F ATTACHMENT 2 Quad Cities Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) Page 1 of 1 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 04/13/2012 CURRENT OPERATING CYCLE: 22 ANALYSIS OF RECORD Evaluation Model: 'Westinghouse BWR ECCS Evaluation Model: Calculations: | | Calculations: "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009. |
| Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004. "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009. Fuel Analyzed in Calculation: | | Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and 1.0 double-ended guillotine break in the recirculation Location: pump suction line Reference Peak Cladding Temperature (PCT) |
| SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and Location: | | MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CF10 CFR 50.46 Report dated May 7,2010 (See Note 1) ~PCT =11°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) ~PCT = OaF NetPCT 2161°F B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) ~PCT = 18°F Total PCT change from current assessments L~PCT = 18°F Cumulative PCT change from current assessments L I~PCTI = 18°F Net PCT 2179°F |
| 1.0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT) MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CF10 CFR 50.46 Report dated May 7,2010 (See Note 1) 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) NetPCT B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) Total PCT change from current assessments Cumulative PCT change from current assessments Net PCT | | |
| =11°F
| | ATTACHMENT 3 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes Page 1 of 1 |
| = OaF 2161°F
| | : 1. Prior Loss-of-Coolant Accident (LOCA) Assessment The referenced letter reported a new Westinghouse LOCA analysis of record (AOR), |
| = 18°F
| | 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 of. 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 of and 2 of, respectively. |
| = 18°F L | |
| = 18°F 2179°F ATTACHMENT 3 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes Page 1 of 1 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 of. 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 of and 2 of, respectively.
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| ==Reference:== | | ==Reference:== |
| | | Letter from Jeffery Hansen (Exelon Generation Company, LLC) to |
| Letter from Jeffery 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,' | | : u. S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,' Annual Reporf' for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, dated May 7,2010] |
| Annual Reporf' for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, dated May 7,2010] 2. Prior LOCA Assessment The referenced letter reported no new PCT assessment for the Westinghouse LOCA analysiS. | | : 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 QCNPS that affected the assumptions of the ECCS analyses. |
| Also, no emergency core cooling system (ECCS)-related changes or modifications occurred at QCNPS that affected the assumptions of the ECCS analyses. | |
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| ==Reference:== | | ==Reference:== |
| | | Letter from Jeffery Hansen (Exelon) to U.S. NRC, "10 CFR 50.46, |
| Letter from Jeffery Hansen (Exelon) to U.S. NRC, "10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,' | | 'Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,' Annual Reporf' for QCNPS, Units 1 and 2, dated May 6, 2011] |
| Annual Reporf' for QCNPS, Units 1 and 2, dated May 6, 2011] 3. Current Assessment In the referenced letter, Westinghouse reported errors in the current QCNPS LOCA analysis associated with the use of incorrect R-factors. | | : 3. Current Assessment In the referenced letter, Westinghouse reported errors in the current QCNPS LOCA analysis associated with the use of incorrect R-factors. The impact of this change was determined to be an 18 of increase in PCT. For 10 CFR 50.46 reporting purposes, the PCT increase is conservatively applied to all bundle types including the fresh bundles. |
| The impact of this change was determined to be an 18 of increase in PCT. For 10 CFR 50.46 reporting purposes, the PCT increase is conservatively applied to all bundle types including the fresh bundles. This PCT increase will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future QCNPS, Unit 1 and Unit 2 cores are evaluated with the correct R-factors. | | This PCT increase will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future QCNPS, Unit 1 and Unit 2 cores are evaluated with the correct R-factors. |
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| ==Reference:== | | ==Reference:== |
| | | Westinghouse letter LTR-LAM-11-63, Revision 0, "Dresden Units 2 & 3 and Quad Cities Units 1 & 2, 10 CFR 50.46 Notification and Report for R-Factor Error Evaluation," September 14,2011]}} |
| Westinghouse letter L TR-LAM-11-63, Revision 0, "Dresden Units 2 & 3 and Quad Cities Units 1 & 2, 10 CFR 50.46 Notification and Report for R-Factor Error Evaluation," September 14,2011]}} | |
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Category:Annual Operating Report
MONTHYEARRS-22-067, 10 CFR 50.46 Annual Report2023-05-0404 May 2023
10 CFR 50.46 Annual Report
RS-22-059, CFR50.46 Annual Report2022-05-0404 May 2022
CFR50.46 Annual Report
RS-21-052, 10 CFR 50.46 Annual Report2021-05-0404 May 2021
10 CFR 50.46 Annual Report
RS-20-059, 10 CFR 50.46 Annual Report2020-05-0404 May 2020
10 CFR 50.46 Annual Report
SVP-18-029, Submittal of Radioactive Effluent Release Report for 20172018-04-27027 April 2018
Submittal of Radioactive Effluent Release Report for 2017
RS-17-052, Transmittal of 10 CFR 50.46 Annual Report2017-05-0202 May 2017
Transmittal of 10 CFR 50.46 Annual Report
RS-16-093, 10 CFR 50.46 Annual Report2016-05-0202 May 2016
10 CFR 50.46 Annual Report
RS-15-104, Annual Property Insurance Status Report2015-04-0101 April 2015
Annual Property Insurance Status Report
RS-14-142, 10 CFR 50.46 Annual Report2014-05-0202 May 2014
10 CFR 50.46 Annual Report
SVP-13-035, Annual Radiological Environmental Operating Report, 1 January Through 31 December 2012, Page 93 of 122 Through End2013-05-31031 May 2013
Annual Radiological Environmental Operating Report, 1 January Through 31 December 2012, Page 93 of 122 Through End
ML13135A5832013-05-31031 May 2013
Annual Radiological Environmental Operating Report, 1 January Through 31 December 2012, Cover Through Page 92 of 122
SVP-12-045, Annual Radiological Environmental Operating Report2012-05-11011 May 2012
Annual Radiological Environmental Operating Report
RS-12-080, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, Annual Report2012-05-0404 May 2012
CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, Annual Report
ML12123A0192012-04-27027 April 2012
Submittal of Radioactive Effluent Release Report for 2011
SVP-11-042, Annual Radiological Environmental Operating Report2011-05-11011 May 2011
Annual Radiological Environmental Operating Report
RS-11-077, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report2011-05-0606 May 2011
CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report
RS-10-191, Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report2010-10-29029 October 2010
Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report
RS-10-087, CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report2010-05-0707 May 2010
CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report
SVP-09-018, Annual Radiological Environmental Operating Report for 1 January Through 31 December 20082009-05-12012 May 2009
Annual Radiological Environmental Operating Report for 1 January Through 31 December 2008
RS-09-059, 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report2009-05-0707 May 2009
10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report
RS-08-064, CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Annual Report2008-05-0707 May 2008
CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Annual Report
SVP-07-029, Annual Radiological Environmental Operating Report for 20062007-05-11011 May 2007
Annual Radiological Environmental Operating Report for 2006
SVP-07-001, CFR 50.59/10 CFR 72.48 Summary Report for January 2004 to December 31, 20062007-01-0303 January 2007
CFR 50.59/10 CFR 72.48 Summary Report for January 2004 to December 31, 2006
SVP-05-037, Annual Radiological Environmental Operating Report2005-05-13013 May 2005
Annual Radiological Environmental Operating Report
SVP-03-063, Transmittal of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, Annual Report for Quad Cities Nuclear Power Station, Units 1 & 22003-05-0808 May 2003
Transmittal of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, Annual Report for Quad Cities Nuclear Power Station, Units 1 & 2
2023-05-04
[Table view]
Category:Letter type:RS
MONTHYEARRS-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-23-060, 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-23-059, 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-23-042, 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-22-102, 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-22-093, 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-22-095, 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
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2024-01-03
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Text
Exelon Nuclear Security, LLC www.exeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL 60555 RS-12-0BO 10 CFR 50.46(a)(3)(ii)
May 4,2012 u.s. Nuclear Regulatory Commission ATIN: 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, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Annual Report
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 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, Units 1 and 2. The attachments describe the changes in accumulated peak cladding temperature since the previous annual report submitted in the referenced letter.
This report covers the period from May 6, 2011, through May 4, 2012.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2B04.
Respectfully, 7?fJ~
David M. Gullott Manager - Licensing Attachments:
- 1. Quad Cities Nuclear Power Station, Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel)
- 2. Quad Cities Nuclear Power Station, Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel)
- 3. Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes
ATTACHMENT 1 Quad Cities Nuclear Power Station, Unit 1 10 CFR 50.46 Report (Westinghouse Fuel)
Page 1 of 1 PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 04/13/2012 CURRENT OPERATING CYCLE: 22 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: "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009.
Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and 1.0 double-ended guillotine break in the recirculation pump Location: 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 =11°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) ~PCT = OaF Net PCT 2161°F B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) ~PCT = 18°F Total PCT change from current assessments L~PCT = 18°F Cumulative PCT change from current assessments L I~PCT I = 18° F Net PCT 2179°F
ATTACHMENT 2 Quad Cities Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel)
Page 1 of 1 PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 04/13/2012 CURRENT OPERATING CYCLE: 22 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: "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 5, Westinghouse Electric Company, LLC. September 2009.
Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-960ptima2 Limiting Single Failure: Low Pressure Coolant Injection System injection valve Limiting Break Size and 1.0 double-ended guillotine break in the recirculation Location: pump suction line Reference Peak Cladding Temperature (PCT)
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CF10 CFR 50.46 Report dated May 7,2010 (See Note 1) ~PCT =11°F 10 CFR 50.46 Report dated May 6, 2011 (See Note 2) ~PCT = OaF NetPCT 2161°F B. CURRENT LOCA MODEL ASSESSMENTS Incorrect R-factors (See Note 3) ~PCT = 18°F Total PCT change from current assessments L~PCT = 18°F Cumulative PCT change from current assessments L I~PCTI = 18°F Net PCT 2179°F
ATTACHMENT 3 Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46 Report Assessment Notes Page 1 of 1
- 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 of. 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 of and 2 of, respectively.
[
Reference:
Letter from Jeffery 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 Reporf' for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, 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 QCNPS that affected the assumptions of the ECCS analyses.
[
Reference:
Letter from Jeffery Hansen (Exelon) to U.S. NRC, "10 CFR 50.46,
'Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,' Annual Reporf' for QCNPS, Units 1 and 2, dated May 6, 2011]
- 3. Current Assessment In the referenced letter, Westinghouse reported errors in the current QCNPS LOCA analysis associated with the use of incorrect R-factors. The impact of this change was determined to be an 18 of increase in PCT. For 10 CFR 50.46 reporting purposes, the PCT increase is conservatively applied to all bundle types including the fresh bundles.
This PCT increase will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future QCNPS, Unit 1 and Unit 2 cores are evaluated with the correct R-factors.
[
Reference:
Westinghouse letter LTR-LAM-11-63, Revision 0, "Dresden Units 2 & 3 and Quad Cities Units 1 & 2, 10 CFR 50.46 Notification and Report for R-Factor Error Evaluation," September 14,2011]
