|
|
Line 16: |
Line 16: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:;. 1 tio RS-13-250 October 25, 2013 U. S. | | {{#Wiki_filter:1 |
| Regulatory Commission ArrN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 10 CFR 50.46 | | ;. |
| | tio RS-13-250 10 CFR 50.46 October 25, 2013 U. S. Nucl~ar Regulatory Commission ArrN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 |
|
| |
|
| ==Subject:== | | ==Subject:== |
| Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Dresden Nuclear Power Station, Units 2 and 3 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Exelon Generation Company, LLC (EGC) is submitting this letter and its attachments to meet the annual reporting requirements of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for Dresden Nuclear Power Station, Units 2 and 3. This report covers the period from October 27, 2012, through October 25, 2013. There are no regulatory commitments contained in this letter. If there are any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819. | | Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Dresden Nuclear Power Station, Units 2 and 3 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Exelon Generation Company, LLC (EGC) is submitting this letter and its attachments to meet the annual reporting requirements of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for Dresden Nuclear Power Station, Units 2 and 3. This report covers the period from October 27, 2012, through October 25, 2013. |
| Respectfu Ily, Patrick R. Simpson Manager -Licensing Exelon Generation Company, LLC Attachments: | | There are no regulatory commitments contained in this letter. If there are any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819. |
| : 1. Dresden Nuclear Power Station Unit 2 -10 CFR 50.46 Report (Westinghouse Fuel) 2. Dresden Nuclear Power Station Unit 3 -10 CFR 50.46 Report (Westinghouse Fuel) 3. Dresden Nuclear Power Station Units 2 and 3 -10 CFR 50.46 Report Assessment Notes Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) PLANT NAME: Dresden Nuclear Power Station, Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 10/3/2013 CURRENT OPERATING CYCLE: 23 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: "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC. September 2010. Fuel Analyzed in Calculation: | | Respectfu Ily, Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments: |
| SVEA-960ptima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure: Low Pressure Coolant Injection system injection valve Limiting Break Size and Location:
| | : 1. Dresden Nuclear Power Station Unit 2 - 10 CFR 50.46 Report (Westinghouse Fuel) |
| 1.0 Double-Ended Guillotine Break in the Recirculation Pump Suction Line Reference Peak Cladding 2150°F Temperature (PCT) Page 1 of 2 MARGIN ALLOCATION Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) A. PRIOR LOCA MODEL ASSESSMENTS I 10 CFR 50.46 report dated October 30, 2009 (See Note 1) 10 CFR 50.46 report dated October 29, 2010 (See Note 2) 10 CFR 50.46 report dated October 28, 2011 (See Note 3) 10 CFR 50.46 report dated October 26,2012 (See Note 4) Net PCT B. CURRENT LOCA MODEL ASSESSMENTS
| | : 2. Dresden Nuclear Power Station Unit 3 - 10 CFR 50.46 Report (Westinghouse Fuel) |
| = 2°F
| | : 3. Dresden Nuclear Power Station Units 2 and 3 - 10 CFR 50.46 Report Assessment Notes |
| = 12°F
| |
| = 18°F
| |
| = OaF 2182°F None (see Note 5)
| |
| = OaF Total PCT change from current assessments
| |
| = OaF Cumulative PCT change from current assessments T I = OaF Net PCT Page 2 of 2 Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel) PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: ANALYSIS OF RECORD Evaluation Model: Calculations:
| |
| Fuel Analyzed in Calculation:
| |
| Limiting Fuel Type: Limiting Single Failure: Limiting Break Size and Location:
| |
| Reference Peak Cladding Temperature (PCT) Dresden Nuclear Power Station, Unit 3 USA5 10/3/2013 23 "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004. "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTI MA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC. September 2010. SVEA-96 Optima2 SVEA-96 Optima2 Low Pressure Coolant Injection system injection valve 1.0 Double-Ended Guillotine Break in the Recirculation Pump Suction Line Page 1 of 2 MARGIN ALLOCATION Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel) A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated October 30, 2009 (See !'Jote 1) 10 CFR 50.46 report dated October 29, 2010 (See Note 2) 10 CFR 50.46 report dated October 28,2011 (See Note 3) 10 CFR 50.46 report dated October 26,2012 (See Note 4) Net PCT B. CURRENT LOCA MODEL ASSESSMENTS None (See !'Jote 5) Total PCT change from current assessments Cumulative PCT change from current assessments Net PCT Page 2 of 2
| |
| = 2°F
| |
| = 12°F
| |
| = 18°F
| |
| =OoF 2182°F
| |
| =O°F L I = OaF 2182°F Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes Assessment Notes 1. Prior Loss of Coolant Accident (LOCA) Model Assessment The referenced letter provided the annual 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors," report for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The letter updated the vessel leakage between the lower shroud and the downcomer.
| |
| Westinghouse evaluated this change and demonstrated that all 10 CFR 50.46 criteria were satisfied.
| |
| This evaluation resulted in maximum peak cladding temperature (PCT) impact due to the change in vessel leakage of 2°F for Optima2 fuel with the licensing basis PCT of 2152°F. This PCT update will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for this change. [
| |
|
| |
|
| ==Reference:== | | Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) |
| | PLANT NAME: Dresden Nuclear Power Station, Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 10/3/2013 CURRENT OPERATING CYCLE: 23 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: "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC. September 2010. |
| | Fuel Analyzed in Calculation: SVEA-960ptima2 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 2150°F Temperature (PCT) |
| | Page 1 of 2 |
| | |
| | Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) |
| | MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I 10 CFR 50.46 report dated October 30, 2009 (See Note 1) ~PCT =2°F 10 CFR 50.46 report dated October 29, 2010 (See Note 2) ~PCT = 12°F 10 CFR 50.46 report dated October 28, 2011 (See Note 3) ~PCT = 18°F 10 CFR 50.46 report dated October 26,2012 (See Note 4) ~PCT = OaF Net PCT 2182°F B. CURRENT LOCA MODEL ASSESSMENTS None (see Note 5) ~PCT = OaF Total PCT change from current assessments L~PCT =OaF r---------------------------------------------~ |
| | Cumulative PCT change from current assessments T I = OaF Net PCT Page 2 of 2 |
| | |
| | Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel) |
| | PLANT NAME: Dresden Nuclear Power Station, Unit 3 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 10/3/2013 CURRENT OPERATING CYCLE: 23 ANALYSIS OF RECORD "Westinghouse BWR ECCS Evaluation Model: |
| | Supplement 3 to Code Description, Qualification and Evaluation Model: |
| | Application to SVEA-96 Optima2 Fuel," |
| | WCAP-16078-P-A, November 2004. |
| | Calculations: "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTI MA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC. |
| | September 2010. |
| | 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) |
| | Page 1 of 2 |
| | |
| | Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel) |
| | MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated October 30, 2009 (See !'Jote 1) ~PCT = 2°F 10 CFR 50.46 report dated October 29, 2010 (See Note 2) ~PCT = 12°F 10 CFR 50.46 report dated October 28,2011 (See Note 3) ~PCT = 18°F 10 CFR 50.46 report dated October 26,2012 (See Note 4) ~PCT =OoF Net PCT 2182°F B. CURRENT LOCA MODEL ASSESSMENTS None (See !'Jote 5) ~PCT=O°F Total PCT change from current assessments L~PCT =O°F Cumulative PCT change from current assessments L I~PCT I = OaF Net PCT 2182°F Page 2 of 2 |
|
| |
|
| Letter from Timothy Hanley (SVPL TR: #09-0052, Exelon Generation Company, LLC (EGC)) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 30, 2009.] 2. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DNPS Units 2 and 3. The letter reported the replacement of core spray lower sectional piping in DNPS, Unit 2 during DNPS, Unit 2 Refueling Outage No. 21. Westinghouse evaluated the core spray leakage due to this modification and concluded that the PCT impact was O°F. The letter also identi'fied a change in input for modeling bypass hole flow coefficient in the Westinghouse LOCA analysis.
| | Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes Assessment Notes |
| The impact on PCT due to this change was determined by Westinghouse to be 12°F. Westinghouse established a MAPLHGR limit for the fresh bundles to accommodate the change. For 10 CFR 50.46 reporting purposes, the PCT impact is conservatively applied to all bundle types in the core including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for the change in bypass hole flow coefficient.
| | : 1. Prior Loss of Coolant Accident (LOCA) Model Assessment The referenced letter provided the annual 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors," report for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The letter updated the vessel leakage between the lower shroud and the downcomer. Westinghouse evaluated this change and demonstrated that all 10 CFR 50.46 criteria were satisfied. This evaluation resulted in maximum peak cladding temperature (PCT) impact due to the change in vessel leakage of 2°F for Optima2 fuel with the licensing basis PCT of 2152°F. |
| | This PCT update will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for this change. |
| [ | | [ |
|
| |
|
| ==Reference:== | | ==Reference:== |
| | | Letter from Timothy Hanley (SVPLTR: #09-0052, Exelon Generation Company, LLC (EGC)) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 30, 2009.] |
| Letter from Jeffrey Hansen (RS-1 0-191, EGG) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 29, 2010.] Page 1 of 2 Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes 3. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DI\IPS, Units 2 and 3. The letter reported errors in the current Westinghouse Dresden LOCA analysis associated with the use of incorrect R-factors. | | : 2. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DNPS Units 2 and 3. |
| The impact due to this change was determined to be 18"F in PCT update. For 10 CFR 50.46 reporting purposes, the PCT update is conservatively applied to all bundle types including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated with the correct R-factors. | | The letter reported the replacement of core spray lower sectional piping in DNPS, Unit 2 during DNPS, Unit 2 Refueling Outage No. 21. Westinghouse evaluated the core spray leakage due to this modification and concluded that the PCT impact was O°F. The letter also identi'fied a change in input for modeling bypass hole flow coefficient in the Westinghouse LOCA analysis. The impact on PCT due to this change was determined by Westinghouse to be 12°F. Westinghouse established a MAPLHGR limit for the fresh bundles to accommodate the change. For 10 CFR 50.46 reporting purposes, the PCT impact is conservatively applied to all bundle types in the core including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for the change in bypass hole flow coefficient. |
| [ | | [ |
|
| |
|
| ==Reference:== | | ==Reference:== |
| | Letter from Jeffrey Hansen (RS-1 0-191, EGG) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 29, 2010.] |
| | Page 1 of 2 |
|
| |
|
| Letter from David M. Gullott (RS-11-171, EGC) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 28,2011.]
| | Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes |
| : 4. Prior LOCA Model Assessment No new changes, error corrections, or enhancements have been report for the current DNPS LOCA analysis. | | : 3. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DI\IPS, Units 2 and 3. |
| No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record. The Analysis of Record was updated to Revision 6, but did not result in a change to the analysis results previously reported.
| | The letter reported errors in the current Westinghouse Dresden LOCA analysis associated with the use of incorrect R-factors. The impact due to this change was determined to be 18"F in PCT update. For 10 CFR 50.46 reporting purposes, the PCT update is conservatively applied to all bundle types including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated with the correct R-factors. |
| [ | | [ |
|
| |
|
| ==Reference:== | | ==Reference:== |
| | Letter from David M. Gullott (RS-11-171, EGC) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 28,2011.] |
| | : 4. Prior LOCA Model Assessment No new changes, error corrections, or enhancements have been report for the current DNPS LOCA analysis. No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record. The Analysis of Record was updated to Revision 6, but did not result in a change to the analysis results previously reported. |
| | [ |
|
| |
|
| Letter from Patrick R. Simpson (RS-12-192, EGC) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 26, 2012.] 5. Current LOCA Model Assessment 1\10 changes, error corrections, or enhancements have been reported for the current DNPS LOCA analysis. | | ==Reference:== |
| No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record. Page 2 of 2}} | | Letter from Patrick R. Simpson (RS-12-192, EGC) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 26, 2012.] |
| | : 5. Current LOCA Model Assessment 1\10 changes, error corrections, or enhancements have been reported for the current DNPS LOCA analysis. No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record. |
| | Page 2 of 2}} |
|
---|
Category:Annual Report
MONTHYEARSVPLTR 23-0014, Annual Dose Report for 20222023-04-15015 April 2023 Annual Dose Report for 2022 ML22320A5592022-11-16016 November 2022 2022 Annual Report - Guarantees of Payment of Deferred Premiums ML14321A7052014-11-14014 November 2014 2014 Annual Report - Guarantees of Payment of Deferred Premiums RS-14-296, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2014-10-24024 October 2014 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors SVPLTR 14-0028, 2013 Radioactive Effluent Release Report and Offsite Dose Calculation Manual2014-04-30030 April 2014 2013 Radioactive Effluent Release Report and Offsite Dose Calculation Manual RS-13-250, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2013-10-25025 October 2013 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors RS-12-192, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Dresden Nuclear Power Station, Units 2 and 32012-10-26026 October 2012 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Dresden Nuclear Power Station, Units 2 and 3 ML0712101332007-04-27027 April 2007 2006 Annual Radioactive Effluent Release Report ML0632404372006-11-13013 November 2006 Byron Station, Dresden, LaSalle County Station, Peach Bottom Atomic Station, Limerick Station and Quad Cities Nuclear Power Station, 2006 Annual Report - Guarantees of Payment of Deferred Premiums SVPLTR 06-0026, Annual Radiological Environmental Operating Report for 20052006-05-0505 May 2006 Annual Radiological Environmental Operating Report for 2005 SVPLTR 05-0023, Annual Radiological Environmental Operating Report for 20042005-05-13013 May 2005 Annual Radiological Environmental Operating Report for 2004 RS-05-052, Corporation Annual Financial Statements for 20042005-05-0202 May 2005 Corporation Annual Financial Statements for 2004 ML17252B1871969-12-31031 December 1969 1969 Commonwealth Edison Company Annual Report ML17252B1851965-12-31031 December 1965 1965 Annual Report for Commonwealth Edison Company 2023-04-15
[Table view] Category:Letter type:RS
MONTHYEARRS-24-073, Alternative Schedule to Complete Decommissioning Beyond 60 Years of Permanent Cessation of Operations – Additional Information Supporting Request to Withhold Security-Related Information2024-07-31031 July 2024 Alternative Schedule to Complete Decommissioning Beyond 60 Years of Permanent Cessation of Operations – Additional Information Supporting Request to Withhold Security-Related Information RS-24-063, Response to Request for Additional Information for 10 CFR 50.55a Request No. I5R-222024-06-14014 June 2024 Response to Request for Additional Information for 10 CFR 50.55a Request No. I5R-22 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-24-056, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors2024-05-28028 May 2024 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors RS-24-055, 2023 Corporate Regulatory Commitment Change Summary Report2024-05-17017 May 2024 2023 Corporate Regulatory Commitment Change Summary 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-24-021, Response to Request for Additional Information for 10 CFR 50.55a Request No. I5R-222024-04-0202 April 2024 Response to Request for Additional Information for 10 CFR 50.55a Request No. I5R-22 RS-24-002, Constellation Energy Generation, LLC - Annual Property Insurance Status Report2024-04-0101 April 2024 Constellation Energy Generation, LLC - Annual Property Insurance Status Report RS-24-023, Report on Status of Decommissioning Funding.2024-03-22022 March 2024 Report on Status of Decommissioning Funding. RS-24-016, Request for Alternative Schedule to Complete Decommissioning Beyond 60 Years of Permanent Cessation of Operations2024-03-14014 March 2024 Request for Alternative Schedule to Complete Decommissioning Beyond 60 Years of Permanent Cessation of Operations RS-24-022, Response to Request for Additional Information by the Office of Nuclear Material Safeguards and Safety to Support Review of a Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.2142024-03-0808 March 2024 Response to Request for Additional Information by the Office of Nuclear Material Safeguards and Safety to Support Review of a Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 RS-24-017, Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for Dresden Nuclear Power Station – Holtec MPC-68MCBS2024-02-23023 February 2024 Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for Dresden Nuclear Power Station – Holtec MPC-68MCBS RS-23-121, Sixth Ten-Year Interval Inservice Testing Program2023-11-30030 November 2023 Sixth Ten-Year Interval Inservice Testing Program RS-23-119, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information RS-23-107, Relief Request I5R-22, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval2023-11-0808 November 2023 Relief Request I5R-22, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval RS-23-102, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-16016 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation 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-23-080, 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-23-071, Application to Adopt TSTF-564, Safety Limit MCPR2023-08-30030 August 2023 Application to Adopt TSTF-564, Safety Limit MCPR RS-23-078, Supplement to License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram Times2023-06-16016 June 2023 Supplement to License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram Times 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-23-042, 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-23-070, Supplement to License Amendment Request Regarding Transition to GNF3 Fuel2023-05-16016 May 2023 Supplement to License Amendment Request Regarding Transition to GNF3 Fuel RS-23-057, Supplement to License Amendment Request Regarding Transition to GNF3 Fuel2023-04-17017 April 2023 Supplement to License Amendment Request Regarding Transition to GNF3 Fuel 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-23-041, Response to Request for Additional Information for Alternative Request RV-23H2023-03-14014 March 2023 Response to Request for Additional Information for Alternative Request RV-23H 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-23-003, 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-23-004, Response to Request for Additional Information Regarding Transition to GNF3 Fuel License Amendment Request2023-01-23023 January 2023 Response to Request for Additional Information Regarding Transition to GNF3 Fuel License Amendment Request 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-22-121, Notice of Intent to Pursue Subsequent License Renewal Applications2022-11-0909 November 2022 Notice of Intent to Pursue Subsequent License Renewal Applications RS-22-117, Response to Request for Additional Information License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies2022-11-0303 November 2022 Response to Request for Additional Information License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies RS-22-116, Submittal of Relief Requests Associated with the Sixth Inservice Testing Interval2022-11-0101 November 2022 Submittal of Relief Requests Associated with the Sixth Inservice Testing Interval RS-22-115, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2022-10-19019 October 2022 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors 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-22-066, License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram Times2022-08-25025 August 2022 License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram 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-22-065, License Amendment Request Regarding Transition to GNF3 Fuel2022-08-18018 August 2022 License Amendment Request Regarding Transition to GNF3 Fuel RS-22-101, 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-22-099, Submittal of Upcoming Sixth Inservice Inspection Interval Relief Requests I6R-09 and I6R-102022-07-19019 July 2022 Submittal of Upcoming Sixth Inservice Inspection Interval Relief Requests I6R-09 and I6R-10 RS-22-082, Defueled Safety Analysis Report Update, Revision 112022-06-22022 June 2022 Defueled Safety Analysis Report Update, Revision 11 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-22-064, License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.12022-06-0808 June 2022 License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.1 RS-22-051, Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists2022-04-12012 April 2022 Constellation Energy Generation, LLC - Update to Correspondence Addressees and Service Lists RS-22-049, Constellation Energy Generation, LLC, Supplemental Information to Correct Typographical Errors in Constellation'S Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for2022-04-0404 April 2022 Constellation Energy Generation, LLC, Supplemental Information to Correct Typographical Errors in Constellation'S Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for V RS-22-045, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2022-01, Preparation and Scheduling of Operator Licensing Examinations2022-03-25025 March 2022 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2022-01, Preparation and Scheduling of Operator Licensing Examinations RS-22-046, Relief Requests Associated with the Sixth Inservice Inspection Interval2022-03-25025 March 2022 Relief Requests Associated with the Sixth Inservice Inspection Interval RS-22-042, Report on Status of Decommissioning Funding for Shutdown Reactors2022-03-23023 March 2022 Report on Status of Decommissioning Funding for Shutdown Reactors RS-22-027, Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated2022-02-23023 February 2022 Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated P RS-22-023, Constellation Energy Generation, LLC, Executed Trust Fund Agreement Amendment and Subordinate Trust Agreement2022-02-23023 February 2022 Constellation Energy Generation, LLC, Executed Trust Fund Agreement Amendment and Subordinate Trust Agreement 2024-07-31
[Table view] |
Text
1
- .
tio RS-13-250 10 CFR 50.46 October 25, 2013 U. S. Nucl~ar Regulatory Commission ArrN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249
Subject:
Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Dresden Nuclear Power Station, Units 2 and 3 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Exelon Generation Company, LLC (EGC) is submitting this letter and its attachments to meet the annual reporting requirements of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for Dresden Nuclear Power Station, Units 2 and 3. This report covers the period from October 27, 2012, through October 25, 2013.
There are no regulatory commitments contained in this letter. If there are any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.
Respectfu Ily, Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments:
- 1. Dresden Nuclear Power Station Unit 2 - 10 CFR 50.46 Report (Westinghouse Fuel)
- 2. Dresden Nuclear Power Station Unit 3 - 10 CFR 50.46 Report (Westinghouse Fuel)
- 3. Dresden Nuclear Power Station Units 2 and 3 - 10 CFR 50.46 Report Assessment Notes
Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel)
PLANT NAME: Dresden Nuclear Power Station, Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 10/3/2013 CURRENT OPERATING CYCLE: 23 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: "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC. September 2010.
Fuel Analyzed in Calculation: SVEA-960ptima2 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 2150°F Temperature (PCT)
Page 1 of 2
Attachment 1 Dresden Nuclear Power Station, Unit 2 10 CFR 50.46 Report (Westinghouse Fuel)
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I 10 CFR 50.46 report dated October 30, 2009 (See Note 1) ~PCT =2°F 10 CFR 50.46 report dated October 29, 2010 (See Note 2) ~PCT = 12°F 10 CFR 50.46 report dated October 28, 2011 (See Note 3) ~PCT = 18°F 10 CFR 50.46 report dated October 26,2012 (See Note 4) ~PCT = OaF Net PCT 2182°F B. CURRENT LOCA MODEL ASSESSMENTS None (see Note 5) ~PCT = OaF Total PCT change from current assessments L~PCT =OaF r---------------------------------------------~
Cumulative PCT change from current assessments T I = OaF Net PCT Page 2 of 2
Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel)
PLANT NAME: Dresden Nuclear Power Station, Unit 3 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 10/3/2013 CURRENT OPERATING CYCLE: 23 ANALYSIS OF RECORD "Westinghouse BWR ECCS Evaluation Model:
Supplement 3 to Code Description, Qualification and Evaluation Model:
Application to SVEA-96 Optima2 Fuel,"
WCAP-16078-P-A, November 2004.
Calculations: "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTI MA2-TR021 DR-LOCA, Revision 6, Westinghouse Electric Company, LLC.
September 2010.
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)
Page 1 of 2
Attachment 2 Dresden Nuclear Power Station, Unit 3 10 CFR 50.46 Report (Westinghouse Fuel)
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated October 30, 2009 (See !'Jote 1) ~PCT = 2°F 10 CFR 50.46 report dated October 29, 2010 (See Note 2) ~PCT = 12°F 10 CFR 50.46 report dated October 28,2011 (See Note 3) ~PCT = 18°F 10 CFR 50.46 report dated October 26,2012 (See Note 4) ~PCT =OoF Net PCT 2182°F B. CURRENT LOCA MODEL ASSESSMENTS None (See !'Jote 5) ~PCT=O°F Total PCT change from current assessments L~PCT =O°F Cumulative PCT change from current assessments L I~PCT I = OaF Net PCT 2182°F Page 2 of 2
Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes Assessment Notes
- 1. Prior Loss of Coolant Accident (LOCA) Model Assessment The referenced letter provided the annual 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors," report for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The letter updated the vessel leakage between the lower shroud and the downcomer. Westinghouse evaluated this change and demonstrated that all 10 CFR 50.46 criteria were satisfied. This evaluation resulted in maximum peak cladding temperature (PCT) impact due to the change in vessel leakage of 2°F for Optima2 fuel with the licensing basis PCT of 2152°F.
This PCT update will remain in effect only until the maximum average planar linear heat generation rate (MAPLHGR) limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for this change.
[
Reference:
Letter from Timothy Hanley (SVPLTR: #09-0052, Exelon Generation Company, LLC (EGC)) to NRC, "Plant Specific ECCS Evaluation Changes -10 CFR 50.46 Report," dated October 30, 2009.]
- 2. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DNPS Units 2 and 3.
The letter reported the replacement of core spray lower sectional piping in DNPS, Unit 2 during DNPS, Unit 2 Refueling Outage No. 21. Westinghouse evaluated the core spray leakage due to this modification and concluded that the PCT impact was O°F. The letter also identi'fied a change in input for modeling bypass hole flow coefficient in the Westinghouse LOCA analysis. The impact on PCT due to this change was determined by Westinghouse to be 12°F. Westinghouse established a MAPLHGR limit for the fresh bundles to accommodate the change. For 10 CFR 50.46 reporting purposes, the PCT impact is conservatively applied to all bundle types in the core including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated for the change in bypass hole flow coefficient.
[
Reference:
Letter from Jeffrey Hansen (RS-1 0-191, EGG) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 29, 2010.]
Page 1 of 2
Attachment 3 Dresden Nuclear Power Station, Units 2 and 3 10 CFR 50.46 Report Assessment Notes
- 3. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for DI\IPS, Units 2 and 3.
The letter reported errors in the current Westinghouse Dresden LOCA analysis associated with the use of incorrect R-factors. The impact due to this change was determined to be 18"F in PCT update. For 10 CFR 50.46 reporting purposes, the PCT update is conservatively applied to all bundle types including the fresh bundles. This PCT update will remain in effect only until the MAPLHGR limits for all bundles in future DNPS, Unit 2 and Unit 3 cores are evaluated with the correct R-factors.
[
Reference:
Letter from David M. Gullott (RS-11-171, EGC) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 28,2011.]
- 4. Prior LOCA Model Assessment No new changes, error corrections, or enhancements have been report for the current DNPS LOCA analysis. No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record. The Analysis of Record was updated to Revision 6, but did not result in a change to the analysis results previously reported.
[
Reference:
Letter from Patrick R. Simpson (RS-12-192, EGC) to NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated October 26, 2012.]
- 5. Current LOCA Model Assessment 1\10 changes, error corrections, or enhancements have been reported for the current DNPS LOCA analysis. No ECCS-related changes or modifications have occurred at DNPS that affect the assumptions in the DNPS LOCA analysis of record.
Page 2 of 2