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Category:Annual Operating Report
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Cook, Units 1 and 2, Submittal of 2017 Annual Radioactive Effluent Release Report AEP-NRC-2018-35, Annual Radioactive Effluent Release Report - Off-Site Dose Calculation Manual2018-04-30030 April 2018 Annual Radioactive Effluent Release Report - Off-Site Dose Calculation Manual AEP-NRC-2017-47, Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes2017-08-30030 August 2017 Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes AEP-NRC-2016-73, Submittal of Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2016-08-26026 August 2016 Submittal of Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2013-32, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2013-08-30030 August 2013 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2012-69, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2012-08-31031 August 2012 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2012-22, Annual Radiological Environmental Operating Report for 20112012-05-0909 May 2012 Annual Radiological Environmental Operating Report for 2011 AEP-NRC-2011-32, Annual Radiological Environmental Operating Report2011-05-13013 May 2011 Annual Radiological Environmental Operating Report AEP-NRC-2011-27, Annual Radioactive Effluent Release Report, January 1, 2010, Through December 31, 20102011-04-29029 April 2011 Annual Radioactive Effluent Release Report, January 1, 2010, Through December 31, 2010 AEP-NRC-2010-37, Submittal of 2009 Annual Environmental Operating Report2010-04-30030 April 2010 Submittal of 2009 Annual Environmental Operating Report AEP-NRC-2009-54, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2009-08-28028 August 2009 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2009-29, Submittal of Annual Radioactive Effluent Release Report for 20082009-04-29029 April 2009 Submittal of Annual Radioactive Effluent Release Report for 2008 ML0913405232009-04-29029 April 2009 Enclosure to AEP-NRC-2009-32 - Donald C. Cook, Units 1 and 2 - 2008 Annual Environmental Operating Report, Cover Through Appendix Iv AEP-NRC-2009-32, Enclosure to AEP-NRC-2009-32 - Donald C. Cook, Units 1 and 2 - 2008 Annual Environmental Operating Report, Appendix V Through End2009-04-29029 April 2009 Enclosure to AEP-NRC-2009-32 - Donald C. Cook, Units 1 and 2 - 2008 Annual Environmental Operating Report, Appendix V Through End AEP-NRC-2008-27, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2008-08-29029 August 2008 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes ML0506302372005-02-25025 February 2005 2004 Annual Operating Report ML0425300392004-08-26026 August 2004 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes ML0407009102004-03-0101 March 2004 Transmittal of 2003 Annual Operating Report ML0313606492003-04-30030 April 2003 Annual Environmental Operating Report ML0312606062003-04-30030 April 2003 Annual Radiological Environmental Operating Report, Appendix D and E, Pages D-101 Through E-3 ML0312606482003-04-30030 April 2003 Annual Radiological Environmental Operating Report, Table of Contents Through Appendix D, Pages D-1 - 100 GNRO-2003/00026, South Mississippi Electric Power Association (Smep) Annual Report for 20022003-04-28028 April 2003 South Mississippi Electric Power Association (Smep) Annual Report for 2002 ML0307005302003-02-28028 February 2003 Annual Operating Report ML0214202912002-04-25025 April 2002 Part a - Donald C. Cook Nuclear Plant, Units 1 & 2 - Annual Environmental Operating Report ML0213002702002-04-25025 April 2002 Part B - Donald C. Cook Nuclear Plant, Units 1 & 2 - Annual Environmental Operating Report ML0206001322002-02-26026 February 2002 Annual Operating Report for Donald C. Cook Nuclear Plant, Units 1 & 2 for 2001 2023-08-29
[Table view] Category:Letter type:AEP
MONTHYEARAEP-NRC-2024-01, Emergency Plan Revision 482024-01-0808 January 2024 Emergency Plan Revision 48 AEP-NRC-2023-56, Report Per Technical Specification 5.6.6 for Inoperability of Unit 1 Post Accident Monitoring Reactor Coolant (Loop 3 Cold Leg) Wide Range Temperature Recorder Thermal Sensor2023-12-20020 December 2023 Report Per Technical Specification 5.6.6 for Inoperability of Unit 1 Post Accident Monitoring Reactor Coolant (Loop 3 Cold Leg) Wide Range Temperature Recorder Thermal Sensor AEP-NRC-2023-45, Unit 2 - Schedular Exemption for Enhanced Weapons, Firearms, Background Checks, and Security Event Notifications Implementation2023-11-16016 November 2023 Unit 2 - Schedular Exemption for Enhanced Weapons, Firearms, Background Checks, and Security Event Notifications Implementation AEP-NRC-2023-40, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2023-08-29029 August 2023 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2023-34, Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation2023-08-0202 August 2023 Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation AEP-NRC-2023-29, Core Operating Limits Report2023-06-19019 June 2023 Core Operating Limits Report AEP-NRC-2023-32, Response to NRC Regulatory Issue Summary 2023-1 Preparation and Scheduling of Operator Licensing Examinations2023-06-0606 June 2023 Response to NRC Regulatory Issue Summary 2023-1 Preparation and Scheduling of Operator Licensing Examinations AEP-NRC-2023-33, Renewable Operating Permit2023-06-0505 June 2023 Renewable Operating Permit AEP-NRC-2023-30, Follow-Up Notification of Ph Non-Compliance for Turbine Room Sump2023-06-0101 June 2023 Follow-Up Notification of Ph Non-Compliance for Turbine Room Sump AEP-NRC-2023-27, Annual Radiological Environmental Operating Report2023-05-15015 May 2023 Annual Radiological Environmental Operating Report AEP-NRC-2023-19, Annual Radioactive Effluent Release Report2023-04-30030 April 2023 Annual Radioactive Effluent Release Report AEP-NRC-2023-23, Annual Report of Individual Monitoring for 20222023-04-24024 April 2023 Annual Report of Individual Monitoring for 2022 AEP-NRC-2023-24, Notification of Ph Non-Compliance for Turbine Room Sump2023-04-12012 April 2023 Notification of Ph Non-Compliance for Turbine Room Sump AEP-NRC-2023-20, Annual Report of Property Insurance2023-04-0303 April 2023 Annual Report of Property Insurance AEP-NRC-2023-15, Decommissioning Funding Status Report2023-03-28028 March 2023 Decommissioning Funding Status Report AEP-NRC-2023-11, Form OAR-1, Owner'S Activity Report2023-01-31031 January 2023 Form OAR-1, Owner'S Activity Report AEP-NRC-2023-02, Request for Approval of Change Regarding Neutron Flux Instrumentation2023-01-26026 January 2023 Request for Approval of Change Regarding Neutron Flux Instrumentation AEP-NRC-2022-66, Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring2022-12-15015 December 2022 Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring AEP-NRC-2022-46, Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in2022-12-12012 December 2022 Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in AEP-NRC-2022-61, Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-062022-10-24024 October 2022 Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-06 AEP-NRC-2022-58, U1C31 Steam Generator Tube Inspection Report2022-10-24024 October 2022 U1C31 Steam Generator Tube Inspection Report AEP-NRC-2022-56, Unit 2 - Transmittal of Report of Changes to the Emergency Plan2022-10-12012 October 2022 Unit 2 - Transmittal of Report of Changes to the Emergency Plan AEP-NRC-2022-30, Core Operating Limits Report2022-10-0606 October 2022 Core Operating Limits Report AEP-NRC-2022-51, Evacuation Time Estimate Analysis2022-08-31031 August 2022 Evacuation Time Estimate Analysis AEP-NRC-2022-50, Form OAR-1, Owner'S Activity Report2022-08-25025 August 2022 Form OAR-1, Owner'S Activity Report AEP-NRC-2022-35, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes2022-08-18018 August 2022 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes AEP-NRC-2022-47, Transmittal of Donald C. 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INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant HWER One Cook Place Bridgman, M149106 A unit of American Electric Power India naMichigan Power.com August 30, 2013 AEP-NRC-2013-32 10 CFR 50.46 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 ANNUAL REPORT OF LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL CHANGES
References:
- 1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S.
Nuclear Regulatory Commission (NRC), "Response to Information Request Pursuant to 10 CFR 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation on the Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation,"
AEP-NRC-2012-13, dated March 19, 2012, (ADAMS Accession No. ML12088A104).
- 2. Letter from M. H. Carlson, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to Request for Information 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation," dated June 11, 2012, (ADAMS Accession No. ML12173A025).
- 3. Letter from P. S. Tam, NRC, to Lawrence J. Weber, I&M, "D. C. Cook Nuclear Plant (DCCNP), Units 1 and 2 - Closeout of Information Request Pursuant to 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation," dated April 2, 2012, (ADAMS Accession No. ML12088A376).
- 4. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Evaluation of Report concerning Significant Emergency Core Cooling System Evaluation Model Error Related to Nuclear Fuel Thermal Conductivity Degradation (TAC NOS. ME8322 and ME8323)," dated March 7, 2013, (ADAMS Accession No. ML13077A137).
U.S. Nuclear Regulatory Commission AEP-NRC-2013-32 Page 2 Pursuant to 10 CFR 50.46, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant (CNP), is transmitting an annual report of loss-of-coolant accident (LOCA) evaluation model changes affecting the peak cladding temperature (PCT) for CNP Unit 1 and Unit 2. CNP is providing, as an enclosure to this letter, the Unit 1 and Unit 2 Large Break and Small Break LOCA Analyses-of-Record PCT values and error assessments for calendar year 2012.
The impacts to Large Break LOCA PCT due to fuel pellet thermal conductivity degradation were previously reported to the Nuclear Regulatory Commission (NRC) in Reference 1 and supplemented by Reference 2. NRC staff review and acceptance of the impact to PCT due to fuel pellet thermal conductivity degradation was documented in Reference 4. These impacts to the LOCA analyses are discussed in the enclosure, and are included on the PCT reporting sheets.
Several other changes were made to the Large Break LOCA and Small Break LOCA evaluation models during the reporting period. The specific details of these changes were evaluated as having no impact on the calculated PCTs. Since there was no PCT impact due to these changes, they are not included in the PCT reporting sheets.
There are no new or revised commitments in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely, Joel P. Gebbie Site Vice President DMB/kmh
Enclosure:
Donald C. Cook Nuclear Plant Units 1 and 2, Large and Small Break Loss-of-Coolant Accident Peak Clad Temperature Summary c: J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosures MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III T. J. Wengert, NRC Washington, DC
ENCLOSURE TO AEP-NRC-2013-32 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE
SUMMARY
Abbreviations:
CNP Donald C. Cook Nuclear Plant OF degrees Fahrenheit ECCS emergency core cooling system EM evaluation methodology FdH nuclear enthalpy rise hot channel factor FQ heat flux hot channel factor HHSI high head safety injection (Safety Injection System at CNP)
I&M Indiana Michigan Power Company LOCA loss of coolant accident MWt megawatts - thermal NRC Nuclear Regulatory Commission PCT peak cladding temperature RHR Residual Heat Removal SGTP steam generator tube plugging TCD thermal conductivity degradation Summary:
By letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), I&M, the licensee for CNP Units 1 and 2, submitted a report describing the impact of fuel pellet TCD on the Large Break LOCA ECCS evaluation model, and an estimate of the effect on the predicted PCT for CNP Units 1 and 2. This report was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.46, Paragraph (a)(3), and referred to a letter from Westinghouse Electric Company dated March 7, 2012 (ADAMS Accession No. ML12072A035).
The report was subsequently found to be acceptable by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13077A137).
A new small break LOCA analysis of record was implemented for CNP Unit 1 in 2012. The revised small break LOCA was provided to the NRC by letter dated August 31, 2012 (ADAMS Accession No. ML12256A685). The enclosure to the letter provides a report of the revised analysis. The results of the revised small break LOCA analysis conform to the emergency core cooling system acceptance criteria of 10 CFR 50.46. The calculated peak cladding temperature of 1725 0 F is below the 10 CFR 50.46 limit of 22001F.
The following pages summarize the impact of TCD, peaking factor burndown, and plant modification evaluations on the CNP Units 1 and 2 Large Break LOCA analyses of record. In addition, pages are included that summarize the small break LOCA PCT analyses of record for CNP Units 1 and 2.
Enclosure to AEP-NRC-2013-32 Page 2 CNP UNIT 1 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Evaluation Model: ASTRUM (2004)
FQ= 2.15 FdH = 1.55 SGTP = 10%(a) Break Size: Split Analysis Date: November 20, 2007 LICENSING BASIS Analysis-of-Record PCT = 2128°F MARGIN ALLOCATIONS (Delta PCT)
A. PREVIOUS 10 CFR 50.46 ASSESSMENTS
- 1. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS -381 °F(a)
C. NEW 10 CFR 50.46 ASSESSMENTS 384 0 F(a)
D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2131°F Notes:
- a. These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown and design input changes (e.g., reduction in the maximum allowed steam generator tube plugging from 10% to 2%
and maximum FdH reduced to 1.545). Evaluation details provided in a letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), and subsequently found acceptable by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13077A137).
Enclosure to AEP-NRC-2013-32 Page 3 CNP UNIT 1 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP FQ=2.32 FdH=1.55 SGTP=10% 3.25 inch cold leg break Analysis Date: January 6, 2012 Notes: 3304 MWt (plus 0.34% calorimetric uncertainty)
LICENSING BASIS Analysis-of-Record PCT = 1725°F MARGIN ALLOCATIONS (Delta PCT)
A. PREVIOUS 10 CFR 50.46 ASSESSMENTS
- 1. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS 0OF C. NEW 10 CFR 50.46 ASSESSMENTS 0OF D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1725°F
Enclosure to AEP-NRC-2013-32 Page 4 CNP UNIT 2 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Evaluation Model: ASTRUM (2004)
FQ = 2.335 FdH = 1.644 SGTP = 10%(a) Break Size: Split Analysis Date: February 9, 2009 LICENSING BASIS Analysis-of-Record PCT = 2107°F MARGIN ALLOCATIONS (Delta PCT)
A. PREVIOUS 10 CFR 50.46 ASSESSMENTS
- 1. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS -2390 F(a)
C. NEW 10 CFR 50.46 ASSESSMENTS 73 0 F(a)
D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 19410 F Notes:
- a. These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown and design input changes (e.g., reduction in the maximum allowed steam generator tube plugging from 10% to 1%
and maximum FdH reduced to 1.61). Evaluation details provided in a letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), and subsequently found acceptable by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13077A137).
Enclosure to AEP-NRC-2013-32 Page 5 CNP UNIT 2 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP FQ= 2.32 FdH= 1.62 SGTP = 10% 4 inch cold leg break Analysis Date: April 25, 2011 Notes: 3600 MWt LICENSING BASIS Analysis-of-Record PCT = 1274°F (a)
MARGIN ALLOCATIONS (Delta PCT)
A. PREVIOUS 10 CFR 50.46 ASSESSMENTS
- 1. None 0°F B. PLANNED PLANT MODIFICATION EVALUATIONS 0OF C. NEW 10 CFR 50.46 ASSESSMENTS 0OF D. OTHER 0°F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1274°F Notes:
- a. The 3600 MWt power level used in this analysis bounds the Unit 2 3468 MWt steady state power limit in the operating license.