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| {{#Wiki_filter:February 28, 2006Mr. Mano K. NazarSenior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 | | {{#Wiki_filter:February 28, 2006 Mr. Mano K. Nazar Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806) | | DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806) |
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| ==Dear Mr. Nazar:== | | ==Dear Mr. Nazar:== |
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| On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to RenewedFacility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. Theamendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, "Reactor Trip System Instrumentation" of the units' Technical Specifications.Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting theamendments references General Design Criterion 10, despite the fact that both units wereconstructed before Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error. | | On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293 to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to Renewed Facility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation of the units' Technical Specifications. |
| | Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting the amendments references General Design Criterion 10, despite the fact that both units were constructed before Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error. |
| Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you. | | Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you. |
| | Sincerely, |
| | /RA/ |
| | Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316 |
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| Sincerely,/RA/Peter S. Tam, Senior Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-315 and 50-316
| | ==Enclosures:== |
| | As stated cc w/encls: See next page |
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| ==Enclosures:==
| | February 28, 2006 Mr. Mano K. Nazar Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 |
| As stated cc w/encls: See next page February 28, 2006Mr. Mano K. NazarSenior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
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| ==SUBJECT:== | | ==SUBJECT:== |
| DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806) | | DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806) |
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| ==Dear Mr. Nazar:== | | ==Dear Mr. Nazar:== |
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| On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to RenewedFacility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. Theamendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, "Reactor Trip System Instrumentation" of the units' Technical Specifications.Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting theamendments references General Design Criterion 10, despite the fact that both units wereconstructed before Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error. | | On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293 to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to Renewed Facility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation of the units' Technical Specifications. |
| | Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting the amendments references General Design Criterion 10, despite the fact that both units were constructed before Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error. |
| Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you. | | Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you. |
| | Sincerely, |
| | /RA/ |
| | Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316 |
| | |
| | ==Enclosures:== |
| | As stated cc w/encls: See next page DISTRIBUTION PUBLIC LPLIII-1 R/F RidsNrrPMPTam RidsNrrLATHarris RidsOgcRp GHill(2) RidsAcrsAcnwMailCenter RidsNrrDorlLple RidsRgn3MailCenter RidsNrrDirsItsb RidsNrrDorl SSun ADAMS Accession Number: ML060460417 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/SPWB NRR/SPWB/BC NRR/LPL3-1/BC(A) |
| | NAME PTam THarris SSun JNakoski TKobetz DATE 2/28/06 2/24/06 2/28/06 2/28/06 2/28/06 OFFICIAL RECORD COPY |
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| Sincerely,/RA/
| | SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316 |
| Peter S. Tam, Senior Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-315 and 50-316
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| ==Enclosures:== | | ==1.0 INTRODUCTION== |
| As stated cc w/encls: See next pageDISTRIBUTIONPUBLICLPLIII-1 R/F RidsNrrPMPTamRidsNrrLATHarrisRidsOgcRpGHill(2)RidsAcrsAcnwMailCenterRidsNrrDorlLple RidsRgn3MailCenterRidsNrrDirsItsbRidsNrrDorlSSunADAMS Accession Number: ML060460417 OFFICENRR/LPL3-1/PMNRR/LPL3-1/LANRR/SPWBNRR/SPWB/BCNRR/LPL3-1/BC(A)NAMEPTam THarrisSSunJNakoskiTKobetzDATE2/28/062/24/062/28/062/28/062/28/06OFFICIAL RECORD COPY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 293 TORENEWED FACILITY OPERATING LICENSE NO. DPR-58AND AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74INDIANA MICHIGAN POWER COMPANYDONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2DOCKET NOS. 50-315 AND 50-31
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| ==61.0INTRODUCTION==
| | By application dated August 10, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052300238), Indiana Michigan Power Company (the licensee) requested amendments to the Technical Specifications (TSs) for the Donald C. Cook (D. C. Cook) Nuclear Plant, Units 1 and 2. The proposed amendments would delete the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation. |
| By application dated August 10, 2005 (Agencywide Documents Access and ManagementSystem (ADAMS) Accession No. ML052300238), Indiana Michigan Power Company (the licensee) requested amendments to the Technical Specifications (TSs) for the Donald C. Cook (D. C. Cook) Nuclear Plant, Units 1 and 2. The proposed amendments would delete the power range neutron flux high negative rate trip function from Table 3.3.1-1, "Reactor Trip System Instrumentation." The licensee stated that the proposed change would allow elimination of an unnecessary tripfunction and thereby reduce the potential for a transient due to reactor shutdown in meeting TSrequirements for a limiting condition for operation (LCO) of the trip function. The Nuclear Regulatory Commission (NRC) staff's review of the application is set forth below. | | The licensee stated that the proposed change would allow elimination of an unnecessary trip function and thereby reduce the potential for a transient due to reactor shutdown in meeting TS requirements for a limiting condition for operation (LCO) of the trip function. The Nuclear Regulatory Commission (NRC) staffs review of the application is set forth below. |
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| ==2.0REGULATORY EVALUATION== | | ==2.0 REGULATORY EVALUATION== |
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| lThe NRC's regulatory requirements related to the content of TSs are set forth in l10 CFR 50.36, "Technical Specifications." Specifically, 10 CFR 50.36(c)(2)(ii) specifies four screening criteria to be used in determining whether an LCO is required to be included in the TSs:Revised by letter dated 2/28/06l Criterion 1 - Installed instrumentation that is used to detect, and indicate in the controlroom, a significant abnormal degradation of the reactor coolant pressure boundary (RCPB); Criterion 2 - A process variable, design feature, or operating restriction that is an initialcondition of a design-basis accident (DBA) or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier;Criterion 3 - A structure, system or component (SSC) that is part of the primary successpath, and which functions or actuates to mitigate a DBA or transient that either assumesthe failure of, or presents a challenge to, the integrity of a fission product barrier; and Criterion 4 - An SSC which operating experience or probabilistic risk assessment hasshown to be significant to public health and safety. The licensee proposed to delete the power range neutron flux high negative rate trip functionfrom the D. C. Cook TSs. The NRC staff evaluation of the licensee's proposed change waspredicated upon continued compliance with the screening criteria specified in 10 CFR 50.36.l
| | l l |
| | l l |
| | The NRCs regulatory requirements related to the content of TSs are set forth in l 10 CFR 50.36, Technical Specifications. Specifically, 10 CFR 50.36(c)(2)(ii) specifies four screening criteria to be used in determining whether an LCO is required to be included in the TSs: |
| | Revised by letter dated 2/28/06 l |
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| | Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary (RCPB); |
| | Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; Criterion 3 - A structure, system or component (SSC) that is part of the primary success path, and which functions or actuates to mitigate a DBA or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; and Criterion 4 - An SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. |
| | The licensee proposed to delete the power range neutron flux high negative rate trip function from the D. C. Cook TSs. The NRC staff evaluation of the licensees proposed change was predicated upon continued compliance with the screening criteria specified in 10 CFR 50.36. l |
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| ==3.0TECHNICAL EVALUATION== | | ==3.0 TECHNICAL EVALUATION== |
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| The power range neutron flux high negative rate trip function was designed as part of thereactor protecti on system (RPS) to mitigate the consequences of one or more dropped rodcluster control assemblies (RCCAs) event. The dr opped RCCAs event is an anticipatedoperational occurrence, and is caused by a single electric or mechanical failure that results in a number and combination of RCCAs from the same group of a given bank to drop to the bottomof the core. The resulting negative reactivity insertion causes nuclear power to quickly decrease and core radial peaking factors to increase. The reduced power and continued steam generation cause the reactor coolant temperature to decrease. In the manual control mode, thepositive reactivity feedback due to dropping temperature causes the reactor power to rise to initial power level at a reduced reactor vessel inlet temperature with no power overshoot. In the automatic control mode, the plant control system detects the reduction in core power andinitiates control bank withdrawal in order to restore core power. As a result, power overshoot occurs, resulting in lower calculated departure from nucleate boiling ratios (DNBRs). At higherpower levels, in the event of a dropped RCCA event, the RPS will detect the rapidly decreasingneutron flux due to the dropped RCCAs and trip the reactor based on the power range neutronflux high negative rate trip function, thus ending the transient and assuring that DNBR designlimits are maintained. Since the dropped RCCA event is an anticipated operational occurrence,it must be shown that the DNBR design limits are met for the combination of high nuclearl power, high radial peaking factor, and other system conditions that exist following the droppedRCCA event.In a topical report WCAP-10297-P-A, "Dropped Rod Methodology for Negative Flux Rate TripPlants" (see ADAMS Accession No. 8304140120) Westinghouse documented a methodology for the analysis of the dropped RCCA event and concluded that the high negative flux rate tripwas required only when a dropped RCCA (or RCCA bank) exceeded the threshold value ofreactivity worth. Any dropped RCCA having a worth below the threshold value would notRevised by letter dated 2/28/06l Donald C. Cook Nuclear Plant, Units 1 and 2 cc:Regional Administrator, Region IIIU.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville RoadLisle, IL 60532-4351Attorney GeneralDepartment of Attorney General 525 West Ottawa Street Lansing, MI 48913Township SupervisorLake Township Hall | | The power range neutron flux high negative rate trip function was designed as part of the reactor protection system (RPS) to mitigate the consequences of one or more dropped rod cluster control assemblies (RCCAs) event. The dropped RCCAs event is an anticipated operational occurrence, and is caused by a single electric or mechanical failure that results in a number and combination of RCCAs from the same group of a given bank to drop to the bottom of the core. The resulting negative reactivity insertion causes nuclear power to quickly decrease and core radial peaking factors to increase. The reduced power and continued steam generation cause the reactor coolant temperature to decrease. In the manual control mode, the positive reactivity feedback due to dropping temperature causes the reactor power to rise to initial power level at a reduced reactor vessel inlet temperature with no power overshoot. In the automatic control mode, the plant control system detects the reduction in core power and initiates control bank withdrawal in order to restore core power. As a result, power overshoot occurs, resulting in lower calculated departure from nucleate boiling ratios (DNBRs). At higher power levels, in the event of a dropped RCCA event, the RPS will detect the rapidly decreasing neutron flux due to the dropped RCCAs and trip the reactor based on the power range neutron flux high negative rate trip function, thus ending the transient and assuring that DNBR design limits are maintained. Since the dropped RCCA event is an anticipated operational occurrence, it must be shown that the DNBR design limits are met for the combination of high nuclear l power, high radial peaking factor, and other system conditions that exist following the dropped RCCA event. |
| | In a topical report WCAP-10297-P-A, Dropped Rod Methodology for Negative Flux Rate Trip Plants (see ADAMS Accession No. 8304140120) Westinghouse documented a methodology for the analysis of the dropped RCCA event and concluded that the high negative flux rate trip was required only when a dropped RCCA (or RCCA bank) exceeded the threshold value of reactivity worth. Any dropped RCCA having a worth below the threshold value would not Revised by letter dated 2/28/06 l |
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| P.O. Box 818 Bridgman, MI 49106U.S. Nuclear Regulatory CommissionResident Inspector's Office 7700 Red Arrow Highway Stevensville, MI 49127James M. Petro, Jr., EsquireIndiana Michigan Power Company One Cook Place Bridgman, MI 49106Mayor, City of Bridgman P.O. Box 366 Bridgman, MI 49106Special Assistant to the GovernorRoom 1 - State Capitol Lansing, MI 48909Mr. John A. ZwolinskiSafety Assurance Director Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106Michigan Department of Environmental Quality Waste and Hazardous Materials Div.
| | Donald C. Cook Nuclear Plant, Units 1 and 2 cc: |
| Hazardous Waste & Radiological Protection Section Nuclear Facilities Unit Constitution Hall, Lower-Level North 525 West Allegan Street P. O. Box 30241 Lansing, MI 48909-7741Lawrence J. Weber, Plant ManagerIndiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106Mr. Joseph N. Jensen, Site Vice President Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106}} | | Regional Administrator, Region III Michigan Department of Environmental U.S. Nuclear Regulatory Commission Quality Suite 210 Waste and Hazardous Materials Div. |
| | 2443 Warrenville Road Hazardous Waste & Radiological Lisle, IL 60532-4351 Protection Section Nuclear Facilities Unit Attorney General Constitution Hall, Lower-Level North Department of Attorney General 525 West Allegan Street 525 West Ottawa Street P. O. Box 30241 Lansing, MI 48913 Lansing, MI 48909-7741 Township Supervisor Lawrence J. Weber, Plant Manager Lake Township Hall Indiana Michigan Power Company P.O. Box 818 Nuclear Generation Group Bridgman, MI 49106 One Cook Place Bridgman, MI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mr. Joseph N. Jensen, Site Vice President 7700 Red Arrow Highway Indiana Michigan Power Company Stevensville, MI 49127 Nuclear Generation Group One Cook Place James M. Petro, Jr., Esquire Bridgman, MI 49106 Indiana Michigan Power Company One Cook Place Bridgman, MI 49106 Mayor, City of Bridgman P.O. Box 366 Bridgman, MI 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, MI 48909 Mr. John A. Zwolinski Safety Assurance Director Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106}} |
Letter Sequence Other |
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MONTHYEARML0523002382005-08-10010 August 2005
Deletion of the Power Range Neutron Flux High Negative Rate Trip Function
Project stage: Request
ML0602304522006-02-10010 February 2006
License Amendment, Issuance of Amendments to Delete the Power Range Neutron Flux High Negative Rate Trip Function
Project stage: Approval
ML0604400222006-02-10010 February 2006
Technical Specifications, Delete the Power Range Neutron Flux High Negative Rate Trip Function
Project stage: Other
ML0604604172006-02-28028 February 2006
Correction to Amendments Deletion of the Power Range Neutron Flux High Negative Rate Function
Project stage: Other
2006-02-10
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Category:Safety Evaluation
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Relief Request ISIR-5-06 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles
ML22214A0012022-10-0707 October 2022
Issuance of Amendment Nos. 361 and 343 Regarding Adoption of TSTF-554, Revise Reactor Coolant Leakage Requirements
ML22166A3302022-07-26026 July 2022
Review of Quality Assurance Program Changes
ML22046A2332022-06-21021 June 2022
Issuance of Amendment Nos. 360 and 342 Regarding Change to the Technical Specification Requirement for Containment Water Level Instrumentation
ML22055A0012022-06-0808 June 2022
Issuance of Amendment No. 341 Updating the Reactor Coolant System Pressure Temperature Limits
ML22102A0122022-05-0202 May 2022
Issuance of Amendment Nos. 359 and 340 Regarding Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
ML21295A0092021-11-0303 November 2021
Relief Request ISIR-4-11 Limited Coverage Examinations During the Fourth 10 Year Inservice Inspection Interval
ML21141A2612021-06-0202 June 2021
Alternative Request REL-PP2 Related to Fifth 10-Year Inservice Testing Program Interval
ML21130A0082021-05-12012 May 2021
Relief Request ISIR-5-05 Related to ASME Code Case N 729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles (EPID L-2021-LLR-0033 (COVID-19))
ML21062A1882021-03-23023 March 2021
Issuance of Amendment No. 339 One Cycle Extension of Appendix J, Type a, Integrated Leakage Rate Test Interval (EPID L-2020-LLA-0280 (COVID-19))
ML21041A0862021-03-0303 March 2021
Issuance of Amendment No. 338 Regarding One-Time Deferral of the Steam Generator Tube Inspections
ML21034A1552021-02-12012 February 2021
Relief Request ISIR-5-04 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles
ML21006A4582021-02-0202 February 2021
Issuance of Amendments Nos. 358 and 337 Regarding Revision to Technical Specifications to Adopt Technical Specifications Task Force Traveler 541, Revision 2
ML20366A1552021-01-15015 January 2021
Issuance of Amendments Nos. 357 and 336 Regarding Revision to Technical Specifications Bases Control Program
ML20366A1342021-01-13013 January 2021
Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques
ML20329A0012021-01-12012 January 2021
Issuance of Amendment No. 356 Regarding Updating the Reactor Coolant System Pressure-Temperature Limits
ML20322A4282021-01-0606 January 2021
Issuance of Amendment Nos. 355 and 335, Revision to Technical Specifications to Adopt Technical Specification Task Force Traveler 567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues
ML20315A4832020-12-30030 December 2020
Issuance of Amendment Nos. 354 and 334 Adopt Technical Specification Task Force Traveler TSTF-412, Revision 3, Provide Actions for One Steam Supply to the Turbine Driven Afw/Efw Pump Inoperable
ML20247J6562020-09-10010 September 2020
Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI
ML20213C7042020-09-0303 September 2020
Issuance of Amendment No. 353 One Cycle Extension of Appendix J, Type a, Integrated Leakage Rate Test Interval
ML19347B3762020-01-31031 January 2020
Issuance of Amendment Nos. 350 and 331, Revise Technical Specification 5.5.5, Reactor Coolant Pump Flywheel Inspection Program, in Accordance with Technical Specification Task Force TSTF-421
ML19329A0112020-01-23023 January 2020
Issuance of Amendment Numbers 349 and 330 to Apply Leak Before-Break Methodology to Reactor Coolant System Branch Lines and Deletion of Containment Humidity Monitor
ML19304B6722019-12-31031 December 2019
Issuance of Amendment Numbers 348 and 329 to Revise Operating Licenses DPR-58 and DPR-74, to Address Issues Identified in Westinghouse Document NSAL-15-1
ML19259A0542019-10-15015 October 2019
Issuance of Amendment to Revise Operating Licenses DPR-58 and DPR-74, Appendix B, Environmental Technical Specifications, Part II, Non-Radiological Environment Protection Plan
ML19196A0642019-08-23023 August 2019
Relied Request ISIR-4-10 Regarding Fourth Inservice Inspection Program Interval
ML19170A3622019-08-0101 August 2019
Issuance of Amendment Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping
ML19134A3552019-07-11011 July 2019
Issuance of Amendments Technical Specification Task Force (TSTF) Traveler 563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program
ML19031B9662019-04-10010 April 2019
Issuance of Amendments 344, 326 Regarding Request to Adopt TSTF-529, Revision 4, Clarify Use and Application Rules
ML18346A3582019-02-0505 February 2019
Issuance of Amendments 343 and 325 Regarding the Battery Monitoring and Maintenance Program
ML18348A4182019-01-0909 January 2019
Units and 2 Approval of Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N 729-4
ML18337A3942018-12-11011 December 2018
Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination
ML18284A2542018-11-16016 November 2018
Issuance of Amendments Request for Deviation from National Fire Protection Association 805 Requirements
ML18249A0192018-11-13013 November 2018
Issuance of Amendment Nos. 341 and 323 Technical Support Center Relocation
ML18284A3102018-10-26026 October 2018
Approval of Alternative to the ASME Code Regarding Reactor Vessel Weld Examination - Relief Request ISIR-4-08
ML18131A2532018-07-0606 July 2018
Issuance of Amendments Request for Deviation from National Fire Protection Association 805 Requirements
ML18103A0592018-04-19019 April 2018
Request for Use of Alternative Isir 04-05, Revision 1, Associated with Reactor Vessel Closure Head Volumetric/Surface Examination Frequency Requirements for the Inservice Inspection Program
ML17312B0302017-12-19019 December 2017
Issuance of Amendments License Amendment Request to Revise Technical Specifications Section 3.7.2, Steam Generator Stop Valves (CAC Nos. MF9539 and MF9540; EPID L-2017-LLA-0198)
ML17214A5502017-09-21021 September 2017
Issuance of Amendments License Amendment Request Regarding Technical Specification 3.9.3, Containment Penetrations
ML17131A2772017-06-0707 June 2017
Issuance of Amendments License Amendment Request Regarding Containment Leakage Rate Testing Program
ML17103A1062017-05-24024 May 2017
Issuance of Amendments Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing
ML17096A6272017-04-12012 April 2017
Proposed Alternative to Use ASME OM Code Case OMN-20
ML17045A1502017-03-31031 March 2017
Issuance of Amendments Adopting of TSTF0425-A, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B
ML16327A1102016-12-28028 December 2016
Cover Letter for Revised Safety Evaluation for Amendment Nos. 332 and 314 Adoption of TSTF-490, Rev. 0, and Implementation of Full-Scope Alternative Source Term
ML16242A1112016-10-20020 October 2016
DC Cook, Units 1 and 2 - Issuance of Amendments Adoption of TSTF-490,REV.0,Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification and Implementation of Full-Scope Alternative Source Ter
ML16216A1812016-08-19019 August 2016
Issuance of Amendment to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System Instrumentation
ML16195A0042016-08-0404 August 2016
Issuance of Amendments to Revise Technical Specifications to Adopt Technical Specifications Task Force - 523, Generic Letter 2008 01, Managing Gas Accumulation
ML16032A0312016-02-0404 February 2016
Request for Use of Alternative REL-PP1 Associated with Pump Inservice Testing (CAC Nos. MF6548 and MF6549)
ML15327A2172015-12-11011 December 2015
Issuance of Amendments Technical Specifications Surveillance Requirements 3.8.1.10, 3.8.1.11, and 3.8.1.15
ML14197A0972015-11-30030 November 2015
Issuance of Amendment Regarding Restoration of Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions
ML15264A8512015-11-0909 November 2015
Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051
2023-01-04
[Table view] |
Text
February 28, 2006 Mr. Mano K. Nazar Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806)
Dear Mr. Nazar:
On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293 to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to Renewed Facility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation of the units' Technical Specifications.
Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting the amendments references General Design Criterion 10, despite the fact that both units were constructed before Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error.
Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
As stated cc w/encls: See next page
February 28, 2006 Mr. Mano K. Nazar Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTION TO AMENDMENTS RE: DELETION OF THE POWER RANGE NEUTRON FLUX HIGH NEGATIVE RATE TRIP FUNCTION (TAC NOS. MC8805 AND MC8806)
Dear Mr. Nazar:
On February 10, 2006, the Nuclear Regulatory Commission (NRC) issued Amendment No. 293 to Renewed Facility Operating License No. DPR-58 and Amendment No. 275 to Renewed Facility Operating License No. DPR-74 for Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments deleted the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation of the units' Technical Specifications.
Subsequent to issuance, the NRC staff noted that the safety evaluation (SE) supporting the amendments references General Design Criterion 10, despite the fact that both units were constructed before Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 was promulgated. Accordingly, referencing GDC 10 in the SE is an error.
Enclosed please find corrected pages 1 and 2 of the SE, deleting all references to GDC 10, with marginal lines indicating the areas corrected. Please use these pages to replace corresponding pages in the SE. The corrections on pages 1 and 2 of the SE do not change the NRC staff's conclusion in the SE. The NRC staff regrets any inconvenience this error may have caused you.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
As stated cc w/encls: See next page DISTRIBUTION PUBLIC LPLIII-1 R/F RidsNrrPMPTam RidsNrrLATHarris RidsOgcRp GHill(2) RidsAcrsAcnwMailCenter RidsNrrDorlLple RidsRgn3MailCenter RidsNrrDirsItsb RidsNrrDorl SSun ADAMS Accession Number: ML060460417 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/SPWB NRR/SPWB/BC NRR/LPL3-1/BC(A)
NAME PTam THarris SSun JNakoski TKobetz DATE 2/28/06 2/24/06 2/28/06 2/28/06 2/28/06 OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
By application dated August 10, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052300238), Indiana Michigan Power Company (the licensee) requested amendments to the Technical Specifications (TSs) for the Donald C. Cook (D. C. Cook) Nuclear Plant, Units 1 and 2. The proposed amendments would delete the power range neutron flux high negative rate trip function from Table 3.3.1-1, Reactor Trip System Instrumentation.
The licensee stated that the proposed change would allow elimination of an unnecessary trip function and thereby reduce the potential for a transient due to reactor shutdown in meeting TS requirements for a limiting condition for operation (LCO) of the trip function. The Nuclear Regulatory Commission (NRC) staffs review of the application is set forth below.
2.0 REGULATORY EVALUATION
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The NRCs regulatory requirements related to the content of TSs are set forth in l 10 CFR 50.36, Technical Specifications. Specifically, 10 CFR 50.36(c)(2)(ii) specifies four screening criteria to be used in determining whether an LCO is required to be included in the TSs:
Revised by letter dated 2/28/06 l
Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary (RCPB);
Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; Criterion 3 - A structure, system or component (SSC) that is part of the primary success path, and which functions or actuates to mitigate a DBA or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; and Criterion 4 - An SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The licensee proposed to delete the power range neutron flux high negative rate trip function from the D. C. Cook TSs. The NRC staff evaluation of the licensees proposed change was predicated upon continued compliance with the screening criteria specified in 10 CFR 50.36. l
3.0 TECHNICAL EVALUATION
The power range neutron flux high negative rate trip function was designed as part of the reactor protection system (RPS) to mitigate the consequences of one or more dropped rod cluster control assemblies (RCCAs) event. The dropped RCCAs event is an anticipated operational occurrence, and is caused by a single electric or mechanical failure that results in a number and combination of RCCAs from the same group of a given bank to drop to the bottom of the core. The resulting negative reactivity insertion causes nuclear power to quickly decrease and core radial peaking factors to increase. The reduced power and continued steam generation cause the reactor coolant temperature to decrease. In the manual control mode, the positive reactivity feedback due to dropping temperature causes the reactor power to rise to initial power level at a reduced reactor vessel inlet temperature with no power overshoot. In the automatic control mode, the plant control system detects the reduction in core power and initiates control bank withdrawal in order to restore core power. As a result, power overshoot occurs, resulting in lower calculated departure from nucleate boiling ratios (DNBRs). At higher power levels, in the event of a dropped RCCA event, the RPS will detect the rapidly decreasing neutron flux due to the dropped RCCAs and trip the reactor based on the power range neutron flux high negative rate trip function, thus ending the transient and assuring that DNBR design limits are maintained. Since the dropped RCCA event is an anticipated operational occurrence, it must be shown that the DNBR design limits are met for the combination of high nuclear l power, high radial peaking factor, and other system conditions that exist following the dropped RCCA event.
In a topical report WCAP-10297-P-A, Dropped Rod Methodology for Negative Flux Rate Trip Plants (see ADAMS Accession No. 8304140120) Westinghouse documented a methodology for the analysis of the dropped RCCA event and concluded that the high negative flux rate trip was required only when a dropped RCCA (or RCCA bank) exceeded the threshold value of reactivity worth. Any dropped RCCA having a worth below the threshold value would not Revised by letter dated 2/28/06 l
Donald C. Cook Nuclear Plant, Units 1 and 2 cc:
Regional Administrator, Region III Michigan Department of Environmental U.S. Nuclear Regulatory Commission Quality Suite 210 Waste and Hazardous Materials Div.
2443 Warrenville Road Hazardous Waste & Radiological Lisle, IL 60532-4351 Protection Section Nuclear Facilities Unit Attorney General Constitution Hall, Lower-Level North Department of Attorney General 525 West Allegan Street 525 West Ottawa Street P. O. Box 30241 Lansing, MI 48913 Lansing, MI 48909-7741 Township Supervisor Lawrence J. Weber, Plant Manager Lake Township Hall Indiana Michigan Power Company P.O. Box 818 Nuclear Generation Group Bridgman, MI 49106 One Cook Place Bridgman, MI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mr. Joseph N. Jensen, Site Vice President 7700 Red Arrow Highway Indiana Michigan Power Company Stevensville, MI 49127 Nuclear Generation Group One Cook Place James M. Petro, Jr., Esquire Bridgman, MI 49106 Indiana Michigan Power Company One Cook Place Bridgman, MI 49106 Mayor, City of Bridgman P.O. Box 366 Bridgman, MI 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, MI 48909 Mr. John A. Zwolinski Safety Assurance Director Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
