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Category:Letter
MONTHYEARAEP-NRC-2024-62, Unit 2 Independent Spent Fuel Storage Installation - Registration of Dry Spent Fuel Storage Cask2024-08-0808 August 2024 Unit 2 Independent Spent Fuel Storage Installation - Registration of Dry Spent Fuel Storage Cask AEP-NRC-2024-47, Form OAR-1, Owners Activity Report2024-07-30030 July 2024 Form OAR-1, Owners Activity Report ML24183A0162024-07-25025 July 2024 Review of Reactor Vessel Material Surveillance Program Capsule W Technical Report ML24169A2142024-07-25025 July 2024 Issuance of Amendment No. 362 Regarding Change to Technical Specification 3.4.12, Low Temperature Overpressure Protection System IR 05000315/20240022024-07-24024 July 2024 Integrated Inspection Report 05000315/2024002 and 05000316/2024002 ML24197A1262024-07-15015 July 2024 Unit 2 - Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating ML24191A0692024-07-0909 July 2024 Operator Licensing Examination Approval - Donald C. Cook Nuclear Power Plant, July 2024 AEP-NRC-2024-56, Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask2024-07-0808 July 2024 Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask AEP-NRC-2024-48, Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-07-0202 July 2024 Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating ML24176A1012024-06-21021 June 2024 57143-EN 57143 - Paragon Energy Solutions - Update 1 (Final) - 10CFR Part 21 Final Notification: P21-05242024-FN, Rev. 0 AEP-NRC-2024-45, Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring2024-06-13013 June 2024 Report Per Technical Specification 5.6.6 for Inoperability of Post Accident Monitoring Neutron Flux Monitoring ML24163A0132024-06-12012 June 2024 Request for Information for the NRC Age-Related Degradation Inspection: Inspection Report 05000315/2024012 and 05000316/2024012 AEP-NRC-2024-23, Core Operating Limits Report2024-05-23023 May 2024 Core Operating Limits Report ML24141A2162024-05-20020 May 2024 —Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection AEP-NRC-2024-40, Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-05-16016 May 2024 Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating AEP-NRC-2024-41, Annual Radiological Environmental Operating Report2024-05-15015 May 2024 Annual Radiological Environmental Operating Report IR 05000315/20240012024-05-14014 May 2024 Integrated Inspection Report 05000315/2024001 and 05000316/2024001 IR 05000315/20244012024-05-14014 May 2024 – Security Baseline Inspection Report 05000315/2024401 and 05000316/2024401 AEP-NRC-2024-07, Unit 2 - Transmittal of Report of Changes to the Emergency Plan2024-05-14014 May 2024 Unit 2 - Transmittal of Report of Changes to the Emergency Plan AEP-NRC-2024-26, Transmittal of Donald C. Cook Nuclear Plant, Emergency Plan Revision 492024-05-14014 May 2024 Transmittal of Donald C. Cook Nuclear Plant, Emergency Plan Revision 49 ML24115A2152024-05-0707 May 2024 LTR: CNP Non-Acceptance with Opportunity TS 3-8-1 AEP-NRC-2024-24, Form OAR-1, Owners Activity Report2024-05-0707 May 2024 Form OAR-1, Owners Activity Report ML24116A0002024-05-0202 May 2024 – Regulatory Audit in Support of Review of the Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors AEP-NRC-2024-35, Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations2024-04-30030 April 2024 Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations AEP-NRC-2024-28, 2023 Annual Radioactive Effluent Release Report2024-04-29029 April 2024 2023 Annual Radioactive Effluent Release Report AEP-NRC-2024-31, Annual Report of Individual Monitoring2024-04-24024 April 2024 Annual Report of Individual Monitoring IR 05000315/20230042024-01-31031 January 2024 Integrated Inspection Report 05000315/2023004 and 05000316/2023004 ML24004A1582024-01-19019 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0039 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) AEP-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 ML23352A3502023-12-19019 December 2023 Dc. Cook Nuclear Power Plant, Units 1 Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23338A2642023-12-0505 December 2023 Confirmation of Initial License Examination 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 ML23310A1152023-11-0606 November 2023 Notification of the NRC Baseline Inspection and Request for Information, Inspection Report 05000316/2024002 IR 05000315/20234042023-11-0606 November 2023 Cyber Security Inspection Report 05000315/2023404 and 05000316/2023404 IR 05000315/20234032023-09-19019 September 2023 Security Baseline Inspection Report 05000315/2023403 and 05000316/2023403 IR 05000315/20230112023-08-31031 August 2023 Functional Engineering Inspection - Commercial Grade Dedication Report 05000315/2023011 and 05000316/2023011 ML23242A1832023-08-30030 August 2023 Notification of NRC Baseline Inspection and Request for Information; Inspection Report05000315/2023004 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 IR 05000315/20234022023-08-11011 August 2023 Security Baseline Inspection Report 05000315/2023402 and 05000316/2023402 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 IR 05000315/20230022023-07-24024 July 2023 Integrated Inspection Report 05000315/2023002 and 05000316/2023002 ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III IR 05000315/20230122023-06-22022 June 2023 Biennial Problem Identification and Resolution Inspection Report 05000315/2023012 and 05000316/2023012 IR 05000315/20235012023-06-21021 June 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000315/2023501 and 05000316/2023501 AEP-NRC-2023-29, Core Operating Limits Report2023-06-19019 June 2023 Core Operating Limits Report ML23159A0192023-06-13013 June 2023 Information Meeting with a Question and Answer Session to Discuss NRC 2022 End-Of-Cycle Plant Performance Assessment of Donald C. Cook Nuclear Plant, Units 1 and 2 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 2024-08-08
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARML24197A1262024-07-15015 July 2024 Unit 2 - Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating AEP-NRC-2024-48, Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-07-0202 July 2024 Response to Request for Additional Information (RAI) for License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating AEP-NRC-2022-03, Final Supplemental Response to NRC Generic Letter 2004-022022-01-20020 January 2022 Final Supplemental Response to NRC Generic Letter 2004-02 AEP-NRC-2021-68, Response to Request for Additional Information on Requested Change Regarding Containment Water Level Instrumentation2021-12-16016 December 2021 Response to Request for Additional Information on Requested Change Regarding Containment Water Level Instrumentation AEP-NRC-2021-43, Response to Request for Additional Information Regarding Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval2021-07-21021 July 2021 Response to Request for Additional Information Regarding Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval AEP-NRC-2021-16, Unit 2 - Response to Request for Additional Information Regarding CFR 50.55a Request Associated with the Fifth Ten-Year Inservice Testing Interval2021-02-25025 February 2021 Unit 2 - Response to Request for Additional Information Regarding CFR 50.55a Request Associated with the Fifth Ten-Year Inservice Testing Interval AEP-NRC-2021-18, Response to Request for Additional Information Regarding License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency2021-02-18018 February 2021 Response to Request for Additional Information Regarding License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency AEP-NRC-2020-50, Response to Request for Additional Information Regarding License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency2020-07-0909 July 2020 Response to Request for Additional Information Regarding License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency AEP-NRC-2019-56, Seismic Probabilistic Risk Assessment in Response to Near Term Task Force Recommendation 2.1: Seismic2019-11-0404 November 2019 Seismic Probabilistic Risk Assessment in Response to Near Term Task Force Recommendation 2.1: Seismic AEP-NRC-2019-32, Unit 2 - Response to Request for Additional Information Regarding Unit 2 Leak-Before-Break Analysis and Deletion of Containment Humidity Monitors for Unit 1 and Unit 22019-08-22022 August 2019 Unit 2 - Response to Request for Additional Information Regarding Unit 2 Leak-Before-Break Analysis and Deletion of Containment Humidity Monitors for Unit 1 and Unit 2 AEP-NRC-2019-40, Response to Request for Additional Information Regarding License Amendment Request to Address NSAL-15-1, Rev. 02019-07-30030 July 2019 Response to Request for Additional Information Regarding License Amendment Request to Address NSAL-15-1, Rev. 0 AEP-NRC-2018-81, Supplement to Response to Request for Additional Information Regarding License Amendment Request for Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping2018-11-27027 November 2018 Supplement to Response to Request for Additional Information Regarding License Amendment Request for Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping AEP-NRC-2018-82, Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination2018-11-20020 November 2018 Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination AEP-NRC-2018-64, Response to Request for Additional Information Regarding License Amendment Request for Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping2018-09-27027 September 2018 Response to Request for Additional Information Regarding License Amendment Request for Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping ML18334A2722018-09-18018 September 2018 LTR-SDA-II-18-41-NP, Revision 1, Responses to NRC Questions on the Expanded Scope Leak-Before-Break Evaluations for D.C. Cook, Units 1 and 2. AEP-NRC-2018-45, Response to Request for Additional Information Concerning 2017 Decommissioning Funding Status Report2018-08-0909 August 2018 Response to Request for Additional Information Concerning 2017 Decommissioning Funding Status Report AEP-NRC-2018-01, Response to Request for Additional Information Regarding Generic Letter 2016-012018-05-25025 May 2018 Response to Request for Additional Information Regarding Generic Letter 2016-01 AEP-NRC-2018-23, Response to Request for Additional Information Regarding Independent Spent Fuel Storage Installation Decommissioning Funding Plan2018-04-11011 April 2018 Response to Request for Additional Information Regarding Independent Spent Fuel Storage Installation Decommissioning Funding Plan ML18092A0842018-03-28028 March 2018 Donald C. Cook Nuclear Plant Unit 2, Response to Request for Additional Information Regarding Supplemental Information Regarding the Reactor Vessel Internals Aging Management Program ML17346A7662017-12-0808 December 2017 Enclosures 2 & 3 to AEP-NRC-2017-56 - Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels and EAL Technical Basis Manual AEP-NRC-2017-56, Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels2017-12-0808 December 2017 Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels AEP-NRC-2017-30, Response to Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 3.9.3, Containment Penetrations2017-05-26026 May 2017 Response to Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 3.9.3, Containment Penetrations AEP-NRC-2017-16, Submittal of Focused Evaluation in Response to March 12, 2012, Request for Information Regarding Near- Term Task Force Recommendation 2.1: Flooding2017-05-11011 May 2017 Submittal of Focused Evaluation in Response to March 12, 2012, Request for Information Regarding Near- Term Task Force Recommendation 2.1: Flooding AEP-NRC-2017-09, Response to Request for Additional Information Regarding the License Amendment Request for the Containment Leakage Rate Testing Program2017-02-27027 February 2017 Response to Request for Additional Information Regarding the License Amendment Request for the Containment Leakage Rate Testing Program AEP-NRC-2016-81, Unit 2 - Supplemental Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTP-425, Relocate Surveillance Frequencies Program to Licensee Control-Risk Informed ...2016-11-0303 November 2016 Unit 2 - Supplemental Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTP-425, Relocate Surveillance Frequencies Program to Licensee Control-Risk Informed ... AEP-NRC-2016-80, Response to NRC Generic Letter 2016-01: Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools2016-10-31031 October 2016 Response to NRC Generic Letter 2016-01: Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools AEP-NRC-2016-79, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2016-10-12012 October 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident AEP-NRC-2016-69, Follow-up Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies Program to License Control-Risk Informed Technical Specification Task Force.2016-09-0909 September 2016 Follow-up Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies Program to License Control-Risk Informed Technical Specification Task Force. AEP-NRC-2016-56, Response to Seventh Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2016-07-12012 July 2016 Response to Seventh Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2016-48, Unit 2 - Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies Program to Licensee-Control...2016-06-16016 June 2016 Unit 2 - Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies Program to Licensee-Control... AEP-NRC-2016-54, Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force Initiative 582016-06-16016 June 2016 Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force Initiative 58 ML16169A1152016-05-0606 May 2016 Donald C. Cook Nuclear Plant Units 1 and 2 - Response to Sixth Request for Additional Information the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2016-24, Response to Fifth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2016-02-19019 February 2016 Response to Fifth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2016-14, Response to Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2016-01-21021 January 2016 Response to Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation AEP-NRC-2015-11, Response (Part 2) to Fourth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2015-12-17017 December 2015 Response (Part 2) to Fourth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term ML15323A4342015-11-16016 November 2015 Response (Part 1) to Fourth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term ML15323A4332015-11-16016 November 2015 Supplemental Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions. AEP-NRC-2015-98, Supplemental Response to Follow-Up Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program2015-10-30030 October 2015 Supplemental Response to Follow-Up Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program AEP-NRC-2015-99, Response to Request for Additional Information Re License Amendment Request to Revise Technical Specification Section 3.8.1, AC Sources - Operating, Surveillance Requirements 3.8.1.10, 3.8.1.11 and 3.8.1.152015-10-30030 October 2015 Response to Request for Additional Information Re License Amendment Request to Revise Technical Specification Section 3.8.1, AC Sources - Operating, Surveillance Requirements 3.8.1.10, 3.8.1.11 and 3.8.1.15 ML15308A0932015-10-15015 October 2015 Pressurized Water Reactor Owners Group (Pwrog), 15066-NP, Revision 1, Responses to Follow-Up NRC RAI 2 on the D.C. Cook, Units 1 and 2, Reactor Internals Aging Management Program. AEP-NRC-2015-86, Supplemental Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent With. Previously Licensed Conditions.2015-09-18018 September 2015 Supplemental Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent With. Previously Licensed Conditions. AEP-NRC-2015-80, Response to Third Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2015-08-28028 August 2015 Response to Third Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2015-75, Response to Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2015-08-24024 August 2015 Response to Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2015-88, Response to Request for Additional Information Regarding Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI Repair Requirements2015-08-24024 August 2015 Response to Request for Additional Information Regarding Proposed Alternative to the American Society of Mechanical Engineers Code, Section XI Repair Requirements AEP-NRC-2015-69, Response to Follow-Up Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program2015-08-0606 August 2015 Response to Follow-Up Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program ML15223A4362015-07-28028 July 2015 PWROG-15066-NP, Revision 0, Responses to Follow-Up NRC RAI 2 on the DC Cook Units 1 and 2 Reactor Internals Aging Management Program. AEP-NRC-2015-63, Response to Request for Additional Information Regarding 2014 Unit 1 Steam Generator Tube Inspection2015-07-17017 July 2015 Response to Request for Additional Information Regarding 2014 Unit 1 Steam Generator Tube Inspection AEP-NRC-2015-64, Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term2015-07-17017 July 2015 Response to Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term AEP-NRC-2015-60, Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification Section 3.8.1, AC Sources - Operating, Surveillance Requirements 3.8.1.10, 3.8.1.11, and 3.8.1.152015-07-0909 July 2015 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification Section 3.8.1, AC Sources - Operating, Surveillance Requirements 3.8.1.10, 3.8.1.11, and 3.8.1.15 AEP-NRC-2015-66, Response to Request for Additional Information Regarding Exigent License Amendment Request Regarding Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2015-07-0202 July 2015 Response to Request for Additional Information Regarding Exigent License Amendment Request Regarding Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation 2024-07-02
[Table view] |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS >
DI - TRIBLITION FOR INCOMING MATERIAL S0 1 316 REC: DEhlTQN H R GRG: TILLINGHAST J DOCDATE: 0&f2'?/7-NRC IN 5 MI PWR DATE RCVD: 10/ 06/ 78 DGCTYPE: LETTER NOTARIZED: YES COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 1 RESPONSE TG NRC REQUEST OF 07f 12f78... FORWARDING ADDL INFO CONCERNING SUBJECT FAl ILITY "S ICE-BASKET STRE- S ANALYSIS... NOTARIZED 09/29f7-.
PLANT NAt'IE: COOK -- UNIT 1 REVIEWER INITIAL: X JM COOK UNIT 2 DISTRIBUTER INITIAL: Q DISTRIBUTION OF THIS MATERIAL I. AS FOLLOWS NOTES:
I 5 E "- 3 CYS ALL MATERIAL GENERAl DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING t I CENSE.
( DISTRIBUTION CODE A001 )
FGR ACTION: BR CHIEF ORB@1 BC4<W/7 ENCL 1NTERNAL: ca eric>>~w~ew~
-->>v /2 ENCL a) f NRC PDR4 4 W ENCL GELD>>>LTR ONLY HANAUER>>'~ l ENCL lf CORE PERFORMANCE BR++W/ENCL AD FOR SYS 5 PROJ>>i>>W/ENCL ENGINEERlNG BR->>W/ENCL REACTOR SAFETY BR+<W/ENCL PLANT SYSTEMS BR++llfEhlCL EEB>>>>W/ENCL EFFLUFNT TREAT SYS++W/ENCL EXTERNAL: LPDR'S ST. JOSEPH, MI.>>~WfEl lCL TERA><W/ENCL N= IC< >WfFNCL ACRS CAT B~i>>W/16 ENCL DISTRIBUTION: LTR 40 ENCL 39 CONTROL NBR: 7'='0740"->>06
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'& THE END ga y
INDIANA II MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 September 29, 1978 AEP:NRC:00057
!!,p!!!gap )g[(g 3',!.t; QPY Donald C. Cook Nuclear Plant Unit Nos. 1 8 2 Docket Nos: 50-315 and 50-316 License Nos: DPR-58 and DPR-74 rtr~
cia Su lementar Information on Ice Basket Stress Anal sis s r tJ CD Mr. Harold R. Denton, Director Pl Cll Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
In his letter dated July 12, 1978 Mr. A. Schwencer of your office advised us that he would require additional information to complete the NRC staff's review of the Donald C. Cook Nuclear Plant's Ice -Basket Stress Analysis. He requested that we furnish information for all items contained in the request for additional information enclosed therein.
For further clarification of the items contained in the request for additional information, a conference call was arranged with Dr. Gluckman of your staff on August 4, 1978. Members of the American Electric Power Service Corporation staff and Westinghouse Electric Corporation staff along with M. M. Mlynczak of your office participated in this call.
Enclosed herein is an item by item response to your request for additional information, prepared by the Westinghouse Electric Corporation in a format agreed upon during the above mentioned conference call.
Very truly yours, JT:em Zo rI, "ii r
/-C:-dC:'~
'linghast
~I, Vice President Sworn and subscribed to before me this <~day of September, 1978 in New York County, New York. KMNTflWM<lKI'KllWY Notary Public 1 as. hLLrLr'8 BARR< 78074030rc, NCrTAkY a'uu',C, >',eto ct New York i~so. 4l-;O0o(92 Quaiitieo in Ciueens County Certiiicaie tiled in New York County cc: (Attached) Cornrrussroir expires n>arcn 30, 197Z
Mr. Harold R. Denton, Director AEP:NRC:00057 cc: R. C. Cal 1 en G. Charnoff P. W. Steketee R. J. Vollen R. Walsh D. V. Shaller-Bridgman R. W. Jurgensen
ENCLOSURE 1 SUPPLEt~EiNTARY Ii!FORtlAfION ON DONALD C. COOK NUCLEAR PLANT ICE BASKET STRESS ANALYSIS uestlon No. 5 Provide Tables 2 through 7, referred to in ICE-TR-079, page 4,- subsection 4.0. ;--
Additional Res onse Ho. 5 Tables 2 through 7, wnich are referenced on page 4, subsection 4.0 of Test Report ICE-TR-079 are again submitted as an attachment to this response.
Question No. 6 There is still some concern that the test arrangement is not furnishing the most conservative results. It is correct that the ice has in sot.,e cases a stiffening effect. It is however possible to state that for certain buckling shapes the effect of ice pressure on the vertical ligament may make conditions worse.
Discuss this possibility. Discuss also the adequacy of the cruciform supports inserts to carry the ice load for which they are not designed.
Additional Res onse No. 6 Assuming that the ice mass inside the ice baskets is fluid and generates an internal pressure (hydraulic) on the vertical ligaments c ". the basket, it is possible 'to state that for certain buckling shapes the test arrangement is less conservative. However, the internal force generated by a six foot head of ice, supported by the cruciform support inserts, when calcula.ed wou'.d be a maximum of 0.4 lb/in acting against the vertical basket ligaments. Th. horizontal test qualification loads, as shown in Table 1, page 10 of ICE-TR-079, varies from 769 lbs to 1025 lbs. This corresponds to a uniformly distributed load, acting against the vertical ligaments of the ice basket 0 5.3. lb/in to 7.12 lb/in, respectively. The effect of the ice pressure on the vertical ligaments would mean a 7.5Ã to 5.65, respectively, increase in horizontal test lualification loads.
The test results are conserva ive because the test qualification loads envelope the worst possib',e load combinations of horizontal and vertical design loads, and they were incre sed to inc'iude ice maldis-ribution, and single basket test factors. In addition, an examination o the actual test results shows that .he baskets were tested to horizontal loads varyirg from 871 lbs .o 1721 lbs which corresponds to increases of 13> to 68;l over the required test qualification loads, which more than adequately reflects any increased load due to ice pressure.
The cruciform support inserts were designed, and have the primary function, to support the weight of ice above it. The cruciform inserts prevent he ice above it from falling down through into the next 6 foot basket section below, in the event of a meltout during LOCA conditions. The purpose o ho1ding up the ice in the upper sections of the ice column is to maintain ice bed geometry
and no. allow a bypass flow passage to open up due to some random configuration of various ice baskets having voids at various elevations.
uestior. <<o. 7A The response does not include a discussion of the effec oi conc ntrated loads in the hoop direction. Provide this discussion.
Additional Response No. 7A The effect o concentrated loads (test load condition) is more censervative than
- - uniformly distributed loads (actual load condition) in the hoop directiGA. A comparison of the shear diagram for the test case model versus the actual load mod 1 shows that on the test condition th shear loads on th hoop elements will be a minimum of ll~ higher in the local area where the concentrated loads are applied as opposed to the sa'e area on a beam with a uni ormly disiributed
- load. The test setup is therefore justified by the conservative load application to the basket hoop elements.
Ouestion No. 7C Discuss the effect of the semicircular bracket in the axial direction.
Additional Res onse No. 7C The semicircular bracket distributes the concentrated traverse .test loads on the ice basket, such that localized failure of the perforated metal is avoided.
The application of concentrated test loads along the axial directiGA Gf the ice basket is ~:Gl e c-Ase. Vative than t. e actual load appl ication wi ere load would be uniformly distributed. A comparison of the shear diagram for the test condition and actual load condition shows that the axial shear dis ribution
.is the same maximum value in both load cases with th maximum shear occurring at, the ends of the basket. A comparison of the axial distr',bu ion of bending moment from the moment diagram shows that the maximum value occurs at the center span of the baske length for both cases, however, the t-axirum moment for the test case condition is 8.8'~ higher than the actual load case.
The concentrated traverse load application th. ough the sem'.circular bracket is
'a more severe and conservative load condition for a bucklirg tv~e failuresince the vertical mesh ligaments are subjected to high shear loads locally in the area of the semicircular brackets.
uestion No. 98 Include a proof demonstrating that the requirements of General Ice Condenser Design Criteria, Section C-6: Experimental or Test Yerifica.inn of Design are satisfied.
Additional Response ho. 98 The ice baske. bottom end assembly test verification via Test Report ICE-TR-079 demonstra es that the requir ments of the General Ice Condenser Design Criteria Section C-6 has been satisfied. For the test identi fied in ICE-TR-079, ice basket bottom end assemblies, which were manufactured to actual production drawings and procedures, with g.A. releases (ref. para. 2.0 of TP-069), were tested to loads in excess of the, required qualification loads. The qualification test loads identified in Table of TP-069 were developed with the required 1
test qualification load factors as identified in the General Ice Condenser Design Criteria, and w re derated by 10 percent for single test sample.
The following table lists the actual test loads achieved i'n comparison to the required qualification test loads:
l gualification Actual Test Loads Test Loads
, 0 + 1/2 SSE Load Case Horizontal 1176 963 Vertical 5339 5133 D + DBA + SSE Load Case Horizontal 1095 1025 Vertical -5641 -4015 guestion No. 10A The concern is the a 2 span be m used as the model does not reproduce the same moment and shear distribution as obtained in the actual multiple span structure.
Your response covered only the end c"nnection. Cover the condition along the whole length of the structure.
Additional Res"onse No. 10A The 2 span beam referenced on F'gure 9 of TP-069 does not. rep.oduce the same momen and shear distribut on as would be obtained in the actual multiple span structure. The actual structure would be a 8 span beam. A comparison o-maximum shear and moment diagrams for the 8 span beam case versus the 2 span beam shows that; the maximum end support shear value for the 8 span beam would be 5> higher, the maximul momenT. at the supports would be 16,. lory r, and the moment in the span would be 10> higner than the 2 span beam model. "'aximum Again, as in our original respons>> to question 10A, Figure 9 of TP-069 is only used .o calculate the bottom connection reaction or shear load that the basket would see from the application of tare weights to the weight and pul;ey system used in applying the horizontal test loads, reference AppendIx 8 of TP-069.
The bottom connection reaction was calculated to verify, via a load cell a the basket bottom, that the required horizontal test load to the basket had been F
~o
/
~ 0 achieved. The 5.. load discrepancy on the shear load would have little im-ortance, on the reliability of the test results since the. baskets were tested loads i'n excess of 1055 of the required qualification test loacs as to'orizontal shown in our additional response to guestion No. 98.
uestion No. 11 Prove perform that a basket which has not its function.
"failed" but is badly deformed can still Additional Res onse No..ll By definition a basket which has not "failed" would not be badly. deformed.
. Examination of ice baskets which have been successfully tested have not been deformed to any noticeable or measureable extent. Baskets which have been tested to failure hav exhibited only small localized deformations since testing would terminate after the failure point was reached. Failure point being defined, as in our original response Ho. 11, as that point in the tes. wnen increasing the horizontal loads to the basket, the maximum vertical load cannot be maintained with the hydraulic jack at the top of .he tes fixture. That is, the vertical load drops down to some value below the maximum and cannot be held, after being re-establisneci, wiihou" continual actuation of the hydraulic jack.
Therefore, a basket which has not "failed", would not be deformed to any extent and would still be able to perform its function.
REGULATORY INFORMATION DISTRIBUTION SYSTEM (BIDS)
DISTRIBUTION FOR INCOMING MATERIAL 50-316 REC: DENTON H B OBG: TILLINGHAST J DOCDATE: 09/22/78 NRC IN 5 MI PWR DATE RCVD: 10/02/78 DOCTYPE: LETTER NOTARIZED: YES COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 0 FURNISHING AUTHORIZATION TO INCLUDE CQP'( OF WESTINGHQUSE ELEC CORP DOCUMENT RE ANALYSIS TO INVESTIGATE THE LONG-TERM CONTAINMENT TEMPERATURE AND PRESSURE RESPONSE TO A POSTULATED STEAM LINE BREAK IN UNIT NO 2 USIhIG THE LOT IC-3 COMPUTER CODE, ON SUBJECT F PLANT NAME: COOK UNIT 2 REVIEWER INITIAL: X JM DISTRIBUTOR INITIAL;QQ DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS NOTES:
- 1. SEND 3 COP IE OF ALL MATERIAL TG I SE GENERAL DISTRIBUTION FOR AFTER ISSUANCE QF Or ERATING LICENSE.
<DISTRIBUTION CODE A001)
FOR ACTION: BR CIHILI= ORBN1 BC<%LTB ONLY<7)
INTERNAL: Fr~+Lg>>~LTR ONLY) NRC PDR+4LTB ONLY<i)
I SE~uLTB O OELD%<LTB ONLY(i>
IHANAUER4~+LTB ONLY< 1) CORE PERFORMANCE BR<~LTR ONLY(i AD FOR SYS 8. PBOJ~~LTR ONLY(1) ENGINEERING BR>+LTR ONLY(i)
REACTOR SAFETY BR+4LTR ONLY<i) PLANT SYSTEMS BR4%LTR ONLY< 1)
EEB~uLTB ONLY(1) EFFLUEhlT TREAT SYS>+LTR ONLY ( 1 )
J MCGOUGI.I+4LTB ONLY(i)
EXTERNAL: LPDB S ST. JOSEPI.li MI+~LTB ONLY (1 >
TERA~>>LTB ONLY<1)
NSIC>+LTB ONLY(1)
ACBS CAT B+>LTR ONLY<16)
DISTRIBUTION: LTR 40 ENCL 0 COhlTROL NBR: 78160 > 4 SIZE: 3P
H It 44
INDIANA II MICHIGAN POWER COHPANY P, O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 1D004 September 22, 1978 AEP:NRC: 00090
~~1 Donald C. Cook Nuclear Plant Unit No. 2 t~l Docket No. 50-316 Vl C/I Pl License No. DPR-74 Wl W pl
~
Containment Long-Term Temperature and Pressure Response QJ ~ Q7 OC)
Pc)~ Dl~
CA~
Jun nr Q7 Mr. Harold R. Denton, Director C7 5
Office of Nuclear Reactor Regulation Vl U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Denton:
Please. be advised that the Westinghouse Electric Corporation has submitted forty (40) copies of an analysis to investigate the Long-Term Containment Temperature and Pressure Response to a postulated steam line break in Unit No. 2 of the Donald C. Cook Nuclear Plant using the LOTIC-3 computer code. This analysis was performed to fulfill License Condition 2.C.(3)(g) of the Operating License of the above captioned facility.
This analysis was transmitted to you as an enclosure to the Westinghouse Electric Corporation letter (NS-TMA-1946) dated September 20, 1978 from Mr. T.M. Anderson to you.
This letter is to authorize the inclusion of the above Westinghouse document on our Docket No. 50-316, issued to the Unit No. 2 of the Donald C. Cook Nuclear Plant.
Very truly yours, 1 ghast ce President Sworn and subscribed to before, me this 2a' day of September, 1978 78i600284 in New York County, New York Notary Pu GREGORY M. GuRICAN Ifotary Public, state of Ilaw York Het 8 No. 31-4643431 cc: (attached) Qualified In flew York CountV~~
Cornrnlsslcn Exp!res March 30, 19'. e l
1 ~
lCJ E7 c
lt O'
CO
J.'Ir>. Harold'R. Denton September 22, 1978 cc: R. C. Callen G. Charnoff P. W. Steketee R. J. Vollen R. Walsh R. W. Jurgensen D. V. Shaller - Bridgman
J'r. Harold R. Dento eptember 22, 1978
~ A bc: S. J. Hilioti J. I. Castresana/P. K. Eapen K. J. Vehstedt R. F. Hering/S. H. Steinhart R. F. Kroeger H. N. Scherer, Jr.
J. F. Stietzel - Bridgman M. N. Nlynczak - NRC N. H. Judkis - Westinghouse J. D. Woodward - Westinghouse DC-N-6015.1 DC-N-6079 AEP:NRC: 00090