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Category:Letter
MONTHYEARAEP-NRC-2024-77, U2C28 Steam Generator Tube Inspection Report2024-10-21021 October 2024 U2C28 Steam Generator Tube Inspection Report AEP-NRC-2024-80, Independent Spent Fuel Storage Installation Registration of Dry Spent Fuel Storage Cask2024-10-15015 October 2024 Independent Spent Fuel Storage Installation Registration of Dry Spent Fuel Storage Cask AEP-NRC-2024-79, Unit 2, Independent Spent Fuel Storage Installation - Registration of Dry Spent Storage Cask2024-09-26026 September 2024 Unit 2, Independent Spent Fuel Storage Installation - Registration of Dry Spent Storage Cask AEP-NRC-2024-78, Reply to a Notice of Violation: EA-24-0472024-09-23023 September 2024 Reply to a Notice of Violation: EA-24-047 05000316/LER-2024-002-01, Manual Reactor Trip Following Rapid Downpower for Steam Leak2024-09-12012 September 2024 Manual Reactor Trip Following Rapid Downpower for Steam Leak IR 05000315/20244022024-09-10010 September 2024 Security Baseline Inspection Report 05000315/2024402 and 05000316/2024402, Independent Spent Fuel Storage Installation Security Inspection Report 07200072/2024401 AEP-NRC-2024-69, Core Operating Limits Report2024-09-0909 September 2024 Core Operating Limits Report IR 05000315/20243012024-09-0505 September 2024 NRC Initial License Examination Report 05000315/2024301 and 05000316/2024301 ML24225A0022024-09-0303 September 2024 Issuance of Amendment Nos. 363 and 344 Revising Technical Specifications Section 3.8.1, AC Sources-Operating, for a One-Time Extension of a Completion Time IR 05000315/20240112024-08-30030 August 2024 NRC Inspection Report 05000315/2024011 and 05000316/2024011 and Notice of Violation AEP-NRC-2024-51, Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes2024-08-28028 August 2024 Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes AEP-NRC-2024-76, Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-08-28028 August 2024 Unit 2 - Supplement to License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating 05000316/LER-2024-003, Plant Shutdown Required by Technical Specifications Due to Reactor Coolant System Identified Leakage2024-08-22022 August 2024 Plant Shutdown Required by Technical Specifications Due to Reactor Coolant System Identified Leakage IR 05000315/20240052024-08-21021 August 2024 Updated Inspection Plan for Donald C. Cook Nuclear Plant, Units 1 and 2 (Report 05000315/2024005 and 05000316/2024005) AEP-NRC-2024-61, Unit 2 - Response to Request for Additional Information for Neutron Flux Instrumentation License Amendment Request2024-08-15015 August 2024 Unit 2 - Response to Request for Additional Information for Neutron Flux Instrumentation License Amendment Request ML24221A2702024-08-0808 August 2024 Unit 2 Independent Spent Fuel Storage Installation - Registration of Dry Spent Fuel Storage Cask AEP-NRC-2024-62, Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask2024-08-0707 August 2024 Unit 2, Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Storage Cask ML24256A1482024-08-0202 August 2024 2024 Post Examination Submittal Letter 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 05000316/LER-2024-002, Manual Reactor Trip Following Rapid Downpower for Steam Leak2024-07-15015 July 2024 Manual Reactor Trip Following Rapid Downpower for Steam Leak 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 ML24159A2522024-05-30030 May 2024 10 CFR 50.71(e) Update and Related Site Change Reports AEP-NRC-2024-23, Core Operating Limits Report2024-05-23023 May 2024 Core Operating Limits Report 05000316/LER-2024-001, Reactor Trip Due to Main Turbine Trip from a High-High Thrust Bearing Position Trip2024-05-20020 May 2024 Reactor Trip Due to Main Turbine Trip from a High-High Thrust Bearing Position Trip 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 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 IR 05000315/20244012024-05-14014 May 2024 – Security Baseline Inspection Report 05000315/2024401 and 05000316/2024401 AEP-NRC-2024-24, Form OAR-1, Owners Activity Report2024-05-0707 May 2024 Form OAR-1, Owners Activity Report ML24115A2152024-05-0707 May 2024 LTR: CNP Non-Acceptance with Opportunity TS 3-8-1 ML24256A1472024-05-0606 May 2024 DC Cook 2024 NRC Examination Submittal Letter: Submittal 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 AEP-NRC-2024-02, Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating2024-04-0303 April 2024 Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating AEP-NRC-2024-29, (CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-072024-04-0303 April 2024 (CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-07 AEP-NRC-2024-19, Annual Report of Property Insurance2024-04-0101 April 2024 Annual Report of Property Insurance 2024-09-09
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARAEP-NRC-2024-61, Unit 2 - Response to Request for Additional Information for Neutron Flux Instrumentation License Amendment Request2024-08-15015 August 2024 Unit 2 - Response to Request for Additional Information for Neutron Flux Instrumentation License Amendment Request 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 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-2024-11, Unit 2 - Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation2024-02-27027 February 2024 Unit 2 - Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation 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-2016-54, 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 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... 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 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. 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 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 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 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 2024-08-15
[Table view] |
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NBC FQRM 195 U.S. NUCLEAR REGULAT ISS ION DOCKET (2.7G) -31 16 F IL NRC,.DISTRIBUTION FOR PART 50 DOCKET MATERIAL TO: A. ~~~ FROM: DATE OF DOCUMENT Mr. Dennis L. Ziemann Exxon Nuclear Co., Inc. 11/30/76 Richland, Washington DATE RECEIVED G. F. Owsle 12/9/76 PfLETTER 0 NOTO R I7 E D PROP INPUT FORM NUMBER OF COPIES RECEIVED
@ORIGINAL PIUNC LASS IF I E 0 One signed Ocopv 40 co ies encl. recvd.
DESCIIIPTION ENCLOSURE Ltr. re our 11/23/76 ltr....trans the following: Furnishes reques ted additional information regarding Exxon Nuc1ear Company Report XN-76-51, concerning information supporting the operation of the Cook Plant Unit 1 for fuel Cycle 2.
(1-P) (9-P)
DO NOT RE~OS PLANT NAME:
Coolc 1 & 2 ACKN0%LEDGZD SAFETY FOR ACTION/INFORMATION 12976 ASSIGNED AD:
N ~ ~ Ziemann/Kniel RO EC MANA E Fle tcher/Benedic t PROJECT MANAGER'IC LIC ASST Diggs Lee ASST 0 INTERNAL D IST R I BUTION REG FILE SYSTEMS SAFETY PLANT SYSTEMS SXTE SAFE HEINEMAN TEDESCO IGE SCHROEDER OELD GOSSICK & STAFF ENGINEERING IPPOLITO ENV MXPC MACARRY ERNST CASE KNIGHT BALLARD HANAUER ~
SIlÃEIL OPERATXNG REACTORS SPANGLER HARLESS PA CK STELLO SITE TECH PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAMMILL BOYD ROSS EXSENHUT STEPP P ~ COLLINS NOVAK HUL)IAN HOUSTON ROSZTOCZY PETERSON CHECK E SXTE ANALYSIS MELTZ VOLLMER HELTEMES AT& I BUNCH SKOVHOLT SALTZMAN J ~ COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION CONTROL NUMBER LPDR ~ St. Josenh MicI . NAT LAB ~ WÃMIQDEK1 TIC: REG V ~ XE ULR KSON OR NSXC: LA PDR ASLB: CONSULTANTS:
ACRS CYS 88&9%H6/ E T ' ~
NRC FORM 196 {2.7G)
h ~ ~
it
E)f(ON NUCLEAR COMPANY, Inc.
2101 Horn Rapids Road, Richland, Washington 99352 PHON f: t509) 946-9621 November 30, 1976 5
File
'Regulatory Docket Mr. Dennis L. Ziemann, Chief Operating Reactors Branch //2 Division of Operating Reactors
'uclear Regulatory Commission Washington, D.C. 20555
Dear Dennis:
In your letter to Mr. John Tillinghast dated November 23, 1976 you requested additional information be supplied regarding Exxon Nuclear Company Report XN-76-51. This report provided information supporting the operation of the D. C..Cook Nuclear Plant Unit //1 for fuel Cycle 2.
1 This letter transmits responses to this request for additional infor-mation for your review. One copy of these responses is being transmitted via telecopier; forty (4P) copies are being transmitted under separate cover.
Ver truly you G. F. Owsley, Manager Reload Licensing GFO:gf Attachments As above c~.$ .7 CC: Mr. John Tillinghast P'cP~i++p cP~>~
~i E
12&4'N AFFILIATE OF EXXON CORPORATION
1 4
l I N I
I4 I II
ADDITIONAL INFORMATION FOR D. C. COOK QUESTION 1 The nodalisation diagram shoran on Pigure S.1 of XN-76-36, for the RELAP reflood calculation uses an axially split doumcomer a)ith accumulator and Safety Injection System (SIS) input to the dovncomer regions. Pvovide additionaZ description and justification for the use of this nodaZisation and pater injection assumption relative to the previous ENC- REM reflood model.
RESPONSE
As discussed in XN-76-36, the RELAP4-EM/FLOOD model used for the D. C.
Cook analysis used two volumes to represent the downcomer region. This nod'alization is a more realistic representation of the actual reactor system than the previously used single-volume downcomer, in that,' si gnificant cross-
=sectional area change occurs in the downcomer region. The lower downcomer node includes the regions on both sides of the thermal shield, the region between the core barrel and the core baffle, and the core bypass. The .
total flow area for this region is 49.53 ft . The upper downcomer region includes the volume between the reactor vessel and the core barrel with a flow area of 32.41 ft2 . Since the liquid height in the downcomer equals the liquid volume divided by the horizontal flow area, a change in flow area will alter the liquid height; and hence, the driving head for reflood.
The two-volume downcomer permits the area change to be modeled for the D. C.
Cook reactor.
The Safety Injection System (SIS) modeling for the D. C. Cook reflood calculations is described in XN-76-36. Initially SIS flow and accumulator flow are modeled as a fill system injecting into the lower downcomer region
with fluid conditions near saturation (saturation at lowest containment pressure); This model is consistent with the approved ENC-WREN .PWR model.,
Pressure drop penalties are applied at the intact loop junction to the pressure 'vessel to account for interaction effects of ECC fluid and super heated steam in the cold leg pipes. This model is in accordance with the approved ENC-WREN PWR model as described in XN-75-41, Supplement 5, Revision l.
When steam flow is established in the intact loops, the SIS flow is switched from the lower downcomer to the upper downcomer region and is input at the actual fluid temperature of the SIS water. At the time steam flow is established'in the intact loops, the steam flow and enthalpy is sufficient to heat the SIS fluixl to saturation by condensing some of the intact loop steam flow. Since SIS fluid and intact loop steam must flow through the same pipes, mixing of these fluids is 'expected. Thus, the assumption of homogeneous equilibrium inherent in the RELAP4-EH/FLOOD program realistically represents the expected conditions and conservatively minimizes the steam flow from the upper downcomer to the containment. This results in a pressure drop at the steam slip flow junctions (to the containment)
'in'imum and a conservative reflood system pressure.
QUESTION 2 A fluid temperature in the upper head region equal to the hot leg temperature shall be used unless a lessee temperature is justified bp actual measurements in a simile plant oz unless a lesser temperature results in a highe>'eak cladding temper'atmo t'hcvt the hot leg tempest;u>'c.
RESPONSE
The upper head fluid temperature sensitivity study has been completed.
The base case consisted of the 1.0 DECLS reported in XN-76-51, the limiting break for the D. C. Cook Nuclear Unit 1 with the upper head temperature.set
to the hot leg temperature. A second calculation was performed, identical to the first, except the upper head temperature was set equal to the average of the'ot leg and cold leg temperatures. This results in a decrease in cladding 'temperature at EOBY of 52'F and decrease in volume averaged'emperature at EOBY of 47'F. This results in an approximate 24'F decrease in peak cladding temperature if the lower value of upper head temperature was used., Thus, the use of. the hot leg temperature for the upper head is conservative.
QUESTION 3 Describe and justify the phase separation model assumed for the upper head dur ing &Ecedoam; P
RESPONSE
A phase 'separation model was input to the upper head region using the available RELAP4 bubble rise model. The use of the phase separation model is justified in that the upper head region is relatively stagnant and thus phase separation is expected to occur. A]so, the flow path from the upper head to the upper plenum occurs at the top of the control rod guide structure and the possibility exists that when the mixture level falls below this level, only steam will flow to the upper plenum with the remaining upper head liquid being held up in the upper head region and thus unavailable for core cooling'during blowdown. A phase separation model is necessary to, consider these effects.
Hodeling the upper head with phase separation model is a more conservati ve representation than a homogeneous model and is also a more realistic representation of this region. The parameter values used for the phase'
separation model are derived empirically for blowdowns from nearly stagnant vessels similar to the upper head region. A sensitivity study w'as performed
. for D. C; Cook system which confirmed that the separated upper head model resul'ted in a higher PCT than a comparable calculation using a homogeneous model.
Upper head nodalization studies have been completed. The first case was reported in XN-76-36, as the 4-loop ice condenser sample problem and was done with a homogeneous model and with the upper head temperature set to the hot leg temperature. A second calculation has been performed identical to the first, except the upper head was modeled as a separated volume. This results in less water ava'ilable for core cooling during the blowdown phase since some of the upper head water is trapped in the upper head, below the top of RCC assembly guide tubes.
The reduced cooling causes an increase in cladding temperature at EOBY of,142'F and an increase in fuel, averaged temperature at EOBY of 90'F.
This difference results in an approximate 45 F increase PCT due to the use of a separated rather than a homogeneous volume.
QUESTION 4 Provide a calculated effect on peak cladding temperature for each of the fo72otuing model changes incor'por'ated in the ENC-VBEM-II model:
A. Floe Blockage B. FLECHT/ENC3 multiplier s C. Hot eall delay D. Steam cooling E. Be flood model-doumcomer nodali "ation (g1 above)
RESPONSE
As requested by the NRC Staff, ENC is providing the change in PCT resulting from each of the model 'changes, comprising ENC-NREM:II, as well as the total change in going from ENC-WREM-I to ENC-WREM-II. These were originally provided in Reference 1, but the model used for these calcu-lations has been modified, resulting from NRC Staff review, thus, a new insensitivity study was performed. The sensiti vity studies are based on the 1.0 DECLS for. the D. C. Cook Unit 1 nuclear plant reported in Reference 2 except the maximum LHGR is 12.14 kW/ft rather than 13.68 kW/ft. It is important to recognize that if the sensitivity study had been performed for a plant without an ice condenser containment system (H. B. Robinson or Palisades), the only significant effect of the new model (ENC-WREM-II) would result from the 'improved hot wall delay model and the LPCT would be 45 to 60'F.
Table 1 summarizes the results of the requested sensitivity studies.
It is to be noted that the ENC-WREM-I is a conservative model which was never intended to be used for plants with extended period of reflood rates less than 1.0 inch/sec. Note that when ENC-WREN-II is applied to this calculation, rupture is not calculated to occur. In. order to make a realistic comparison between the two models, the rupture temperature was reduced about 50'F. This causes rupture to occur and the code to switch
'o the steam cooling model. By forcing the code to switch to the steam cooling model, representative comparisons can be made between the old and new models. However, the data for the calculations in which rupture was not forced is included"for completeness.'
4 The reduction in PCT due to ENC-WREN-II is about 540 " for the D. C.
Cook plant. As indicated by the table, the temperature decrease is due mainly to the improved calculation of flow around the blockage and to the FLECHT/ENC3 multipliers. This is not unexpected since the ENC-WREN-I model assumed an 805 flow area reduction (maximum calculated flow area reduction) near the blockage; but only a 23K area reduction is calculated to occur. ENC-WREM-II takes into account the actual amount of blackage+
which is calculated to occur; thus, ENC-WREN-II calculates a more realistic flow at and downstream of the blockage. The improved low flood rate FLECHT heat transfer i s based on the expanded data base available in Reference 2 which was not available during ENC-WREN-I development. This data showed improved heat transfer during the. initial 50 seconds of the transient to the original data. 'elative Although the ENC-WREN-II steam cooling model with the ENC3 multipliers conservatively predicts 100/ of the low flood rate low pressure data between the 4 and 8 foot elevations, the steam cooling model change led to an increase in cladding temperature as compared to the previous model.
The 90;F reduction in PCT due to the hot wall delay model change is caused by the reduction in the ECC water spilled during the refill period-.
This water becomes available for cooling the core and for filling the downcomer.
- This flow area reduction due to ballooning calculated with the ENC flow blockage model approved in ENC-WREM-I.
The axially split downcomer properly accounts for the change in the cross-sectional area of the downcomer at the top of the thermal shield.
This allows for an accurate calculation of the liquid height in the down-comer. Since the liquid velocity in the downcomer is small, the additional inertial and frictional terms are negligible; thus, it only affects the calculation through the determination of the liquid height. For the D. C.
Cook case, the downcomer is filled by the accumulators very early in the transient and the downcomer remains filled for the remainder of the tran-sient. Therefore, this change has no effect on PCT for D. C. Cook.
'I
TABLE 1 ENC-WREM-II MODEL CHANGE SENSITIVITY STUDIES (ICE CONTAINMENT REACTOR)
Hot Wall . ENC3 SC ENC-WREM-1 Dela Multi liers Blockaqe Model ENC-WREM-II PCT 2 i53. 2363. 1727.1 2250.2 1999. 2629. 31 1905.c (F)
Time (sec) of PCT 344. 332. 342. 350. 274. 366. 336 '82,.
Location of PCT 7.1 7.1 8.9. 7.4. 7.1 7.1 8.9 (ft) 7.4'ime of Rupture 120.
- 129. 131 .120. '1 20. 143.
(sec)
Location of Rupture 6.9 6.9 7.1 6.9 6.9 7 1 (ft)
Max. 2r02 11.5 9.4 1.40 6.9 3.7 15.9 1.36 2.6
(~)
Location Hax. 2r02 7.1 7.1 7.9 7.4 7.1 7.1 7.9 7.4 (ft) aPCT -90. -726. -203. -454. +176. -730. -548 If No rupture.
Forced rupture.
~ ~ ~ ~
REFERENCES
- 1. Worley, L. C., Rowe, D. S., and Galbraith,- K. P., "Exxon Nuclear Company WREH-Based Generic PWR ECCS Evaluation Hodel Update ENC-WREN-II",
XN-76-27, July, '1976.
- 2. "Donald C. Cook Unit 1 LOCA Analysis Using the ENC WREH-Based PWR ECCS Evaluation Hodel '(ENC-WREH-II)", XN-76-51, October 1976.