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Category:Letter
MONTHYEARIR 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 IR 05000315/20234042023-11-0606 November 2023 Cyber Security Inspection Report 05000315/2023404 and 05000316/2023404 ML23310A1152023-11-0606 November 2023 Notification of the NRC Baseline Inspection and Request for Information, Inspection Report 05000316/2024002 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 AEP-NRC-2023-27, Annual Radiological Environmental Operating Report2023-05-15015 May 2023 Annual Radiological Environmental Operating Report ML23131A3282023-05-11011 May 2023 D.C. Cook Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000315/2023404 and 05000316/2023404 IR 05000315/20230012023-05-0303 May 2023 Integrated Inspection Report 05000315/2023001 and 05000316/2023001 AEP-NRC-2023-19, Annual Radioactive Effluent Release Report2023-04-30030 April 2023 Annual Radioactive Effluent Release Report ML23117A0062023-04-27027 April 2023 Review of the Spring 2022 Steam Generator Tube Inspections Report ML23114A1142023-04-25025 April 2023 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection AEP-NRC-2023-23, Annual Report of Individual Monitoring for 20222023-04-24024 April 2023 Annual Report of Individual Monitoring for 2022 AEP-NRC-2023-24, Notification of Ph Non-Compliance for Turbine Room Sump2023-04-12012 April 2023 Notification of Ph Non-Compliance for Turbine Room Sump AEP-NRC-2023-20, Annual Report of Property Insurance2023-04-0303 April 2023 Annual Report of Property Insurance AEP-NRC-2023-15, Decommissioning Funding Status Report2023-03-28028 March 2023 Decommissioning Funding Status Report ML23076A0212023-03-20020 March 2023 Request for Information for NRC Commercial Grade Dedication Inspection; Inspection Report 05000315/2023011; 05000316/2023011 IR 05000315/20234012023-03-16016 March 2023 Security Baseline Inspection Report 05000315/2023401 and 05000316/2023401 ML23066A1882023-03-0707 March 2023 Information Request to Support Upcoming Problem Identification and Resolution (Pi&R) Inspection at Donald C. Cook Nuclear Plant IR 05000315/20220062023-03-0101 March 2023 Annual Assessment Letter for Donald C. Cook Nuclear Plant, Units 1 and 2 (Report 05000315/2022006 and 05000316/2022006) IR 05000315/20220042023-02-0101 February 2023 Integrated Inspection Report 05000315/2022004 and 05000316/2022004 and Exercise of Enforcement Discretion AEP-NRC-2023-11, Form OAR-1, Owner'S Activity Report2023-01-31031 January 2023 Form OAR-1, Owner'S Activity Report IR 05000315/20230102023-01-31031 January 2023 Phase 4 Post-Approval License Renewal Inspection Report 05000315/2023010 and 05000316/2023010 AEP-NRC-2023-02, Request for Approval of Change Regarding Neutron Flux Instrumentation2023-01-26026 January 2023 Request for Approval of Change Regarding Neutron Flux Instrumentation ML22363A5622023-01-0404 January 2023 Relief Request ISIR-5-06 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles AEP-NRC-2022-66, Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring2022-12-15015 December 2022 Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring AEP-NRC-2022-46, Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in2022-12-12012 December 2022 Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in ML22340A1392022-11-30030 November 2022 Submittal of Revision 31 to Updated Final Safety Analysis Report and 10CFR50.71(e) Updated and Related Site Change Reports IR 05000315/20220112022-11-0404 November 2022 Design Basis Assurance Inspection (Teams) Inspection Report 05000315/2022011 and 05000316/2022011 IR 05000315/20220032022-10-28028 October 2022 Integrated Inspection Report 05000315/2022003 and 05000316/2022003 AEP-NRC-2022-58, U1C31 Steam Generator Tube Inspection Report2022-10-24024 October 2022 U1C31 Steam Generator Tube Inspection Report AEP-NRC-2022-61, Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-062022-10-24024 October 2022 Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-06 2024-01-08
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARAEP-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 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-2015-98, 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-2015-88, 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-2015-75, 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-2015-69, 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-2015-64, 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-2015-63, 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-2015-60, 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 ML18219D0951978-09-29029 September 1978 D.C Cook - Acknowledges 07/12/1978 Letter Advising of Additional Information to Complete NRC Staff'S Review of D.C Cook Nuclear Plant'S Ice - Basket Stress Analysis ML18219D7991978-09-29029 September 1978 Response to Request for Additional Information on Ice Basket Stress Analysis, Prepared by Westinghouse Electric Corporation 2022-01-20
[Table view] |
Text
!
I gRC.FORM 15$
IZ.7B)
NRC DISTRIBUTION FoR PA Mr. George E. Lear GO FROM:
DOCKET MATERIAL EExxon Nuclear Co.
RichMnd,Washington Inc.
t U.S. NUCLEAR REGULATORY COMMISSION 99352 DOCICET NUMBER FILE NUMBER DATE OF DOCUMENT 07-01-77 S.
John A. White- DATE RECEIVED 07-05-77 gLE3'T E R 0 NOTOR IZE D PROP INPUT FORM NUMBER OF COPIES RECEIVED SIORIGINAL @UNCLASSIFIED Cl cop v S I C A)EjD DESCRIPTION ENCLOSU RE Furnishing esponse to NRC's questions of 06-08-77 concernin , Second-round Responses to XN-76-51, Supplement ., "Flow Blaokage and Exposu Sensitivity Study fog;ENC D. C. Cook Unit 1 Reloa Fuel Using ENC WREM N Model", dated 01-27-77...
I 1 page 4 pages 1 PIIANT N~IE. CDOOK UNIT ¹ 1 jcm 07-11-77 FOR ACTION/INFORMATION ENVIRH&iENTAL ASSIGNED AD: Va MOORE LTR BRANCH CHIEF:
ECT MANAGER: PROJECT MANAGER!
ICENSING ASSISTANT: LICENSING ASSISTANT:
Bo HARLESS INTERNAL D ISTRI BUTION SYSTFMS SAFETY PLAiNT SYSTEMS SITE SAFETY &
HEINEMAN TEDESCO ENVIRON ANALYSIS ROEDER BENAROYA DENTON & MULLER ENGINEERING IPPOLITO ENVIRO TECH ERNST OPERATING REACTORS BALLARD YOUNiGBLOOD BAER VAK BUTLER GAMMILL 2 RO ZTOCZY GRIIP CHECK SI'ZH ANALYSiS VOLLMER AT&I BUNCH ALTZMAN J ~ COLLINS RUTBERG KREGER EXTERNAL OISTRIBUTION CONTROL NUiVIBER TIC NSIC EG IV J HANCHETT 771 P2P 1 7EI 16 CYS ACRS SENT CAT GO Y
I A II C
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NUCLEAR COMPANY, Inc.
'QON 2101 Horn Rapids Road P. 8. Box 130, Richland, Washington 99352 Phone: (509j 943-8100 Telex: 32.8353 July 1, 1977 Mr. George E. Lear Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555
SUBJECT:
Second-round Responses to XN-76-51, Supplement 1, "Flow Blockage and Exposure Sensitivity Study for ENC D. C. Cook Unit~ Reload Fuel Using ENC WREM-II Model", dated January 27, 1977
Dear George:
Transmitted herewith is Exxon Nuclear Company's response to NRC questions received informally June 8, 1977 on the subject document.
The enclosed information was telecopied to Mr. P. E. Norian of the NRC on June 30, 1977.
h lit If you have any comments or questions regarding'he enclosed respons'es, please call.
Sincerely, ON NUCLEAR CO , INC.
John A. White Reload Licensing JAW:gf Enclosures As noted CC: Mr. Paul E. Norian (NRC)
Mr. James G. Feinstein (AEP) IQQ 5 IN1.
ggs /0 AN AFFILIATE OF EXXON CORPORATION
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ATTACHMENT A SECOND ROUND UESTIONS FOR XN-76-51, SUPPLEMENT 1 QUES'TION 2 Since'TOODEE2 twas designed to envelop bur'st str'ains, shy are "uniform" strains used as input?
RESPONSE
Uniform strains are not input to TOODEE2. Rupture strains are input.
The code then assumes that the. uniform striah is 0.2 times the rupture strain (c .). Thus, uniform strain is affected by the input rupture strain. Section 2.2 of XN-76-51, Supplement 1, indicates that a value of 0.263 x cR is a better estimate of the uniform strain. The use of 0.2 x cR is conservative since it results in higher pin pressures.
QUES'TION 2 Does Figure 4 contian meed strains (burst and "uniform" strains"? I' this figure used for. any purpose other'than as an illustration;
. RESPONSE Figure 4 represents 'measured flow area reduction for all cases except for ORNL single rod data. For the ORNL data, the single rod rupture strain was used in conjunction with a probability of coplanar blockage obtained from ORNL Multirod Hurst Tests. The purpose of Figure 4 is to indicate that the approved ENC model is very conservative when compared to blockage data steam environment from tests which are prototypic of the anticipated LOCA conditions.
Figure 4 is not used in the ENC model.
The rupture temper ture, as Listed in the ORNA repor ts, is the hottest
'I r'ecorded temperature, not necessarily the temperature obtained from the thermocouple located closest to the rupture. If the mpture temperatures
0 used in the data anaLysis vere aZZ the hottest recorded temper'atur'es, vouZd
, the resuZts be aZtered?
RESPONSE
If temperatures from the therriiocouples measuring the highest temperatures are used instead of the temperatures from the thermocouples nearest the failure location, then the correlating temperatures'or tests PS-17, PS-19, SR-l, SR-2, SR-5, and SR-8 would be altered. The location of data points on Figure 3, XN-76-51, Supplement 1, would be altered as shown in the attached'igure l.
All the measured uniform strains would be above the design correlation and no changes would be necessary for the application of the data. In some cases the highest temperatures were measured at locations over a foot from ,
the ruptured area and were n'ot thought to be as representative of the failure conditions ys the temperatures recorded nearer the failure location. Less than half of the data points were not at the maximum temperatures and only in the cases of PS-17 and SR-5, which were the lowest temperature failures,
.did the highest and failure zone temperatures differ by more than 100'F.
QUES'TION 4 a
In regard to question 4 of the fir'st r'ou>>d questions, Bar lou)'s,thi>>
u)aZZed equation is aZso.used in design for eZastic applications. Justify shy it is more appZicabZe than Iame's thick vaZZed equation.
RESPONSE
The'use of stresses calculated with Barlow's thin walled equation is more appropriate than the use of stresses calculated with Lame's thick walled equation in all applications where the inelastic components of de-formations are of -similar or greater. magnitude than the elastic components.
In many design applications concerned with primary stress limits, reference to yi.eld conditions has to do with the load carrying criteria associated with
,f the limit load. At the limit load, the full thickness of the section is
. 'at the yield stress. This type of criteria is used, for example, in the A.S.H.E. Boiler code. For the limit type evaluation, Barlow's equation is applicable.
For high cycle fatigue, for the determination of the onset of initial .
plastic yielding and for the deformation of brittle materials, the elastic solution o'f Lame is applicable. Such application's are most common in machine design, for rotating equipment and for vibrational and cyclic loadings.
Application of the appropriatedesign relationships have to be related to the situation being modeled and to the source of test data used to establish the design correlation. In the application for LOCA analyses, the cladding inelastic deformations .are orders of magnitude larger than the elastic -deformations, and therefore, the elastic solution does not properly describe the deformation process. The criteria. are associa-ted with the ultimate strength rather than with the fatigue strength. Thus, Barlow's equation is used and not Lame's.
b 0.5 CO I I Clt Clt 0.4 LSl 0 Rupture Strain t
C/)
CL N + Uniform Strain I
LA CA I ~CL Clt t O t oCt I'7 I 0 O CC Clt 0
0.1 Design Non-Rupture Straan Lcm>t (c u = 0.2
0
'400 '500 1600 1700 1800 1900 2000 . 2100 2200 Ropture Temperature ('F)
FIGURE 1 DESIGN STRAIN LIllITS DEPENDENCE ON TEMPERATURE AND ORNL ROD SIMULATOR TEST RESULTS IN STEA'1 STHOSPHERES
~ ~
' ~ t.+
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