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Category:Letter type:RA
MONTHYEARRA-24-0012, Independent Spent Fuel Storage Installation - Annual Radioactive Effluent Release Report2024-02-0505 February 2024 Independent Spent Fuel Storage Installation - Annual Radioactive Effluent Release Report RA-24-0015, Submittal of 2023 Sea Turtle Annual Report2024-01-10010 January 2024 Submittal of 2023 Sea Turtle Annual Report RA-23-0242, Proposed Alternative for the Inspection of Reactor Vessel Closure Head Penetrations in Accordance with 10 CFR 50.55a(z)(2)2024-01-10010 January 2024 Proposed Alternative for the Inspection of Reactor Vessel Closure Head Penetrations in Accordance with 10 CFR 50.55a(z)(2) RA-23-0325, Submittal of Procedures CSD-EP-HNP-0101-01, 02, CSD-EP-ONS-0101-01, CSD-EP-RNP-0101-01, and EP-RNP-EPLAN-ANNEX2024-01-0808 January 2024 Submittal of Procedures CSD-EP-HNP-0101-01, 02, CSD-EP-ONS-0101-01, CSD-EP-RNP-0101-01, and EP-RNP-EPLAN-ANNEX RA-23-0295, Delay of Planned End Date for Fourth 10-Year Inservice Testing (1ST) Program Interval2024-01-0404 January 2024 Delay of Planned End Date for Fourth 10-Year Inservice Testing (1ST) Program Interval RA-24-0006, 10 CFR 50.54(q) Evaluation2024-01-0404 January 2024 10 CFR 50.54(q) Evaluation RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0320, Notification of Deviation from MRP-227, Revision 1-A Baffle-to-Former Bolt Inspection Frequency2023-12-14014 December 2023 Notification of Deviation from MRP-227, Revision 1-A Baffle-to-Former Bolt Inspection Frequency RA-23-0306, Procedures CSD-EP-BNP-0101-01, EAL Technical Basis Document, Revision 006 and CSD-EP-CNS-0101-01, EAL Technical Basis Document, Revision 005, Summary Of.2023-12-12012 December 2023 Procedures CSD-EP-BNP-0101-01, EAL Technical Basis Document, Revision 006 and CSD-EP-CNS-0101-01, EAL Technical Basis Document, Revision 005, Summary Of. RA-23-0318, Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-0404 December 2023 Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RA-23-0284, RA-23-0284 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-16016 November 2023 RA-23-0284 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RA-23-0182, Duke Energy Carolinas, LLC - License Amendment Request to Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program for a One-Time Extension of the Units 1, 2 and 3 Type a Leak Rate Test Frequency2023-11-16016 November 2023 Duke Energy Carolinas, LLC - License Amendment Request to Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program for a One-Time Extension of the Units 1, 2 and 3 Type a Leak Rate Test Frequency RA-23-0288, End of Cycle 27 (C1 R27) Steam Generator Tube Inspection Report2023-11-16016 November 2023 End of Cycle 27 (C1 R27) Steam Generator Tube Inspection Report RA-23-0304, Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 2 Cycle 32 (Revision 0)2023-11-14014 November 2023 Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 2 Cycle 32 (Revision 0) RA-23-0276, Response to Request for Additional Information Regarding License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specific2023-11-0606 November 2023 Response to Request for Additional Information Regarding License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specific RA-23-0279, Supplement to Application to Adopt Risk-Informed Completion Times TSTF-505, Revision 2 and Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power2023-11-0202 November 2023 Supplement to Application to Adopt Risk-Informed Completion Times TSTF-505, Revision 2 and Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power RA-23-0281, Procedure EP-ALL-EPLAN, Duke Energy Common Emergency Plan, Revision 5, Summary of Changes2023-11-0101 November 2023 Procedure EP-ALL-EPLAN, Duke Energy Common Emergency Plan, Revision 5, Summary of Changes RA-23-0275, Subsequent License Renewal Application, Appendix E, Responses to Requests for Additional Information (Rai), and Request for Confirmation of Information (RCI)2023-10-12012 October 2023 Subsequent License Renewal Application, Appendix E, Responses to Requests for Additional Information (Rai), and Request for Confirmation of Information (RCI) RA-23-0121, License Amendment Request to Adopt TSTF-258-A, Revision 4, Regarding Changes to Technical Specification Section 5.7, High Radiation Area2023-10-0505 October 2023 License Amendment Request to Adopt TSTF-258-A, Revision 4, Regarding Changes to Technical Specification Section 5.7, High Radiation Area RA-23-0268, Transmittal of Sixth 10-Year Inservice Testing Interval Program Plan2023-09-29029 September 2023 Transmittal of Sixth 10-Year Inservice Testing Interval Program Plan RA-23-0234, Update to the Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations for Oconee Nuclear Station (ONS)2023-09-28028 September 2023 Update to the Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations for Oconee Nuclear Station (ONS) RA-23-0230, Refuel 27 (C1 R27) Inservice Inspection (ISI) Report2023-09-25025 September 2023 Refuel 27 (C1 R27) Inservice Inspection (ISI) Report RA-23-0218, Review Request for the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals2023-09-21021 September 2023 Review Request for the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals RA-23-0225, Procedure AD-EP-ALL-0109, Offsite Protective Action Recommendations, Revision 9, and the Joint Information Center (JIC) Relocation, Summary of Changes2023-09-20020 September 2023 Procedure AD-EP-ALL-0109, Offsite Protective Action Recommendations, Revision 9, and the Joint Information Center (JIC) Relocation, Summary of Changes RA-23-0228, Registration for Use of General License Spent Fuel Cask Number 1752023-09-14014 September 2023 Registration for Use of General License Spent Fuel Cask Number 175 RA-22-0290, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology2023-08-30030 August 2023 License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology RA-23-0215, Independent Spent Fuel Storage Installation (ISFSI) ISFSI Docket Number 72-45 Registration of Spent Fuel Storage Casks2023-08-28028 August 2023 Independent Spent Fuel Storage Installation (ISFSI) ISFSI Docket Number 72-45 Registration of Spent Fuel Storage Casks RA-23-0223, Registration for Use of General License Spent Fuel Cask Numbers 173 and 1742023-08-23023 August 2023 Registration for Use of General License Spent Fuel Cask Numbers 173 and 174 RA-23-0222, On-Shift Staffing Analysis (Ossa), Revision 12023-08-23023 August 2023 On-Shift Staffing Analysis (Ossa), Revision 1 RA-23-0216, Independent Spent Fuel Storage Installation, Registration of Use of Spent Fuel Casks2023-08-22022 August 2023 Independent Spent Fuel Storage Installation, Registration of Use of Spent Fuel Casks RA-23-0199, Response to Request for Additional Information (RAI) Regarding Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Inservice Inspection of the Torus Metallic Liner2023-08-18018 August 2023 Response to Request for Additional Information (RAI) Regarding Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Inservice Inspection of the Torus Metallic Liner RA-23-0211, Reply to a Notice of Violation (NOV) 05000270/2023010-022023-08-17017 August 2023 Reply to a Notice of Violation (NOV) 05000270/2023010-02 RA-23-0122, License Amendment Request to Revise the 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Categorization Process to .2023-08-17017 August 2023 License Amendment Request to Revise the 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Categorization Process to . RA-23-0154, Duke Energy - Response to Request for Additional Information Regarding Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2023-07-20020 July 2023 Duke Energy - Response to Request for Additional Information Regarding Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) RA-23-0136, Duke Energy Response to Request for Additional Information Regarding Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)2023-07-20020 July 2023 Duke Energy Response to Request for Additional Information Regarding Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) RA-23-0195, Notice of Intentions Regarding Apparent Violation Response; (EA-23-060)2023-07-20020 July 2023 Notice of Intentions Regarding Apparent Violation Response; (EA-23-060) RA-23-0186, Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 2 Cycle 31 (Revision 3)2023-07-18018 July 2023 Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 2 Cycle 31 (Revision 3) RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0184, Registration for Use of General License Spent Fuel Cask Numbers 171 and 1722023-07-10010 July 2023 Registration for Use of General License Spent Fuel Cask Numbers 171 and 172 RA-23-0142, Mcguire, Units 1 and 2, Oconee, Units 1, 2, and 3, H. B. Robinson, Unit 2, Response to Request for Additional Information Regarding the License Amendment Request to Revise Restrictive Technical Specification Surveillance Require2023-07-0707 July 2023 Mcguire, Units 1 and 2, Oconee, Units 1, 2, and 3, H. B. Robinson, Unit 2, Response to Request for Additional Information Regarding the License Amendment Request to Revise Restrictive Technical Specification Surveillance Require RA-23-0127, 10 CFR 50.55a(g)(4)(iv) Request to Use a Subsequent Edition of ASME Section XI Referenced in 10 CFR 50.SSa(a) for ISI Code of Record2023-06-28028 June 2023 10 CFR 50.55a(g)(4)(iv) Request to Use a Subsequent Edition of ASME Section XI Referenced in 10 CFR 50.SSa(a) for ISI Code of Record RA-23-0045, Inservice Inspection Program Owner'S Activity Report for Unit 2 Refueling Outage 262023-06-28028 June 2023 Inservice Inspection Program Owner'S Activity Report for Unit 2 Refueling Outage 26 RA-23-0146, Subsequent License Renewal Application, Appendix E, Responses to Requests for Additional Information (Rai), and Request for Confirmation of Information (RCI)2023-06-20020 June 2023 Subsequent License Renewal Application, Appendix E, Responses to Requests for Additional Information (Rai), and Request for Confirmation of Information (RCI) RA-18-0007, License Amendment Application to Revise Control Room Cooling Technical Specifications2023-06-19019 June 2023 License Amendment Application to Revise Control Room Cooling Technical Specifications RA-23-0145, Independent Spent Fuel Storage Installation (Isfsi), Registration for Use of General License Spent Fuel Casks2023-06-15015 June 2023 Independent Spent Fuel Storage Installation (Isfsi), Registration for Use of General License Spent Fuel Casks RA-23-0135, Submittal of 30 Day Report Per 10 CFR 26.719(c), Unsatisfactory Performance of Health and Human Services Certified Laboratory2023-06-0707 June 2023 Submittal of 30 Day Report Per 10 CFR 26.719(c), Unsatisfactory Performance of Health and Human Services Certified Laboratory RA-23-0139, Registration for Use of General License Spent Fuel Casks2023-06-0606 June 2023 Registration for Use of General License Spent Fuel Casks RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0005, License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications2023-05-31031 May 2023 License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications RA-23-0041, Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations2023-05-30030 May 2023 Response to NRC Regulatory Issue Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations 2024-02-05
[Table view] Category:Report
MONTHYEARRA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation ML23270B8362023-09-26026 September 2023 Code Case N-752 Audit September 26, 2023, E-mail Providing Additional Information Regarding the Use of Owner'S Requirements and Engineering Judgment in Lieu of Code and Standards RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023 Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) ML22269A3882022-09-0707 September 2022 02-Gamma Spectroscopy of Concrete Pucks RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022 Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary) RA-22-0180, Supplement to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant2022-07-0707 July 2022 Supplement to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant RA-22-0165, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 242022-06-0909 June 2022 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 24 RA-22-0134, Unit 1 Cycle 23 Mellla+ Eigenvalue Tracking Data2022-05-25025 May 2022 Unit 1 Cycle 23 Mellla+ Eigenvalue Tracking Data RA-22-0115, Request for Acceptance of Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue (EAF) for the Safety Injection Nozzle2022-04-21021 April 2022 Request for Acceptance of Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue (EAF) for the Safety Injection Nozzle RA-21-0144, Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant2022-01-20020 January 2022 Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-21-0312, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2021-11-22022 November 2021 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) RA-19-0245, of DPC-NE-1007-P, Conditional Exemption of the EOC Mtc Measurement Methodology2021-10-25025 October 2021 of DPC-NE-1007-P, Conditional Exemption of the EOC Mtc Measurement Methodology IR 05000261/20210052021-08-25025 August 2021 Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005) RA-21-0198, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 252021-06-21021 June 2021 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 25 RA-21-0176, Cycle 24 Mellla+ Eigenvalue Tracking Data2021-06-0707 June 2021 Cycle 24 Mellla+ Eigenvalue Tracking Data RA-21-0145, Revision to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method2021-04-24024 April 2021 Revision to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method RA-21-0133, Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method2021-04-23023 April 2021 Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits2021-03-15015 March 2021 Notification of Permit Revision Request Regarding Copper and Zinc Limits RA-20-0353, Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1)2021-02-24024 February 2021 Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1) RA-20-0381, CFR 50.54( Q) Screening Evaluation Form for Revisions2020-12-0808 December 2020 CFR 50.54( Q) Screening Evaluation Form for Revisions RA-20-0363, Post Accident Monitoring Report 369-2020-012020-12-0101 December 2020 Post Accident Monitoring Report 369-2020-01 RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in . RA-20-0347, 10 CFR 71.95 Report on the 3-60B Cask2020-11-16016 November 2020 10 CFR 71.95 Report on the 3-60B Cask RA-20-0280, Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw2020-09-14014 September 2020 Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw ML20150A3322020-04-23023 April 2020 Revision of Slc'S 16-11-2, 16-11-7 and Loes RA-20-0032, License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits2020-03-0606 March 2020 License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits RA-19-0479, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-12-31031 December 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0403, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z2019-10-23023 October 2019 Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z RA-19-0411, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-10-23023 October 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies ML19242C7282019-08-28028 August 2019 Attachments 5 - 12: Thermal-Hydraulic Models for High Energy Line Break Transient Analysis, Duke Energy/Framatome Affidavits, Regulatory Requirements, Definitions, Time Critical Operator Actions and Feasibility Assessment for New Proposed T RA-19-0338, Pressure and Temperature Limits Report for Unit Nos. 1 and 22019-08-15015 August 2019 Pressure and Temperature Limits Report for Unit Nos. 1 and 2 RA-19-0243, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-07-0202 July 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 RA-19-0240, Response to Request for Additional Information Regarding Advanced Framatome Methodologies License Amendment Request2019-05-29029 May 2019 Response to Request for Additional Information Regarding Advanced Framatome Methodologies License Amendment Request RA-19-0139, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-03-18018 March 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies ML19058A4162019-03-0808 March 2019 Staff Evaluation Related to Reactor Vessel Internals Inspection Plan Based on MRP-227-A RA-18-0100, Request for License Amendment Regarding Application of Advanced Framatome Methodologies2018-10-11011 October 2018 Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-18-0179, 90-Day Special Report2018-10-0404 October 2018 90-Day Special Report ML18230A7592018-08-18018 August 2018 Fault Investigation, Progress Report, Volume 1 of 2 ML18230A7562018-08-18018 August 2018 Fault Investigation, Responses to Mr. W. R. Butler'S Letter of May 16, 1975 RA-18-0131, Technical Requirements Manual, Revision 662018-08-13013 August 2018 Technical Requirements Manual, Revision 66 ML18249A1592018-08-13013 August 2018 Technical Requirements Manual, Revision 73 RA-18-0024, Supplemental Information to Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report2018-05-29029 May 2018 Supplemental Information to Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report HNP-18-023, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes2018-05-0202 May 2018 Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes ML18135A0722018-04-30030 April 2018 ANP-3658, Revision 0, Evaluation of the McGuire Units 1 and 2 Upflow Modification for 60 Years. BSEP 18-0044, Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D2018-04-11011 April 2018 Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D 2023-09-26
[Table view] Category:Miscellaneous
MONTHYEARML23270B8362023-09-26026 September 2023 Code Case N-752 Audit September 26, 2023, E-mail Providing Additional Information Regarding the Use of Owner'S Requirements and Engineering Judgment in Lieu of Code and Standards RA-22-0165, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 242022-06-0909 June 2022 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 24 RA-22-0115, Request for Acceptance of Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue (EAF) for the Safety Injection Nozzle2022-04-21021 April 2022 Request for Acceptance of Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue (EAF) for the Safety Injection Nozzle RA-21-0312, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2021-11-22022 November 2021 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) IR 05000261/20210052021-08-25025 August 2021 Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005) RA-21-0198, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 252021-06-21021 June 2021 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 25 RA-20-0353, Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1)2021-02-24024 February 2021 Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1) RA-19-0479, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-12-31031 December 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies ML19242C7282019-08-28028 August 2019 Attachments 5 - 12: Thermal-Hydraulic Models for High Energy Line Break Transient Analysis, Duke Energy/Framatome Affidavits, Regulatory Requirements, Definitions, Time Critical Operator Actions and Feasibility Assessment for New Proposed T RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 ML19058A4162019-03-0808 March 2019 Staff Evaluation Related to Reactor Vessel Internals Inspection Plan Based on MRP-227-A ML18230A7562018-08-18018 August 2018 Fault Investigation, Responses to Mr. W. R. Butler'S Letter of May 16, 1975 ML18230A7592018-08-18018 August 2018 Fault Investigation, Progress Report, Volume 1 of 2 HNP-18-023, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes2018-05-0202 May 2018 Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes BSEP 18-0044, Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D2018-04-11011 April 2018 Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D RNP-RA/18-0024, Report of Changes Pursuant to 10 CFR 50.59(d)(2)2018-04-0202 April 2018 Report of Changes Pursuant to 10 CFR 50.59(d)(2) MNS-17-050, Review Request for the Aging Management Program and Inspection Plan for the Reactor Vessel Internals to Implement MRP-227-A2017-12-13013 December 2017 Review Request for the Aging Management Program and Inspection Plan for the Reactor Vessel Internals to Implement MRP-227-A CNS-17-043, Report 16C4437-RPT-002, Revision 0, 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation for Catawba Nuclear Station.2017-07-0707 July 2017 Report 16C4437-RPT-002, Revision 0, 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation for Catawba Nuclear Station. ML17171A2772017-06-12012 June 2017 Selected Licensee Commitments List of Revised Effected Sections CNS-17-022, Summary of 10CFR 50.54(q) Evaluation2017-04-25025 April 2017 Summary of 10CFR 50.54(q) Evaluation HNP-17-006, Submittal of Cycle 20 Activity Report2017-01-30030 January 2017 Submittal of Cycle 20 Activity Report ML17031A4312017-01-26026 January 2017 Notification of Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events and FLEX Final Integrated Plan ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 HNP-16-083, Mitigating Strategies Assessment Report for Flooding Hazard Information2016-12-21021 December 2016 Mitigating Strategies Assessment Report for Flooding Hazard Information ML16281A5102016-12-15015 December 2016 Staff Assessment of the Reactor Vessel Internals Aging Management Program Plans RNP-RA/16-0087, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-31031 October 2016 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors ML16293A6662016-10-31031 October 2016 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood Causing Mechanism Reevaluation (CAC Nos. MF3623 and MF3624) ML16294A2532016-10-12012 October 2016 Issue MNS Technical Specification 5.6 Rev. 2 Amend. 288/267 RNP-RA/16-0078, Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-0505 October 2016 Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors RA-16-0024, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P2016-10-0303 October 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P ML16237A3542016-08-31031 August 2016 Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 ML16223A7252016-08-17017 August 2016 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Phase 2 of Order EA-13-109 (Severe Accident Capable Hardened Events) CAC Nos. MF4467 and MF4468) ML16152A0522016-05-31031 May 2016 ANP-3477NP, Rev. O, MRP-227-A Applicant/Licensee Action Item 6 Analysis RNP-RA/16-0038, Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-1742016-05-25025 May 2016 Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-174 ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report ML16041A4352016-03-0101 March 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force HNP-16-017, Transmittal of Summary of a 10 CFR 50.54(q) Evaluation2016-02-29029 February 2016 Transmittal of Summary of a 10 CFR 50.54(q) Evaluation ONS-2016-017, Request for Alternative to Codes and Standards Requirements Pursuant to 10 CFR 50.55a(z) to Satisfy 10 CFR 50.55a(h)(2)2016-02-15015 February 2016 Request for Alternative to Codes and Standards Requirements Pursuant to 10 CFR 50.55a(z) to Satisfy 10 CFR 50.55a(h)(2) CNS-15-101, License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants2015-12-11011 December 2015 License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants RA-15-0047, Annual Report of Changes Pursuant to 10 CFR 50.462015-11-17017 November 2015 Annual Report of Changes Pursuant to 10 CFR 50.46 ML15301A5572015-11-0202 November 2015 Supplement to Staff Assessment of Response to 10 CFR 50.54(f) Information Request Flood Causing Mechanisms Reevaluations ML15280A1992015-10-19019 October 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review ML15313A1642015-10-0808 October 2015 Submittal of 10 CFR 71.95 Report on the 8-120B Cask MNS-15-072, Selected Licensee Commitment Special Report 2015-01, Nuclear Condition Report No. 1943414, Standby Shutdown System Was Non-functional for a Period Greater than 7 Days2015-09-10010 September 2015 Selected Licensee Commitment Special Report 2015-01, Nuclear Condition Report No. 1943414, Standby Shutdown System Was Non-functional for a Period Greater than 7 Days ML15238A7392015-08-24024 August 2015 Submittal of 10 CFR 71.95 Report on the 8-120B Cask RNP-RA/15-0053, Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04)2015-08-19019 August 2015 Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04) ML15246A1452015-08-11011 August 2015 Submittal of 10 CFR 71.95 Report on the 8-120B Cask ML15218A2972015-08-0606 August 2015 Historic Temperature Data Including Record Highs for the Seneca ML15201A0082015-07-22022 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review ML15182A0672015-07-20020 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(F), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near Term Task Force Review of Insights From.. 2023-09-26
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Steve Snider Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 Steve.Snider@duke-energy.com 10 CFR 50.46 May 30, 2019 Serial: RA-19-0223 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Catawba Nuclear Station, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. NPF-35 and NPF-52 Docket Nos. 50-413 and 50-414 H. B. Robinson Steam Electric Plant, Unit 2 Renewed Facility Operating License No. DPR-23 Docket No. 50-261 McGuire Nuclear Station, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. NPF-9 and NPF-17 Docket Nos. 50-369 and 50-370 Shearon Harris Nuclear Power Plant, Unit 1 Renewed Facility Operating License No. NPF-63 Docket No. 50-400 Oconee Nuclear Station, Unit Nos. 1, 2 and 3 Renewed Facility Operating License Nos. DPR-38, DPR-47 and DPR-55 Docket Nos. 50-269, 50-270 and 50-287
Subject:
Annual Report of Changes Pursuant to 10 CFR 50.46 Ladies and Gentlemen:
Pursuant to 10 CFR 50.46(a)(3)(ii), Duke Energy hereby submits the enclosed annual reports of changes to or errors in Emergency Core Cooling System (ECCS) evaluation models. These reports cover the period from January 1, 2018 to December 31, 2018 for the Brunswick Steam Electric Plant, Catawba Nuclear Station, H. B. Robinson Steam Electric Plant, McGuire Nuclear Station, Shearon Harris Nuclear Power Plant and the Oconee Nuclear Station.
There are no regulatory commitments contained in this letter.
U.S. Nuclear Regulatory Commission Page 2 Serial: RA-19-0223 Should you have any questio ns concerning this letter and its enclosu res, please contact Art Zaremba, Manager - Nuclea r Fleet Licensing at (980) 373-2062.
Sincerely, Steve Snider Vice President - Nuclear Engineering JLV
Enclosures:
- 1. Brunswick 10 CFR 50.46 Annual Report
- 2. Catawba 10 CFR 50.46 Annual Report
- 3. Robinson 10 CFR 50.46 Annual Report
- 4. McGuire 10 CFR 50.46 Annual Report
- 5. Shearon Harris 10 CFR 50.46 Annual Report
- 6. Oconee 10 CFR 50.46 Annual Report cc: Ms. C. Haney, NRC Regional Administrator, Region II Mr. D. J. Galvin, NRC Project Manager, BNP Mr. G. Smith, NRC Sr. Resident Inspector, BNP Mr. M. Mahoney, NRC Project Manager, CNS and MNS Mr. J. D. Austin, NRC Sr. Resident Inspector, CNS Mr. A. Hutto, NRC Sr. Resident Inspector, MNS Mr. N. Jordan, NRC Project Manager, HBRSEP2 Mr. M. Fannon, NRC Sr. Resident Inspector, HBRSE P2 Ms. M. Barillas, NRC Project Manager, HNP Mr. J. Zeiler, NRC Sr. Resident Inspector, HNP Ms. A. Klett, NRC Senior Project Manager, ONS Mr. E. L. Crowe, NRC Sr. Resident Inspector, ONS Chair - North Carolina Utilities Commission (Electronic Copy Only)
Ms. L. Garner, Manager, SC DHEC Ms. A. Nair-Gimmi, SC DHEC Mr. A. Wilson, Attorney General (SC)
U.S. Nuclear Regulatory Commission Page 3 Serial: RA-19-0223 bcc (with Enclosure):
W.R. Gideon K.K. Moser J. Ratliff J.L. Pierce T. Simril C.E. Curry A.B. Linker M.B. Hare E. Kapopoulos J.A. Krakuszeski B.J. Foster K.M. Ellis T. Ray E.R. Pigott N.E. Kunkel J. Thomas T.M. Hamilton J. Dills C. Kidd B.C. McCabe E. Burchfield P.V. Fisk T. Grant S.A. Dalton S. Snider M.C. Nolan A.H. Zaremba S.B. Thomas M.C. Handrick D.A. Cummings ELL File: (Corporate)
U.S. Nuclear Regulatory Commission Page 1 of 3 Serial: RA-19-0223 Serial: RA-19-0223 Brunswick Steam Electric Plant, Units 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 Enclosure 1 Brunswick 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-19-0223 A10XM Summary 10 CFR 50.46 Report for Brunswick Steam Electric Plant Units 1, and 2 Plant: Brunswick Steam Electric Plant, Units 1 and 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable):
Evaluation Model: EMF-2361(P)(A), Revision 0 EXEM BWR-2000 ECCS Evaluation Model, May 2001 Fuel: ATRIUM 10XM (A10XM)
A. Analysis of Record PCT 1923 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported 0 °F 0 °F C. Baseline PCT for assessing new changes for significance (A + B) 1923 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Error in the Thermal Conductivity 0 °F Degradation analysis for top lattices with a top peaked axial power shape. BNPs limiting lattice is a bottom lattice with a mid- peaked axial power shape -
no impact. (Framatome Report FS1-0040060 Rev. 1.0).
E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1923 °F
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-19-0223 ATRIUM 11 Summary 10 CFR 50.46 Report for Brunswick Steam Electric Plant Units 1, and 2 Plant: Brunswick Steam Electric Plant, Unit 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable):
Evaluation Model: EMF-2361(P)(A), Revision 0 EXEM BWR-2000 ECCS Evaluation Model, May 2001 Fuel: ATRIUM 11 (A11)
A. Analysis of Record PCT 1762 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported 0 °F 0 °F C. Baseline PCT for assessing new changes for significance (A + B) 1762 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Error in the Thermal Conductivity 0 °F Degradation analysis for top lattices with a top peaked axial power shape. BNPs limiting lattice is a bottom lattice with a mid- peaked axial power shape -
no impact. (Framatome Report FS1-0041575 Rev. 1.0).
E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1762 °F
U.S. Nuclear Regulatory Commission Page 1 of 6 Serial: RA-19-0223 Serial: RA-19-0223 Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 / Renewed License Nos. NPF-35 and NPF-52 Enclosure 2 Catawba 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 6 Serial: RA-19-0223 Westinghouse identified and communicated in a letter that there were modeling changes and errors in the LOCA evaluation models that were assessed for impact to PCT in 2018. None of the assessments resulted in changes to PCT. The following items are included for information.
VAPOR TEMPERATURE RESETTING Affected Evaluation Model(s):
1996 Westinghouse Best Estimate Large Break LOCA In the WCOBRA/TRAC and WCOBRA/TRAC-TF2 codes, when the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. It was discovered that this vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. The correction of this error represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Engineering judgement supported by sensitivity calculations showed that correcting this error had minimal impact on LOCA transient calculations, leading to an estimated peak cladding temperature impact of 0 °F.
UO2 Fuel Pellet Heat Capacity Affected Evaluation Model: 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP A typographical error was discovered in the implementation of the UO2 fuel pellet heat capacity as described by Equation C-4 of WCAP-8301 [1] for fuel rod heat-up calculations within the Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an overprediction of heat capacity that increases with fuel temperature. The corrected formulation results in a maximum decrease in heat capacity on the order of approximately 1.2%
for existing analyses of record. This represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
The small over-prediction in UO2 fuel pellet heat capacity has been evaluated to have a negligible effect on existing large and small break LOCA analysis results due to the small magnitude of the change, leading to an estimated PCT impact of 0 °F.
Reference
- 1) WCAP-8301, LOCTA-IV Program: Loss-of-Coolant Transient Analysis, June 1974.
U.S. Nuclear Regulatory Commission Page 3 of 6 Serial: RA-19-0223 10 CFR 50.46 Report for Catawba Unit 1 - Large Break LOCA Plant: Catawba Nuclear Station, Unit 1 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +58 °F 378 °F C. Baseline PCT for assessing new changes for significance (A + B) 2086 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Vapor Temperature Resetting 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 2086 °F
U.S. Nuclear Regulatory Commission Page 4 of 6 Serial: RA-19-0223 10 CFR 50.46 Report for Catawba Unit 1 - Small Break LOCA Plant: Catawba Nuclear Station, Unit 1 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA A. Analysis of Record PCT 1323 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new changes for significance (A + B) 1323 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. UO2 Fuel Pellet Heat Capacity 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1323 °F
U.S. Nuclear Regulatory Commission Page 5 of 6 Serial: RA-19-0223 10 CFR 50.46 Report for Catawba Unit 2 - Large Break LOCA Plant: Catawba Nuclear Station, Unit 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +42 °F 362 °F C. Baseline PCT for assessing new changes for significance (A + B) 2070 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Vapor Temperature Resetting 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 2070 °F
U.S. Nuclear Regulatory Commission Page 6 of 6 Serial: RA-19-0223 10 CFR 50.46 Report for Catawba Unit 2 - Small Break LOCA Plant: Catawba Nuclear Station, Unit 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA A. Analysis of Record PCT 1243 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new changes for significance (A + B) 1243 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. UO2 Fuel Pellet Heat Capacity 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1243 °F
U.S. Nuclear Regulatory Commission Page 1 of 3 Serial: RA-19-0223 Serial: RA-19-0223 H. B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 / Renewed License No. DPR-23 Enclosure 3 Robinson 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for H. B. Robinson Unit 2 - Large Break LOCA Plant: H. B. Robinson , Unit 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if Large Break applicable):
Evaluation Model: EMF-2103(P)(A), Revision 0 Realistic Large Break LOCA for PWRs Fuel: 15x15 HTP A. Analysis of Record PCT 2084 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +4 °F 24 °F C. Baseline PCT for assessing new changes for significance (A + B) 2088 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Estimated effect of including a +31 °F fuel clad swelling and rupture model, inclusive of (1) M5 LOCA swelling and rupture model update and (2) error corrections to cladding oxidation calculation due to use of cold cladding dimensions.
E. Sum of 10 CFR 50.46 Changes Net PCT Effect Absolute PCT Effect and Error Corrections against Baseline PCT +31 °F 31 °F F. Licensing Basis PCT (C + E) 2119 °F
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for H. B. Robinson Unit 2 - Small Break LOCA Plant: H. B. Robinson , Unit 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if Small Break applicable):
Evaluation Model: EMF-2328(P)(A), Revision 0 PWR Small Break LOCA Evaluation Model Fuel: 15x15 HTP A. Analysis of Record PCT 1492 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +60 °F 98 °F C. Baseline PCT for assessing new changes for significance (A + B) 1552 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes Net PCT Effect Absolute PCT Effect and Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1552 °F
U.S. Nuclear Regulatory Commission Page 1 of 4 Serial: RA-19-0223 Serial: RA-19-0223 McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 / Renewed License Nos. NPF-9 and NPF-17 Enclosure 4 McGuire 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 4 Serial: RA-19-0223 Westinghouse identified and communicated in a letter that there were modeling changes and errors in the LOCA evaluation models that were assessed for impact to PCT in 2018. None of the assessments resulted in changes to PCT. The following items are included for information.
VAPOR TEMPERATURE RESETTING Affected Evaluation Model(s):
1996 Westinghouse Best Estimate Large Break LOCA In the WCOBRA/TRAC and WCOBRA/TRAC-TF2 codes, when the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. It was discovered that this vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. The correction of this error represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Engineering judgement supported by sensitivity calculations showed that correcting this error had minimal impact on LOCA transient calculations, leading to an estimated peak cladding temperature impact of 0 °F.
UO2 Fuel Pellet Heat Capacity Affected Evaluation Model: 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP A typographical error was discovered in the implementation of the UO2 fuel pellet heat capacity as described by Equation C-4 of WCAP-8301 [1] for fuel rod heat-up calculations within the Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an overprediction of heat capacity that increases with fuel temperature. The corrected formulation results in a maximum decrease in heat capacity on the order of approximately 1.2%
for existing analyses of record. This represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
The small over-prediction in UO2 fuel pellet heat capacity has been evaluated to have a negligible effect on existing large and small break LOCA analysis results due to the small magnitude of the change, leading to an estimated PCT impact of 0 °F.
Reference
- 1) WCAP-8301, LOCTA-IV Program: Loss-of-Coolant Transient Analysis, June 1974.
U.S. Nuclear Regulatory Commission Page 3 of 4 Serial: RA-19-0223 10 CFR 50.46 Report for McGuire Units 1 & 2 - Large Break LOCA Plant: McGuire Nuclear Station, Units 1 & 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +58 °F 378 °F C. Baseline PCT for assessing new changes for significance (A + B) 2086 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. Vapor Temperature Resetting 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 2086 °F
U.S. Nuclear Regulatory Commission Page 4 of 4 Serial: RA-19-0223 10 CFR 50.46 Report for McGuire Units 1 & 2 - Small Break LOCA Plant: McGuire Nuclear Station, Units 1 & 2 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA A. Analysis of Record PCT 1323 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new changes for significance (A + B) 1323 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. UO2 Fuel Pellet Heat Capacity 0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1323 °F
U.S. Nuclear Regulatory Commission Page 1 of 3 Serial: RA-19-0223 Serial: RA-19-0223 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 Enclosure 5 Shearon Harris 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for Shearon Harris Unit 1 - Large Break LOCA Plant: Shearon Harris, Unit 1 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if Large Break applicable):
Evaluation Model: EMF-2103(P)(A), Revision 0 Realistic Large Break LOCA for PWRs Fuel: 17x17 HTP A. Analysis of Record PCT 1935 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +160 °F 160 °F C. Baseline PCT for assessing new changes for significance (A + B) 2095 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes Net PCT Effect Absolute PCT Effect and Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 2095 °F
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for Shearon Harris Unit 1 - Small Break LOCA Plant: Shearon Harris, Unit 1 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if Small Break applicable):
Evaluation Model: EMF-2328(P)(A), Revision 0 PWR Small Break LOCA Evaluation Model Fuel: 17x17 HTP A. Analysis of Record PCT 1664 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +63 °F 63 °F C. Baseline PCT for assessing new changes for significance (A + B) 1727 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes Net PCT Effect Absolute PCT Effect and Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1727 °F
U.S. Nuclear Regulatory Commission Page 1 of 3 Serial: RA-19-0223 Serial: RA-19-0223 Oconee Nuclear Station, Units 1, 2 and 3 Docket Nos. 50-269, 50-270 and 50-287 Renewed License Nos. DPR-38, DPR-47 and DPR-55 Enclosure 6 Oconee 10 CFR 50.46 Annual Report
U.S. Nuclear Regulatory Commission Page 2 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for Oconee Units 1, 2, & 3 - Large Break LOCA Plant: Oconee Nuclear Station, Units 1, 2, & 3 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Large Break Evaluation Model: BAW-10192P-A, Revision 0, BWNT LOCA Evaluation Model for Once-Through Steam Generator Plants Fuel: 15x15 Mark-B-HTP A. Analysis of Record PCT 1852 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +2 °F 858 °F C. Baseline PCT for assessing new changes for significance (A + B) 1854 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1854 °F
U.S. Nuclear Regulatory Commission Page 3 of 3 Serial: RA-19-0223 10 CFR 50.46 Report for Oconee Units 1, 2, & 3 - Small Break LOCA Plant: Oconee Nuclear Station, Units 1, 2, & 3 Reporting Period: January 1, 2018 - December 31, 2018 LOCA Analysis Type (if applicable): Small Break Evaluation Model: BAW-10192P-A, Revision 0, BWNT LOCA Evaluation Model for Once-Through Steam Generator Plants Fuel: 15x15 Mark-B-HTP A. Analysis of Record PCT 1598 °F Full Power (FP) - 100% FP B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +0 °F 0°F C. Baseline PCT for assessing new changes for significance (A + B) 1598 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1598 °F A. Analysis of Record PCT 1480 °F Reduced Power - 50% FP B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections
- Previously Reported +0 °F 0°F C. Baseline PCT for assessing new changes for significance (A + B) 1480 °F D. Cumulative 10 CFR 50.46 Changes and Error Corrections
- This Reporting Period
- 1. None E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline PCT 0 °F 0 °F F. Licensing Basis PCT (C + E) 1480 °F