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Category:Code Relief or Alternative
MONTHYEARML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23230A0652023-08-31031 August 2023 William B. McGuire Nuclear Station, Units 1 and 2 - Relief Request Use of Later Edition of ASME Code ML23159A2712023-06-20020 June 2023 William B. McGuire Nuclear Station, Unit 1 - Relief Request Impractical Reactor System Welds ML23151A3482023-05-30030 May 2023 Duke Fleet - Request for Additional Information Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML23118A0762023-05-0101 May 2023 Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-22-0257, Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2023-02-17017 February 2023 Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML22096A0032022-11-18018 November 2022 McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations ML22266A0782022-09-26026 September 2022 William B. Mcguire Nuclear Station, Unit 2 Pressurizer Power Operated Relief Valve Relief Request ML22213A2532022-08-31031 August 2022 Authorization and Safety Evaluation for Alternative Request RA-21-0144 to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method ML22242A1602022-08-30030 August 2022 Hardship Relief Request Pilot-Operated Relief Valve Acceptance Review Results ML22046A2802022-02-23023 February 2022 Relief from ASME Code Paragraph ISTP-3540(b) Related to High Pressure Injection Pumps Vibration Measurements ML22046A2452022-02-18018 February 2022 Regarding Alternative to Implement Code Case OMN-22, Smooth Running Pumps ML22028A3652022-02-0707 February 2022 Authorization and Safety Evaluation for Alternative Acceptance Criteria for Code Case N-853 ML22007A3452022-01-11011 January 2022 Proposed Alternative to Use the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-885 ML21306A1592021-11-16016 November 2021 Request to Use Provisions of Later Edition and Addenda of the ASME Code, Section XI (EPID; L-2021-LLR-0048) ML21259A0992021-11-0404 November 2021 Proposed Alternatives to American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants ML21214A2222021-08-10010 August 2021 Proposed Alternatives to American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants ML21209A9962021-08-0202 August 2021 Proposed Alternative Request IST-RR-8 Related to the Inservice Testing Program ML21202A3982021-07-27027 July 2021 Relief Request Proposed Alternatives to American Society of Mechanical Engineers Operation and Maintenance Code Isolation Valve Seal Water System ML21068A4172021-03-17017 March 2021 Summary of Meeting with Duke Energy Progress, LLC, Regarding a Proposed Relief Request Concerning Containment Inspections of Liners and Moisture Barriers for the H. B. Robinson Steam Electric Plant, Unit 2 ML21029A3352021-02-16016 February 2021 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-20-0312, Relief Request, Service Water Pinhole Leak in a Socket Weld, Inservice Inspection Program, Fourth Ten-Year Interval2020-11-0404 November 2020 Relief Request, Service Water Pinhole Leak in a Socket Weld, Inservice Inspection Program, Fourth Ten-Year Interval ML20230A2052020-08-21021 August 2020 Relief Request RA-20-0031, Delay to Update the Code of Record for Inservice Inspection ML20090F8932020-04-29029 April 2020 Relief Request RA-19-0138 Regarding Proposed Alternative to ASME Code, Section XI Volumetric Examination Frequency Requirements ML20080G9502020-04-0505 April 2020 Relief from the Requirements of the ASME Code RA-20-0036, Request for Alternative to Defect Removal Prior to Performing Repair/Replacement Activities on Low Pressure Service Water (LPSW) System Piping2020-03-0202 March 2020 Request for Alternative to Defect Removal Prior to Performing Repair/Replacement Activities on Low Pressure Service Water (LPSW) System Piping ML20042D7142020-02-13013 February 2020 Relief Request 19-ON-002 for Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in ASME Code Case N-770-2 RA-19-0428, Relief Request RA-19-0428 for One-Time Pressure Relief Valve Test Frequency Extension2020-02-0404 February 2020 Relief Request RA-19-0428 for One-Time Pressure Relief Valve Test Frequency Extension ML19217A3242019-08-14014 August 2019 Relief Request MC-SRV-NC-03, Alternate Testing for Pressurizer Power Operated Relief Valve Block Valve 2NC-35B ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins RA-19-0028, Submittal of Relief Request (19-ON-001) to Utilize Code Case N-853, PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material Susceptible to Primary Water Stress2019-03-21021 March 2019 Submittal of Relief Request (19-ON-001) to Utilize Code Case N-853, PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material Susceptible to Primary Water Stress ML19018A0252019-02-26026 February 2019 Relief Request I3R-18, Regarding Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval RA-19-0026, Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001)2019-02-11011 February 2019 Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001) ML18283B5442018-11-20020 November 2018 Relief from the Requirements of the ASME Code, Section XI, Reactor Vessel Closure Head Penetration Nozzle Repair Technique RA-18-0180, Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 22018-11-12012 November 2018 Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 ML18016A1782018-01-24024 January 2018 Request for Relief No. 17-CN-001, Regarding Category B-J Pressure Retaining Welds for the Third 10-Year Inservice Testing Interval (CAC Nos. MF9807 and MF9808; EPID L-2017-LLA-0032) ML17331A0862017-12-26026 December 2017 Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads ML17104A1102017-04-26026 April 2017 Relief Request CN Grr 01, Alternative to the Testing Frequencies in the American Society of Mechanical Engineers Operation and Maintenance Code, by Adoption of Code Case OMN-20, Inservice Test Frequencies (CAC Nos. MF9595-6) CNS-17-020, Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing2017-04-11011 April 2017 Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing ML17034A3622017-02-22022 February 2017 Mcguire Nuclear Station, Unit 1 - Request 16-MN-002, Alternative to Defect Removal Prior to Performing Repair Activities on Nuclear Service Water System Piping Relief ML16358A6962017-01-17017 January 2017 Proposed Relief Request MC-SRV-NC-02, Alternate Testing for Pressurizer Power Operated Relief Valve (Porv) Block Valve 2NC-31B ML16266A0262016-10-18018 October 2016 Proposed Relief Request Serial No. 16-MN-003 for Alternate Repair of Nuclear Service Water Piping ML16294A2542016-10-13013 October 2016 Response to Request for Additional Information Regarding Relief Request 16-MN-002, Alternative to Defect Removal Prior to Performing Repair Activities on Nuclear Service Water System Piping ML16225A6542016-08-11011 August 2016 Email - McGuire Unit No. 1: Acceptance REVIEW- Relief Request 16-MN-003 Alternative to Defect Removal Prior to Performing Temporary Repair Activities on Three-Inch-Diameter Nuclear Service Water System Piping. MNS-16-062, Relief Request 16-MN-003 Alternative to Defect Removal Prior to Performing Temporary Repair Activities on Three-Inch-Diameter Nuclear Service Water System Piping2016-08-10010 August 2016 Relief Request 16-MN-003 Alternative to Defect Removal Prior to Performing Temporary Repair Activities on Three-Inch-Diameter Nuclear Service Water System Piping ML16210A0382016-07-28028 July 2016 Draft Relief Request Serial No. 16-MN-003 MNS-16-053, Relief Request 16-MN-002, Alternative to Defect Removal Prior to Performing Repair Activities on Nuclear Service Water System Piping2016-06-23023 June 2016 Relief Request 16-MN-002, Alternative to Defect Removal Prior to Performing Repair Activities on Nuclear Service Water System Piping ML16152A6012016-06-0808 June 2016 Withdrawal of Relief Request ML16118A2302016-04-30030 April 2016 Relief Request 14-ON-022 Alternative Requirements for Class 2 Residual Heat Removal Heat Exchange Welds ML16053A5012016-03-0101 March 2016 Relief Request ON-GRR-01 Grace Period for OM Code Frequencies (CAC Nos. MF7130, MF7131, and MF7132) 2023-09-25
[Table view] 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-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-23-0211, 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-23-0154, 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-23-0146, 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-23-0135, 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-23-0120, 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 RA-23-0128, Refuel 32 (O1R32) Steam Generator Tube Inspection Report2023-05-18018 May 2023 Refuel 32 (O1R32) Steam Generator Tube Inspection Report RA-23-0018, Relief Request (RA-23-0018) to Utilize Code Case N-853 PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material Susceptible to Primary Water Stress Corrosion Cracking Section XI, Division 12023-05-0404 May 2023 Relief Request (RA-23-0018) to Utilize Code Case N-853 PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material Susceptible to Primary Water Stress Corrosion Cracking Section XI, Division 1 RA-23-0105, Duke Energy Carolinas, LLC (Duke Energy) - Core Operating Limits Report (COLR) for Unit 1 Cycle 28 Reload Core2023-05-0404 May 2023 Duke Energy Carolinas, LLC (Duke Energy) - Core Operating Limits Report (COLR) for Unit 1 Cycle 28 Reload Core RA-23-0118, Cycle 29, Revision 1, Core Operating Limits Report (COLR)2023-05-0303 May 2023 Cycle 29, Revision 1, Core Operating Limits Report (COLR) RA-23-0117, Report of Changes to Duke Energy McGuire Emergency Plan Annex2023-05-0101 May 2023 Report of Changes to Duke Energy McGuire Emergency Plan Annex RA-23-0111, Withdrawal of License Amendment Request to Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies2023-04-27027 April 2023 Withdrawal of License Amendment Request to Revise Technical Specification 5.5.2, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies 2024-02-05
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARRA-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-23-0275, 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-23-0136, 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-23-0154, 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-23-0142, 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-23-0146, 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-23-0104, Response to Request for Additional Information (RAI) Regarding Oconee Unit 3, Refuel 31 (O3R31) Steam Generator Tube Inspection Report2023-04-26026 April 2023 Response to Request for Additional Information (RAI) Regarding Oconee Unit 3, Refuel 31 (O3R31) Steam Generator Tube Inspection Report RA-23-0076, Supplement to Response to Request for Additional Information (RAI) Regarding McGuire Nuclear Station Unit 1 Spring, 2022 Outage Steam Generator Tube Inspection Report2023-03-16016 March 2023 Supplement to Response to Request for Additional Information (RAI) Regarding McGuire Nuclear Station Unit 1 Spring, 2022 Outage Steam Generator Tube Inspection Report RA-23-0051, Response to Request for Additional Information (RAI) Regarding Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities2023-03-0909 March 2023 Response to Request for Additional Information (RAI) Regarding Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities RA-23-0024, Response to Request for Additional Information (RAI) for Relief Request for RPV Reactor Coolant System Welds2023-02-28028 February 2023 Response to Request for Additional Information (RAI) for Relief Request for RPV Reactor Coolant System Welds RA-23-0025, End of Cycle 28 (M1R28) Steam Generator Tube Inspection Report Response to Request for Additional Information (RAI)2023-02-15015 February 2023 End of Cycle 28 (M1R28) Steam Generator Tube Inspection Report Response to Request for Additional Information (RAI) RA-23-0015, Response to Request for Additional Information (RAI) Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-02-0909 February 2023 Response to Request for Additional Information (RAI) Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-22-0270, Response to Request for Additional Information Regarding License Amendment Request to Address Technical Specifications Mode Change Limitations2022-10-0707 October 2022 Response to Request for Additional Information Regarding License Amendment Request to Address Technical Specifications Mode Change Limitations RA-22-0245, Response to Request for Additional Information (RAI) Regarding Removal of 4.160 Kilovolt Bus 2 from Surveillance Requirement 3.8.1.162022-09-0808 September 2022 Response to Request for Additional Information (RAI) Regarding Removal of 4.160 Kilovolt Bus 2 from Surveillance Requirement 3.8.1.16 RA-22-0192, Subsequent License Renewal Application Response to ONS SLRA Second Round RAI B2.1.7-4a2022-09-0202 September 2022 Subsequent License Renewal Application Response to ONS SLRA Second Round RAI B2.1.7-4a RA-22-0210, Supplement to License Amendment Request to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-07-28028 July 2022 Supplement to License Amendment Request to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-22-0160, Subsequent License Renewal Application: Responses to ONS SLRA - Second Round RAIs - Trp 76 (Irradiation Structural) - FE 3.5.2.2.2.62022-07-25025 July 2022 Subsequent License Renewal Application: Responses to ONS SLRA - Second Round RAIs - Trp 76 (Irradiation Structural) - FE 3.5.2.2.2.6 RA-22-0193, Subsequent License Renewal Application Response to ONS SLRA Second Round RAI B2.1.7-4b2022-07-0808 July 2022 Subsequent License Renewal Application Response to ONS SLRA Second Round RAI B2.1.7-4b RA-22-0195, Response to Request for Additional Information (RAI) for the Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue for the Safety Injection Nozzle2022-06-29029 June 2022 Response to Request for Additional Information (RAI) for the Proposed Method to Manage Aging Due to Environmentally Assisted Fatigue for the Safety Injection Nozzle RA-22-0158, Subsequent License Renewal Application - Response to ONS SLRA Second Round RAI B2.1.9-2a2022-06-0808 June 2022 Subsequent License Renewal Application - Response to ONS SLRA Second Round RAI B2.1.9-2a RA-22-0159, Subsequent License Renewal Application Response to ONS SLRA Request for Additional Information (RAI) 3.1.2-12022-05-27027 May 2022 Subsequent License Renewal Application Response to ONS SLRA Request for Additional Information (RAI) 3.1.2-1 RA-22-0137, Subsequent License Renewal Application Response to ONS SLRA Second Round RAI 4.6.1-1a2022-05-20020 May 2022 Subsequent License Renewal Application Response to ONS SLRA Second Round RAI 4.6.1-1a RA-22-0144, Response to NRC Request for Additional Information Regarding Supplemental Response to Generic Letter 2004-022022-05-19019 May 2022 Response to NRC Request for Additional Information Regarding Supplemental Response to Generic Letter 2004-02 RA-22-0147, Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, Response to Request for Additional Information (RAI) Regarding License Amendment Request for Relocating the Duke Energy Emergency Operations Facility2022-05-13013 May 2022 Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, Response to Request for Additional Information (RAI) Regarding License Amendment Request for Relocating the Duke Energy Emergency Operations Facility RA-22-0145, Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information - RAI 3.5.2.2.2.6-L2022-05-11011 May 2022 Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information - RAI 3.5.2.2.2.6-L RA-22-0106, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-04-28028 April 2022 Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-22-0124, Subsequent License Renewal Application Responses to NRC Request for Additional Information Set 42022-04-22022 April 2022 Subsequent License Renewal Application Responses to NRC Request for Additional Information Set 4 RA-22-0129, Subsequent License Renewal Application, Response to ONS SLRA 2nd Round RAI B4.1-32022-04-20020 April 2022 Subsequent License Renewal Application, Response to ONS SLRA 2nd Round RAI B4.1-3 RA-22-0089, Response to Request for Additional Information Regarding Application to Revise Technical Specification 3.7.7, Low Pressure Service Water System, to Extend the Completion Time for One Required Inoperable LPSW2022-04-14014 April 2022 Response to Request for Additional Information Regarding Application to Revise Technical Specification 3.7.7, Low Pressure Service Water System, to Extend the Completion Time for One Required Inoperable LPSW RA-22-0102, Response to Request for Additional Information (RAI) Regarding Revision 1 of DPC-NE-1007-P, Conditional Exemption of the EOC Mtc Measurement Methodology2022-04-0707 April 2022 Response to Request for Additional Information (RAI) Regarding Revision 1 of DPC-NE-1007-P, Conditional Exemption of the EOC Mtc Measurement Methodology RA-22-0111, Subsequent License Renewal Application, Follow-up Request for Additional Information Set 2 and 3 Updates2022-03-31031 March 2022 Subsequent License Renewal Application, Follow-up Request for Additional Information Set 2 and 3 Updates RA-22-0105, Subsequent License Renewal Application - Responses to NRC Requests for Confirmation of Information - Set 42022-03-22022 March 2022 Subsequent License Renewal Application - Responses to NRC Requests for Confirmation of Information - Set 4 ML22075A2032022-03-11011 March 2022 Email from Duke to NRC - Follow-Up Items from March 7, 2022 Public Meeting ML22074A0022022-03-11011 March 2022 Email from Duke to NRC - Follow-up Item from March 7, 2022 Public Meeting - SSW Tendon AMP RA-22-0040, Subsequent License Renewal Application: Responses to NRC Request for Additional Information Set 32022-02-21021 February 2022 Subsequent License Renewal Application: Responses to NRC Request for Additional Information Set 3 RA-22-0003, Response to Requests for Additional Information for Reactor Vessel Closure Stud Exam Extension Alternative2022-01-31031 January 2022 Response to Requests for Additional Information for Reactor Vessel Closure Stud Exam Extension Alternative RA-22-0023, Subsequent License Renewal Application - Response to NRC Requests for Confirmation of Information - Set 32022-01-21021 January 2022 Subsequent License Renewal Application - Response to NRC Requests for Confirmation of Information - Set 3 RA-22-0025, Supplemental Information for Relief Request to Utilize an Alternative Acceptance Criteria for Code Case, PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material2022-01-20020 January 2022 Supplemental Information for Relief Request to Utilize an Alternative Acceptance Criteria for Code Case, PWR Class 1 Primary Piping Alloy 600 Full Penetration Branch Connection Weld Metal Buildup for Material ML22019A1192022-01-0707 January 2022 Appendix 1-A - Oconee Nuclear Station 122.21(r)(2)-(13) Submittal Requirement Checklist ML22019A1242022-01-0707 January 2022 Attachment 1: Oconee Nuclear Station, Units 1, 2 & 3, Subsequent License Renewal Application, Appendix E - HDR, Inc., 2020 Clean Water Act Documents ML22019A1232022-01-0707 January 2022 Subsequent License Renewal Application, Appendix E, Responses to Requests for Additional Information and Request for Confirmation of Information ML22019A1182022-01-0707 January 2022 Enclosures 1,2 & 3: Oconee Nuclear Station, Units 1, 2 & 3, Subsequent License Renewal Application, Appendix E, Environmental Report - Index of Duke Energy'S Responses, Responses to NRC Requests for Confirmation of Information and NRC Reque RA-21-0332, Subsequent License Renewal Application Responses to NRC Request for Additional Information Set 1 and Second Round Request for Additional Information B2.1.27-1a2022-01-0707 January 2022 Subsequent License Renewal Application Responses to NRC Request for Additional Information Set 1 and Second Round Request for Additional Information B2.1.27-1a ML22019A1202022-01-0707 January 2022 Appendix K ML22019A1212022-01-0707 January 2022 Appendix 13-B Peer Reviewer Communication Log RA-22-0002, Appendix 10-B - Pump and Pipe Selection Calculations for a Hypothetical Cooling Tower Retrofit at Oconee Nuclear Station2022-01-0707 January 2022 Appendix 10-B - Pump and Pipe Selection Calculations for a Hypothetical Cooling Tower Retrofit at Oconee Nuclear Station ML22007A0152022-01-0707 January 2022 Subsequent License Renewal Application, Appendix E Responses to Requests for Additional Information (Rai), and Request for Confirmation of Information ML22019A1222022-01-0707 January 2022 Calculation of Permeability by the Falling Head Method RA-21-0325, Response to NRC Requests for Confirmation of Information - Set 22021-12-17017 December 2021 Response to NRC Requests for Confirmation of Information - Set 2 RA-21-0321, Response to Request for Additional Information (RAI) Regarding Refuel 24 (C2R24) Inservice Inspection (ISI) and Steam Generator Inspection (SG-ISI) Report2021-12-14014 December 2021 Response to Request for Additional Information (RAI) Regarding Refuel 24 (C2R24) Inservice Inspection (ISI) and Steam Generator Inspection (SG-ISI) Report 2023-07-07
[Table view] |
Text
STEVE SNID ER Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 Steve.Snider@duke-energy.com
Serial: RA 0026 10 CFR 50.55a February 11, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
-0001 CATAWBA NUCLEAR STATION, UNIT NO. 2 DOCKET NO.
50-414 / RENEWED LICENSE NO. NPF
-52 MCGUIRE NUCLEAR STATION, UNIT NOS. 1 AND 2 DOCKET NO S. 50-369, 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17 OCONEE NUCLEAR STATION, UNIT NOS. 1, 2 AND 3 DOCKET NO S. 50-269, 50-270, 50-287 / RENEWED LICENSE NOS. DPR-38, DPR-47 AND DPR-55 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50
-400 / RENEWED LICENSE NO. NPF
-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50
-261 / RENEWED LICENSE NO. DPR
-23
SUBJECT:
Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix VIII, Supplements 2 and 10 (18
-GO-001)
REFERENCES:
- 1. Duke Energy Letter, Relief Request in Accordance with 10 CFR 50.55a(g)(5)(iii) for an Alternative to the Depth Sizing Qualification Requirement of Appendix VIII, Supplements 2 and 10 (18-GO-001), dated September 6, 2018 (ADAMS Accession No. ML19011A137) 2. Duke Energy Letter, Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix VIII, Supplements 2 and 10 (18-GO-001), dated November 2, 2018 (ADAMS Accession No. ML18316A035) 3. NRC email , Duke Energy Fleet RAIs - Relief Request 18
-GO-001 - Proposed Alternative for Depth Sizing Qualification Examination of Welds (L
-2018-LLR-0117), dated January 10, 2019 (ADAMS Accession No. ML19011A137)
U.S. Nuclear Regulatory Commission RA-19-0026 Page 3 N. Jordan, NRR Project Manager
- RNP A. L. Klett, NRR Project Manager - ONS M. Mahoney, NRR Project Manager - CNS & MNS D. Galvin, NRR Project Manager - Fleet RA-19-0026
Enclosure 1 Response to Request for Additional Information
RA-19-0026 Page 1 of 6 NRC RAI-1: In the proposed relief request, the licensee discusses Code Cases N
-695 and N-695-1 (or N- 696 and N-696-1). For example, Section 4 of the relief request discusses ASME Code Case N- 695. Section 5.1 of the relief request discusses ASME Code Case N
-695-1 and N-696-1. Section 5.2 discusses Code Case N
-695. The NRC staff has approved Code Cases N
-695 and N-696, but not N
-695-1 and N-696-1, in Regulatory Guide 1.147, Revision 18.
The NRC staff notes that the ASME Code has limited the use of Code Cases N
-695 and N-696 to the 2003 Addenda or earlier editions and addenda as stated in the Code Cases for Nuclear Components of the 2015 Edition of the ASME Code. The code of record for all the plants covered under the proposed relief request is the 2007 edition through the 2008 addenda of the ASME Code,Section XI. Regulatory Guide 1.147, Revision 18 does not have the ASME Code Edition limitation on the use Code Cases N
-695 and N-696. While the proposed relief request in Section 5, "Proposed Alternative and Basis for Use,"
describes the proposed alternative use of Code Cases N
-695-1 and N-696-1, the proposed relief request does not describe the basis for use. As the NRC staff has not approved Code Cases N-695-1 and N-696-1 in Regulatory Guide 1.147, Revision 18, provide the basis for the use of Code Cases N
-695-1 and N-696-1. Duke Energy Response to RAI-1: To support the basis for use of ASME Code Cases N
-695-1 and N-696-1, the changes from the NRC approved versions of these Codes Cases (N-695 and N-696, respectively) are described and justified below.
The following summarizes the changes from Code Case N
-695 to N-695-1 1. Allowance of a root
-mean square error (RMSE) up to 0.250 in. for components 2.1 in. or greater in wall thickness. Further justification is below.
- 2. Other editorial changes
The following summarizes the changes from Code Case N
-696 to N-696-1 1. Allowance of a root
-mean square error (RMSE) up to 0.250 in. for components 2.1 in. or greater in wall thickness. Further justification is below.
- 2. Deletion of a requirement for the specimen set for Supplement 3 qualification to include at least three flaws in ferritic material. A statement was added that depth sizing qualification for ferritic piping shall be performed in accordance with Supplement 3.
- 3. Other editorial changes
To date, for components 2.1 in or greater in wall thickness, no process has met the ASME Section XI, Appendix VIII qualification requirements of a root
-mean square error (RMSE) smaller than or equal to 0.125 in. for the depth
-sizing of flaws from the inner surface (ID) in reactor pressure vessel (RPV) nozzles, according to Supplements 2, 10 or 14. These efforts have shown the impracticality of obtaining the RMSE of 0.125", given the challenges of weld geometry, rough surfaces, multiple materials, and microstructural anisotropies.
The Electric Power Research Institute (EPRI) implemented an alternate criterion which has been used by utilities in relief requests to the Nuclear Regulatory Commission (NRC), who have reviewed the criterion performance on multiple occasions. ASME Code Case N
-695-1 and RA-19-0026 Page 2 of 6 N-696-1 have been approved by the ASME Code to allow a maximum RMSE of 0.250 in. for components 2.1 in. or greater in wall thickness as a permanent solution to this issue. EPRI Technical Report 3002000612 (Reference RAI-1.1) contains the basis for this change. Deterministic and probabilistic approaches were applied in EPRI Technical Report 3002000612, which is Publicly Available, to show the acceptability of alternative depth
-sizing RMSE requirements. As shown in Enclosure 2 (Paragraph 5.1 and Reference 8.6
), Duke Energy is revising the proposed relief request to include this EPRI Technical Report. In a Deterministic Assessment, each input is set to a conservative value to account for uncertainty and variability. This methodology compounds various conservative margins in a fashion that can lead to unrealistic results and mask the true extent of conservatism in the final calculation results. These deterministic evaluations demonstrate that a depth
-sizing RMSE of 0.250" provides a structural margin for large
-diameter PWR piping welds compared to that for large-diameter BWR piping welds inspected with a depth
-sizing RMSE of 0.125". The RMSE of 0.125" currently required for the qualification of UT depth
-sizing in accordance with Supplements 2, 10, and 14 of ASME Sec. XI, App. VIII was originally a deterministic assessment based on the depth
-sizing error that was achievable for UT of BWR piping welds in the 1980s.
The use of Probabilistic Evaluations facilitates the incorporation of uncertainties, variability, and randomness important in the evaluation of leakage risk. Probabilistic assessment provides a direct uncertainty estimate for key outputs so the specific degree of conservatism in the result can be assessed. Probabilistic evaluations show that alternative depth
-sizing RMSE requirement of 0.250" has little effect on probability of through
-wall penetration of a PWSCC flaw. The probability of leakage due to through
-wall PWSCC is a key indication of the effect of PWSCC on structural integrity as large flaws are necessary to produce both leakage and pressure-boundary rupture.
The effect of uncertainty in flaw sizing on structural integrity of piping systems were assessed directly through net section collapse calculation given the presence of a circumferential flaw.
The calculations of net section collapse are based on standard equations included in ASME Section XI for evaluating acceptability for continued service of piping systems with circumferential planar flaws connected to the ID surface.
Based on these facts, EPRI recommended and ASME approved changes incorporated in N
-695-1 and N-696-1 that allow RMSE depth
-sizing qualification to be changed from 0.125" to 0.250" for large
-diameter PWR piping welds having a nominal wall thickness of at least 2.1" examined from the ID.
This recommended change to the UT qualification requirement continues to provide reasonable assurance of structural integrity and, thus, an acceptable level of quality and safety as described above.
RAI-1
References:
RAI-1.1. EPRI Technical Report 3002000612, Materials Reliability Program: Technical Basis for Change to American Society of Mechanical Engineers (ASME)Section XI Appendix VIII Root
-Mean-Square Error (RMSE) Requirement for Qualification of Depth-Sizing for Ultrasonic Testing (UT) Performed from the Inner Diameter (ID) of Large
-Diameter Thick
-Wall Supplement 2, 10, and 14 Piping Welds (MRP
-373), October 2013 RA-19-0026 Page 3 of 6 NRC RAI-2: Paragraph 4.3 of the relief request states that the vendors have demonstrated RMS errors between 0.179 inches and 0.212 inches.
Paragraph 5.2.2 of the relief request states that a correction factor equal to the difference between the procedure qualification RMS error and 0.125 inches shall be added to the depths of any measured flaws. It is not evident the exact RMS error that will be used to calculate the correction factor.
Discuss whether different vendors will be used to examine welds at different plants in the fleet.
If yes, confirm that the vendor
-specific RMS error will be used to calculate the correction factor for the specific plant in the fleet that the vendor performs weld examinations.
Duke Energy Response to RAI-2: Duke Energy intends to utilize only vendors who have demonstrated acceptable RMSE in accordance with the limits specified in N
-695-1 and N-696-1. RMS error correction factors will not be necessary with these cases. As shown in Enclosure 2, Duke Energy is revising the proposed relief request by deleting Paragraph 5.2 in its entirety (and updating paragraphs 4.1 and 5.1 to align with this deletion).
RA-19-0026 Page 4 of 6 NRC RAI-3: Note 1 to Table 1E of the relief request states that "-Oconee Unit 1 is in the process of implementing Code Case N
-716-1. Category B
-F, Item B5.10 will be replaced by the applicable Category R
-A, Item Numbers for welds 1
-RPV-WR-53 and 1-RPV-WR-53A when the inservice inspection plan and schedule are revised to implement this case-."
Note 2 of Table 1F of the relief request states that "-Robinson Unit 2 is in the process of implementing Code Case N
-716-1. Category B
-F, Item B5.10 will be replaced by the applicable Category R
-A, Item Numbers for the welds listed in Table 1F when the inservice inspection plan and schedule are revised to implement this case-"
Paragraph 3.3 of the relief request states that "-For Category R
-A welds (Oconee only), examinations are performed in accordance with ASME Code Case N-716-1. This code case does not provide alternative requirements to those specified in IWA
-2232, so the requirements of IWA-2232 apply-"
The NRC staff understands that paragraph 3.3 currently applies to Oconee Unit 2 and 3 and would apply to Oconee Unit 1 when Code Case N
-716-1 is implemented. However, since paragraph 3.3 states "(Oconee Only)", it would not apply to Robinson when Code Case N
-716-1 is implemented. Therefore, the proposed relief request does not address the applicable Code
requirement for when Code Case N
-716-1 is implemented at Robinson.
Identify the applicable Code requirement upon implementing Code Case N
-716-1 at Robinson regarding the proposed relief request.
Duke Energy Response to RAI-3: The Category R
-A weld examination requirements specified in Paragraph 3.3 of the relief request applies to both Oconee and Robinson. This is reflected in the revised relief request, Enclosure 2 , Paragraph 3.3.
RA-19-0026 Page 5 of 6 NRC RAI-4: Paragraph 5.3 of the relief request states that "-For all welds listed in this relief request, if any ID surface
-breaking flaw is detected and measured (from the ID surface) as 50 percent throughwall depth or greater, the flaw shall be considered to be of indeterminate depth. The licensee shall repair the component, or shall perform a volumetric examination from the OD surface of the component to determine the flaw depth and shall evaluate the component for continued service in accordance with the ASME Code,Section XI, IWB
-3132.3-" Paragraph 5.5 of the relief request states that "The proposed alternative for welds less than 2.1 in. (54 m[m]) in thickness is essentially identical to that approved for use during the Catawba
Unit 1 Third Inservice Inspection Interval (Precedent 7.4)-"
By letter dated December 17, 2014 (ADAMS Accession No. ML14352A261), Duke Energy submitted a similar relief request 1 CN-003, for Catawba Nuclear Station Unit 1. In the submittal, Duke Energy stated "If any inner diameter (ID) surface
-breaking flaws are detected and measured as 50% through
-wall depth or greater, Duke Energy shall repair the indications or shall perform flaw evaluations and shall submit the evaluations to the NRC for review and
approval prior to reactor startup." Duke Energy stated that the submitted flaw evaluation will include: (a) information concerning the mechanism that caused the flaw, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or weld, and (c) an estimate of the percentage of potential surface areas with UT probe "lift off' from the surface of the pipe and/or weld. By letter dated October 26, 2015 (ADAMS Accession No.
ML15286A326), the NRC approved the relief request at Catawba.
Further, previous approval of relief requests with the alternative use of Code Cases N
-695-1 and N-696-1, for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem) (ADAMS Accession No. ML17219A186) and Beaver Valley Power Station, Unit No. 2 (Beaver Valley)
(ADAMS Accession No. ML18075A096) included the submittal of flaw evaluations for NRC review and approval prior to reactor startup for detected flaws to be left in service with depths measured greater than or equal to 50 percent through wall thickness.
However, the proposed relief request Serial # 18
-GO-001, paragraph 5.3, does not discuss submitting the flaw evaluation to the NRC or what information will be involved in the flaw evaluation as discussed above and specified in the aforementioned Catawba, Salem, and Beaver Valley relief requests.
Discuss whether a flaw evaluation will be submitted to the NRC for review and approval prior to reactor startup and if it will include the aforementioned information if the detected flaw is 50
percent through
-wall depth or greater and the flaw is not repaired. If not, provide justification.
Duke Energy Response to RAI-4: For all welds listed in this request, if any inner diameter (ID) surface
-breaking flaws are detected and measured (from the ID surface) as 50% through
-wall depth or greater, Duke Energy shall repair the indications or shall perform a volumetric examination from the OD surface of the component to determine the flaw depth and shall perform flaw evaluations and shall submit the evaluations to the NRC for review and approval prior to reactor startup. The submitted flaw evaluation will include: (a) information concerning the mechanism that caused the flaw, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or weld, and (c) an estimate of the percentage of potential surface areas with UT probe "lift RA-19-0026 Page 6 of 6 off' from the surface of the pipe and/or weld. This is reflected in the revised relief request, Enclosure 2, Paragraph 5.3.
RA-19-0026
Enclosure 2 Revised Relief Request Serial #18
-GO-001
Duke Energy Carolinas, LLC Duke Energy Progress, LLC Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Relief Requested in Accordance with 10 CFR 50.55a(g)(5)(iii)
Alternative to the Depth Sizing Qualification Requirement of Appendix VIII, Supplements 2 and 10
Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 2 of 8 1. ASME Code Component(s) Affected
1.1 Class
1 Dissimilar Metal and Alloy 82/182 Welds Listed in Tables 1A through 1G. Table 1A Catawba Unit 1 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 1RPV-W52-01 1RPV-W52-02 1RPV-W52-03 1RPV-W52-04 N-770-2/B Upper Head Injection Upper Tube to Lower Tube Welds (Auxiliary Head Adapter Welds) 0.65" 1RPV-W51-01-SE 1RPV-W51-02-SE 1RPV-W51-03-SE 1RPV-W51-04-SE N-770-2/B Reactor Vessel Closure Head to Upper Head Injection Lower Tube Welds (Auxiliary Head Adapter Welds) 0.65" Table 1B Catawba Unit 2 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 2RPV-W79-101 2RPV-W80-101 2RPV-W81-101 2RPV-W82-101 N-770-2/B Upper Head Injection Upper Tube to Lower Tube Welds (Auxiliary Head Adapter Welds) 0.65" 2RPV-W79-101SE 2RPV-W80-101SE 2RPV-W81-101SE 2RPV-W82-101SE N-770-2/B Reactor Vessel Closure Head to Upper Head Injection Lower Tube Welds (Auxiliary Head Adapter Welds) 0.65" 2RPV-201-121ASE 2RPV-201-121BSE 2RPV-201-121CSE 2RPV-201-121DSE N-770-2/B Reactor Vessel Cold Leg Nozzle to Safe End Welds 2.3" 2RPV-202-121ASE 2RPV-202-121BSE 2RPV-202-121CSE 2RPV-202-121DSE N-770-2/A-2 Reactor Vessel Hot Leg Nozzle to Safe End Welds 2.4" Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 3 of 8 Table 1C McGuire Unit 1 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 1RPV3-445E-SE 1RPV3-445F-SE 1RPV3-445G-SE 1RPV3-445H-SE N-770-2/A-2 Reactor Vessel Hot Leg Nozzle to Safe End Welds 2.5" 1RPV3-445A-SE 1RPV3-445B-SE 1RPV3-445C-SE 1RPV3-445D-SE N-770-2/B Reactor Vessel Cold Leg Nozzle to Safe End Welds 2.4" 1RPV1-462C-SE 1RPV1-462B-SE 1RPV1-462A-SE 1RPV1-462D-SE N-770-2/B Reactor Vessel Closure Head to Upper Head Injection Lower Tube Welds (Auxiliary Head Adapter Welds) 0.63" 1NI1FW-38-1 1NI1FW-38-2 1NI1FW-38-3 1NI1FW-38-4 N-770-2/B Upper Head Injection Upper Tube to Lower Tube Welds (Auxiliary Head Adapter Welds) 0.63" Table 1D McGuire Unit 2 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 2RPV-W51-01-SE 2RPV-W51-02-SE 2RPV-W51-03-SE 2RPV-W51-04-SE N-770-2/B Reactor Vessel Closure Head to Upper Head Injection Lower Tube Welds (Auxiliary Head Adapter Welds) 0.63" 2RPV-W52-01 2RPV-W52-02 2RPV-W52-03 2RPV-W52-04 N-770-2/B Upper Head Injection Upper Tube to Lower Tube Welds (Auxiliary Head Adapter Welds) 0.63" Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 4 of 8 Table 1E Oconee Units 1, 2, and 3 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 1-RPV-WR-53 1-RPV-WR-53A N-770-2/B and Unit 1 Reactor Vessel Cold Leg Core Flood Nozzle
-to-Safe End Welds 1.5" B-F 1/B5.10 2-RPV-WR-53 2-RPV-WR-53A N-770-2/B and Unit 2 Reactor Vessel Cold
Leg Core Flood Nozzle
-to-Safe End Welds 1.5" N-716-1, R-A/R1.11 and R1.15 3-RPV-WR-53 3-RPV-WR-53A N-770-2/B and Unit 3 Reactor Vessel Cold Leg Core Flood Nozzle
-to-Safe End Welds 1.5" N-716-1, R-A/R1.11 and R1.15 Table 1F Robinson Unit 2 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) 107/01DM 107A/01DM 107B/01DM N-770-2/A-2 B-F 2/B5.10 Reactor Vessel Hot Leg Nozzle to Safe End Welds 2.4" 107/14DM 107A/14DM 107B/14DM N-770-2/B B-F 2/B5.10 Reactor Vessel Cold Leg Nozzle to Safe End Welds 2.4" Table 1G Harris Unit 1 Welds Component ID ASME Category or Code Case/ Inspection Item Description Nominal Nozzle Wall Thickness at Weld (Approximate) ll-RV-001 RVNOZAl-N-01SE ll-RV-001 RVNOZBl-N-03SE ll-RV-001 RVNOZCl
-N-05SE N-770-2/B Reactor Vessel Cold Leg Nozzle to Safe End Welds 2.4" ll-RV-001 RVNOZAO-N-06SE ll-RV-001 RVNOZBO-N-02SE ll-RV-001 RVNOZCO
-N-04SE N-770-2/D Reactor Vessel Hot Leg Nozzle to Safe End Welds 2.5" 1 Oconee Unit 1 is in the process of implementing Code Case N
-716-1. Category B
-F, Item B5.10 will be replaced by the applicable Category R
-A, Item Numbers for welds 1-RPV-WR-53 and 1-RPV-WR-53A when the inservice inspection plan and schedule are revised to implement this case.
2 Robinson Unit 2 is in the process of implementing Code Case N
-716-1. Category B-F, Item B5.10 will be replaced by the applicable Category R
-A, Item Numbers for the welds listed in Table 1F when the inservice inspection plan and schedule are revised to implement this case.
Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 5 of 8 2. Applicable Code Edition and Addenda 2.1 ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition with the 2008 Addenda 2.2 The inservice inspection intervals for plants included in this request are identified in Table 2. Table 2 Plant/Unit(s)
ISI Interval Interval Start Date Current Interval End Date Catawba Nuclear Station Units 1 and 2 Fourth 08/19/2015 (Unit 1) 08/19/2015 (Unit 2) 12/06/2024 (Unit 1) 02/24/2026 (Unit 2) McGuire Nuclear Station Units 1 and 2 Fourth 12/1/2011 (Unit 1) 07/15/2014 (Unit 2) 11/30/2021 (Unit 1) 12/14/2024 (Unit 2)
Oconee Nuclear Station Units 1, 2 and 3 Fifth 07/15/2014 07/15/2024 Robinson Nuclear Plant Unit 2 Fifth 07/21/2012 02/19/2023 Shearon Harris Nuclear Plant Unit 1 Fourth 09/09/2017 09/08/2027
3. Applicable Code Requirement
3.1 ASME Code Case N
-770-2, as referenced in 10 CFR 50.55a(g)(6)
(ii)(F), requires ultrasonic examination of Category A-2 , B, and D welds fabricated from Alloy 82/182 material. Table 1, Note 4 of this case requires that ultrasonic examinations meet the applicable requirements of Mandatory Appendix VIII.
3.2 For Category B
-F welds, IWA-2232 requires that ultrasonic examinations be conducted in accordance with Mandatory Appendix I. Mandatory Appendix I, I
-2220 requires that ultrasonic examinations be qualified by performance demonstration in accordance with Mandatory Appendix VIII.
3.3 For Category R
-A welds (Oconee and Robinson
), examinations are performed in accordance with ASME Code Case N
-716-1. This case does not provide alternative requirements to those specified in IWA
-2232, so the requirements of IWA
-2232 apply, as described in 3.2 above. 3.4 Mandatory Appendix VIII, Supplement 10, "Qualification Requirements for Dissimilar Metal Piping Welds", Paragraph 3.3(c) requires that "Examination procedures, equipment, and personnel are qualified for depth
-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)."3 3.5 Mandatory Appendix VIII, Supplement 2, "Qualification Requirements for Wrought Austenitic Piping Welds", Paragraph 3.2(b) requires that examination procedures, equipment, and personnel are qualified for depth
-sizing if the "RMS error of the flaw depths estimated by ultrasonics, as compared with the true depths, do not exceed 0.125 in. (3 mm)."
3 RMS (root mean square) is defined in Mandatory Appendix VIII, VIII
-3120.
Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 6 of 8 4. Impracticality of Compliance: 4.1 ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1,"
is shown as acceptable for use in Regulatory Guide (RG) 1.147, Revision 18, dated March 2017. This case provides alternatives to the requirements of Appendix VIII, Supplements 2 and 10, but paragraph 3.3(c) of this case requires that "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)." The requirement for the 0.125- inch RMS error depth sizing accuracy criteria of Code Case N- 695 is impractical because, although examination vendors have qualified for detection and length sizing in accordance with the requirements for examinations from the inside diameter (ID) surface, vendors have not met the establishe d RMS error of 0.125 inch for indication depth sizing of welds 2.1 in. (54 mm) or greater in thickness. Several process enhancements including systems, new search units, and software modifications have been implemented, but these have not been successful in demonstrating the ability to meet the required measurement error accuracy.
For these reasons, Duke Energy believes that achieving the RMS error of 0.125 inches is impractical for use with the ID ultrasonic examination technology employed in the qualification efforts. 4.2 Compliance with the requirements of Appendix VIII, Supplement 2, paragraph 3.2(b) and Supplement 10, paragraph 3.3(c) is possible for examinations performed from the outside diameter (OD) surface. However, Duke Energy has determined that examinations performed from the OD surface result in significant and unnecessary personnel radiation exposure that can be avoided by performing these examinations remotely from the ID surface.
4.3 Vendors
that Duke Energy is using for performing these examinations have demonstrated RMS errors between 0.179" and 0.212". 5. Proposed Alternative and Basis for Use:
5.1 Duke Energy proposes to use ASME Code Cases N-695-1 and N-696-1 to perform qualified ultrasonic examinations from the ID surface of the welds. Electric Power Research Institute (EPRI) Technical Report 3002000612 (Reference 8.6) contains technical basis to support this requested relief.
5.1.1 Paragraph
3.3(d) of ASME Code Case N
-695-1 states: "(d) For qualifications from the inside
-surface, examination procedures, equipment, and personnel are qualified for depth sizing if the RMS error of the flaw depth measurements, as compared to the true flaw depths, does not exceed 0.125 in. (3 mm) for piping less than 2.1 in. (54 mm) in thickness, or 0.250 in. (6 mm) for piping 2.1 in. (54 mm) or greater in thickness."
5.1.2 Paragraph
3.3(c) of ASME Code Case N
-696-1 states: "(c) Supplement 2 examination procedures, equipment, and personnel are qualified for depth
-sizing if the RMS error of the flaw depth measurements as compared to the true flaw depths, does not exceed 0.125 in. (3 mm) for piping less than 2.1 in. (54 mm) in thickness, or 0.250 in. (6 mm) for piping 2.1 in. (54 m) or greater in thickness, when they are combined with a successful Supplement 10 qualification."
Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 7 of 8 5.1.3 Personnel, procedures, and equipment shall satisfy all requirements of Code Cases N-695-1 and N-696-1. 5.1.4 Flaws detected and measured as less than 50 percent through
-wall depth shall be sized using personnel, procedures, and equipment qualified to meet the requirements of ASME Code Cases N
-695-1 and N-696-1. 5.2 (DELETED) 5.3 For all welds listed in this request, if any inner diameter (ID) surface
-breaking flaws are detected and measured (from the ID surface) as 50% through
-wall depth or greater, Duke Energy shall repair the indications or shall perform a volumetric
examination from the OD surface of the component to determine the flaw depth and shall perform flaw evaluations and shall submit the evaluations to the NRC for review and approval prior to reactor startup. The submitted flaw evaluation will include: (a) information concerning the mechanism that caused the flaw, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or weld, and (c) an estimate of the percentage of potential surface areas with UT probe "lift off' from the surface of the pipe and/or weld.
5.4 All other requirements of the ASME Code,Section XI and Code Case N
-770-2 [as conditioned by 10 CFR 50.55a(g)(6)(ii)(F)] for which relief was not specifically requested apply, including the third party review by the Authorized Nuclear Inservice Inspector.
5.5 The proposed alternative for welds less than 2.1 in. (54 m) in thickness is essentially identical to that approved for use during the Catawba Unit 1 Third Inservice Inspection Interval (Precedent 7.4). 5.6 The proposed alternative may be used in lieu of the alternative approved in Relief Request RR
-08, for the Robinson Nuclear Plant, Unit 2 Fifth Inservice Inspection Interval (Precedent 7.5). 5.7 Because compliance with the applicable requirements is impractical, this request is submitted pursuant to 10 CFR 50.55a(g)(5)(iii). Duke Energy believes that the proposed alternative provides reasonable assurance that flaws detected during examination will be sufficiently sized to disposition in accordance with acceptanc e standards of the ASME Code,Section XI and ASME Code Case N
-770-2. 6. Duration of Proposed Alternative:
This alternative is requested for the inservice inspection intervals listed in Table 2 of this request.
Revised Relief Request Serial #18-GO-001 RA-19-0026 , Enclosure 2 Page 8 of 8 7. Precedents:
The following requests for relief were granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). These requests provide similar alternatives to those proposed above. 7.1 NRC letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 1 - Relief Request NO. AN01
-ISl-025, Relief From American Society Of Mechanical EngineersSection XI Table IWB
-2500-1 Requirements (CAC No. MF7625)", dated August 29, 2016 (ADAMS Accession No. ML16237A082) 7.2 NRC letter to Pacific Gas and Electric Company, "Diablo Canyon Power Plant, Unit No. 2 - Inservice Inspection Program Relief Request NDE
-RCS-SE-2R19, Associated With The Use Of Alternate Sizing Qualification Criteria Through a Protective Clad Layer (CAC NO. MF5348)", dated November 4, 2015 (ADAMS Accession No. ML15299A034) 7.3 NRC letter to Exelon Nuclear, "Three Mile Island Nuclear Station, Unit 1 - Relief Request RR 01 Regarding Alternative Root Mean Square Depth Sizing Requirements (TAC NO. MF4873)", dated September 15, 2015 (ADAMS Accession No. ML15163A249) 7.4 NRC letter to Duke Energy Carolinas, LLC, "Catawba Nuclear Station, Unit 1: Proposed Relief Request 14
-CN-003, American Society Of Mechanical Engineers (ASME) Boiler And Pressure Vessel Code (ASME Code), Code Case N-695 (TAC NO. MF5447)", dated October 26, 2015 (ADAMS Accession No. ML15286A326) 7.5 NRC letter to Carolina Power & Light Company, H. B. Robinson Steam Electric Plant, Unit No.2 - Relief Request
-08 From ASME Code Root Mean Square Error Value For the Fifth 10
-Year Inservice Inspection Program Plan (TAC NO. MF1015), dated July 16, 2013 (ADAMS Accession No. ML13191A930)
- 8.
References:
8.1 ASME Boiler and Pressure Vessel Code,Section XI, Division 1, 2007 Edition with the 2008 Addenda 8.2 ASME Code Case N
-695, Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1 8.3 ASME Code Case N
-695-1, Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1 8.4 ASME Code Case N
-696-1, Qualification Requirements for Mandatory Appendix V III Piping Examinations Conducted From the Inside Surface,Section XI, Division 1 8.5 ASME Code Case N
-770-2, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1 8.6 EPRI Technical Report 3002000612, Materials Reliability Program: Technical Basis for Change to American Society of Mechanical Engineers (ASME)Section XI Appendix VIII Root
-Mean-Square Error (RMSE) Requirement for Qualification of Depth-Sizing for Ultrasonic Testing (UT) Performed from the Inner Diameter (ID) of Large-Diameter Thick
-Wall Supplement 2, 10, and 14 Piping Welds (MRP
-373), October 2013