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Category:Letter
MONTHYEARIR 05000261/20244012024-09-11011 September 2024 Security Baseline Inspection Report 05000261/2024401 ML24242A2612024-08-29029 August 2024 Operator Licensing Examination Approval 05000261/2024301 IR 05000261/20240052024-08-22022 August 2024 Updated Inspection Plan for H.B. Robinson Steam Electric Plant - Report 05000261/2024005 ML24169A2712024-08-14014 August 2024 – Issuance of Amendment No. 280 to Adopt TSTF-258-A, Revision 4, Regarding Changes to Technical Specification 5.7, High Radiation Area IR 05000261/20240022024-08-0101 August 2024 Integrated Inspection Report 05000261/2024002 ML24183A0972024-07-12012 July 2024 ISFSI; Catawba 1, 2 & ISFSI; McGuire 1, 2 & ISFSI; Oconee 1, 2, 3 & ISFSI; Shearon Harris 1; H. B. Robinson 2 & ISFSI; and Radioactive Package Shipping Under 10 CFR 71 (71-266 & 71-345) – Review of QA Program Changes EPID L-2024-LLQ-0002 IR 05000261/20240112024-06-0303 June 2024 Focused Engineering Inspection- Commercial Grade Dedication Report 05000261/2024011 ML24114A0152024-06-0303 June 2024 Unit 2 – Issuance of Amendment No. 279 Regarding Application of Leak-Before-Break Methodology for Auxiliary Reactor Coolant System Piping IR 05000261/20240012024-05-0909 May 2024 Integrated Inspection Report 05000261/2024001 IR 05000261/20240102024-04-30030 April 2024 Biennial Problem Identification and Resolution Inspection Report 05000261/2024010 IR 05000261/20230062024-02-28028 February 2024 Annual Assessment Letter for H.B. Robinson Steam Electric Plant Unit 2 - Report 05000261-2023006 IR 05000261/20243012024-02-0606 February 2024 – Notification of Licensed Operator Initial Examination 05000261/2024301 ML24033A0592024-02-0202 February 2024 Response to Request for Additional Information 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 IR 05000261/20230042024-01-31031 January 2024 Integrated Inspection Report 05000261/2023004 ML24009A2432024-01-25025 January 2024 Unit, No. 2 - Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0047 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML24009A2712024-01-24024 January 2024 Revision to Reactor Vessel Material Surveillance Capsule Withdrawal Schedule ML23354A0052024-01-0808 January 2024 Request for Withholding Information from Public Disclosure, H. B. Robinson Steam Electric Plant, Unit No. 2 ML23342A0902023-12-0808 December 2023 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection IR 05000261/20234202023-11-30030 November 2023 Security Baseline Inspection Report 05000261/2023420 (Cover Letter with Report) IR 05000261/20230102023-11-28028 November 2023 Fire Protection Team Inspection Report 05000261/2023010 ML23317A3462023-11-14014 November 2023 Duke Fleet - Correction Letter to License Amendment Nos. 312 & 340 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 IR 05000261/20230032023-11-0707 November 2023 Integrated Inspection Report 05000261 2023003 and 07200060 2023001 ML23226A0862023-10-12012 October 2023 Issuance of Amendment No. 277 Regarding Revision of TSs to Add High-High Steam Generator Level Function to Table 3.3.2-1 and Remove Obsolete Content from TSs 2.1.1.1 and 5.6.5.b ML23346A1322023-10-0606 October 2023 Communication from C-10 Research & Education Foundation Regarding NextEra Common Emergency Fleet Plan License Amendment Request and Related Documents Subsequently Published ML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23195A0782023-08-29029 August 2023 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1 ML23235A0552023-08-23023 August 2023 Notification of an Fire Protection Team Inspection (FPTI) (NRC Inspection Report 05000261/2023010) and Request for Information (RFI) IR 05000261/20230052023-08-21021 August 2023 Updated Inspection Plan for H.B. Robinson Steam Electric Plant (Report 05000261/2023005) IR 05000261/20230022023-08-0707 August 2023 Integrated Inspection Report 05000261/2023002 IR 05000261/20234022023-05-30030 May 2023 Cyber Security Inspection Report 05000261/2023402 ML23145A1602023-05-25025 May 2023 Submittal of Updated Final Safety Analysis Report (Revision No. 30), Independent Spent Fuel Storage Installation Safety Analysis Report (Revision No. 28), Technical Specifications Bases Revisions, Quality IR 05000261/20230012023-05-0202 May 2023 Integrated Inspection Report 05000261/2023001 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 ML22332A4932023-03-10010 March 2023 William States Lee III 1 and 2 - Issuance of Amendments Regarding the Relocation of the Emergency Operations Facility IR 05000261/20220062023-03-0101 March 2023 Annual Assessment Letter for H.B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2022006) ML23047A4512023-02-21021 February 2023 Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000261/2023402 ML23041A2272023-02-13013 February 2023 2022 Q4 Robinson_Workflow Final ML22329A2982023-01-19019 January 2023 Issuance of Amendment No. 274 Regarding Revision of Technical Specification 3.8.1 Surveillance Requirement 3.8.1.16 ML22294A0922022-12-15015 December 2022 Issuance of Amendment No. 273 Regarding Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1 ML22339A1362022-12-0101 December 2022 2023 Requalification Program Inspection - H.B. Robinson Nuclear Plant ML22322A1322022-11-18018 November 2022 301, NRC Operator License Examination Report & Cover Letter Merged 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 ML22256A2532022-11-14014 November 2022 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-541, Rev. 2 IR 05000261/20220032022-11-0303 November 2022 Integrated Inspection Report 05000261 2022003 IR 05000261/20224022022-10-19019 October 2022 Security Baseline Inspection Report 05000261/2022402 IR 05000261/20220102022-09-26026 September 2022 Design Basis Assurance Inspection (Teams) Inspection Report 05000261/2022010 IR 05000261/20223012022-09-23023 September 2022 301 Operator License Exam Approval Letter (05000261/2022301) ML22258A0302022-09-15015 September 2022 Evacuation Time Estimate Reports IR 05000261/20220052022-08-26026 August 2022 Updated Inspection Plan for the H.B Robinson Steam Electric Plant - Report 05000261/2022005-Final IR 05000261/20214042022-08-10010 August 2022 Reissue - H.B. Robinson Steam Electric Plant Security Baseline Inspection Report 05000261/2021404 2024-09-11
[Table view] Category:Report
MONTHYEARRA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, 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) 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-22-0017, 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) IR 05000261/20210052021-08-25025 August 2021 Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005) 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 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) RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) 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 ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals 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 RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits 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 RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. 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 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 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) RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RNP-RA/15-0018, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel2015-02-26026 February 2015 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel RNP-RA/14-0037, Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System2014-04-0808 April 2014 Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System RNP-RA/14-0012, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3....2014-03-12012 March 2014 Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3.... RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML13365A2912014-02-19019 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14027A0632014-01-24024 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720 ML13267A2122013-09-30030 September 2013 Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML13270A1762013-09-24024 September 2013 Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety RNP-RA/13-0079, Technical Specifications (TS) Section 5.8.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Wide Range Containment Pressure Transmitter2013-08-21021 August 2013 Technical Specifications (TS) Section 5.8.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Wide Range Containment Pressure Transmitter RNP-RA/13-0066, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for the Pressurizer Safety Valve Indication2013-06-24024 June 2013 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for the Pressurizer Safety Valve Indication RNP-RA/13-0037, Independent Spent Fuel Storage Installation, Annual Individual Exposure Monitoring Report for 20122013-04-25025 April 2013 Independent Spent Fuel Storage Installation, Annual Individual Exposure Monitoring Report for 2012 ML12331A1752012-11-26026 November 2012 Draft Bypass Fiber Quantity Test Plan ML12278A3992012-08-31031 August 2012 WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant ML1210707162012-05-17017 May 2012 Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant ML12068A1332012-02-23023 February 2012 Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG RNP-RA/11-0100, Annual Report of Changes to or Errors Discovered in an Acceptable Loss-of-Coolant Accident Evaluation Model Application for the Emergency Core Cooling System2011-11-23023 November 2011 Annual Report of Changes to or Errors Discovered in an Acceptable Loss-of-Coolant Accident Evaluation Model Application for the Emergency Core Cooling System ML1124113592011-09-23023 September 2011 Final Precursor Analysis: Electrical Fault Causes Fire and Subsequent Reactor Trip with a Loss of Reactor Coolant Pump Seal Injection and Cooling ML1124113582011-09-23023 September 2011 Final Precursor Analysis: Concurrent Unavailabilities - EDG B Inoperable Due to Failed Output Breaker and EDG a Unavailable Due to Testing and Maintenance ML1128005282010-12-29029 December 2010 NRC 2011 Hb Robinson ML1019304172010-05-0606 May 2010 Tritium Database Report RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009 WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant. RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 2024-06-28
[Table view] Category:Technical
MONTHYEARRA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, 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) 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-22-0017, 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-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML13365A2912014-02-19019 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14027A0632014-01-24024 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720 ML13267A2122013-09-30030 September 2013 Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML13270A1762013-09-24024 September 2013 Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety ML12331A1752012-11-26026 November 2012 Draft Bypass Fiber Quantity Test Plan ML12278A3992012-08-31031 August 2012 WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant ML1210707162012-05-17017 May 2012 Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant ML12068A1332012-02-23023 February 2012 Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009 WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant. RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 RNP-RA/06-0081, Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP2006-08-31031 August 2006 Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP RNP-RA/06-0027, ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.2006-03-31031 March 2006 ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis. ML0507004082004-02-20020 February 2004 EMF-3030(NP), Revision 0, Robinson Nuclear Plant, Realistic Large Break LOCA Analysis, February 2004, Non-Proprietary Version RNP-RA/03-0075, Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-62003-07-31031 July 2003 Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-6 RNP-RA/03-0031, Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 5042003-04-28028 April 2003 Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 504 ML0311207052003-04-18018 April 2003 Review of 90-day Steam Generator Tube Inservice Inspection Report for a Refueling Outage in 2001 ML0305202692003-02-15015 February 2003 Follow-up Report, Reference Event #39516 RNP-RA/03-0012, Request for Relief Pertaining to Examination Coverage Less than Essentially 100 Percent2003-02-11011 February 2003 Request for Relief Pertaining to Examination Coverage Less than Essentially 100 Percent RNP-RA/02-0172, Steam Generator Tube Plugging During Refueling Outage 212002-11-11011 November 2002 Steam Generator Tube Plugging During Refueling Outage 21 RNP-RA/02-0164, Part 17 of 17, CD-ROM Providing the Review Tool Supporting the Application for Renewal of Operating License2002-11-0606 November 2002 Part 17 of 17, CD-ROM Providing the Review Tool Supporting the Application for Renewal of Operating License ML0230402682002-09-19019 September 2002 Part 4 of 4 - Westinghouse Technology Manual, Course Outline for R-104P and Course Manual ML0221103432002-01-31031 January 2002 Caldon, Inc Engineering Report: ER-267N, Bounding Uncertainty Analysis for Thermal Power Determination at CP&L Robinson Nuclear Power Station Using the LEFM Check Plus System ML18288A3691990-10-31031 October 1990 ANF-88-054(NP)(A), PDC-3: Advanced Nuclear Fuels Corp Power Distribution Control for PWRs & Application of PDC-3 to Hb Robinson Unit 2. NRC Generic Letter 1979-451979-09-25025 September 1979 NRC Generic Letter 1979-045: Transmittal of Reports Regarding Foreign Reactor Operation Experiences 2024-06-28
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARRA-24-0165, Response to Request for Additional Information (RAI) Regarding Proposed Alternative for the Fifth Ten-Year Inservice Inspection Interval Limited Examinations2024-07-26026 July 2024 Response to Request for Additional Information (RAI) Regarding Proposed Alternative for the Fifth Ten-Year Inservice Inspection Interval Limited Examinations ML24033A0592024-02-0202 February 2024 Response to Request for Additional Information 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-23-0136, 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-23-0015, 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-22-0245, 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-22-0210, 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-22-0144, 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-22-0106, 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-21-0230, Duke Energy - Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors2021-09-30030 September 2021 Duke Energy - Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors RA-21-0063, 1, 2; Catawba Nuclear Station 1, 2; H. B. Robinson Steam Electric Plant 2; Mcgguire Nuclear Station 1, 2; Oconee Nuclear Station 1, 2, 3; Shearon Harris Nuclear Power Plant 1 - Response to RAI Re Amend for Emergency Plan2021-03-11011 March 2021 1, 2; Catawba Nuclear Station 1, 2; H. B. Robinson Steam Electric Plant 2; Mcgguire Nuclear Station 1, 2; Oconee Nuclear Station 1, 2, 3; Shearon Harris Nuclear Power Plant 1 - Response to RAI Re Amend for Emergency Plan RA-21-0032, Duke Energy - Response to Requests for Additional Info for Request to Use a Provision of Later Edition & Addenda of the ASME Boiler & Pressure Vessel Code, Section XI for Repair/Replacement Activities in Accordance with 10 CFR 50.55a(g)(42021-02-11011 February 2021 Duke Energy - Response to Requests for Additional Info for Request to Use a Provision of Later Edition & Addenda of the ASME Boiler & Pressure Vessel Code, Section XI for Repair/Replacement Activities in Accordance with 10 CFR 50.55a(g)(4)( RA-20-0048, Response to Request for Additional Information for Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections2020-03-0404 March 2020 Response to Request for Additional Information for Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections RA-20-0018, Supplement to Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding; Integrated Assessment Submittal2020-01-23023 January 2020 Supplement to Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding; Integrated Assessment Submittal RA-19-0460, Revised Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel (RPV) Surveillance Capsule Removal2019-12-19019 December 2019 Revised Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel (RPV) Surveillance Capsule Removal RA-19-0452, Seismic Probabilistic Risk Assessment (Spra), Response to March 12, 2012, Request for Information Regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Related to the Fukushima Dai-ichi Accident2019-12-12012 December 2019 Seismic Probabilistic Risk Assessment (Spra), Response to March 12, 2012, Request for Information Regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Related to the Fukushima Dai-ichi Accident RA-19-0421, Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel (RPV) Surveillance Capsule Removal (RA-19-0145)2019-11-13013 November 2019 Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel (RPV) Surveillance Capsule Removal (RA-19-0145) RA-19-0386, Response to Request for Additional Information (RAI) Regarding License Amendment Request Proposing to Revise Technical Specification 3.8.2, AC Sources - Shutdown, Surveillance Requirement 3.8.2.12019-10-24024 October 2019 Response to Request for Additional Information (RAI) Regarding License Amendment Request Proposing to Revise Technical Specification 3.8.2, AC Sources - Shutdown, Surveillance Requirement 3.8.2.1 RA-19-0154, Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors2019-05-0606 May 2019 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors 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) RA-18-0185, Response to Request for Additional Information (RAI) Regarding 10 CFR 50.46 Annual Report, Including Revised Robinson Large Break Loss of Coolant Accident Report2018-12-10010 December 2018 Response to Request for Additional Information (RAI) Regarding 10 CFR 50.46 Annual Report, Including Revised Robinson Large Break Loss of Coolant Accident Report RA-18-0194, Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt TSTF-425, Revision 32018-11-13013 November 2018 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt TSTF-425, Revision 3 RA-18-0193, Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power ..2018-11-13013 November 2018 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power .. RNP-RA/18-0050, Response to Supplemental Request for Additional Information Regarding License Amendment Request Proposing to Add a Qualified Offsite Circuit to Technical Specification 3.8.1, AC Sources - Operating and the Use of Load Tap Change2018-08-0101 August 2018 Response to Supplemental Request for Additional Information Regarding License Amendment Request Proposing to Add a Qualified Offsite Circuit to Technical Specification 3.8.1, AC Sources - Operating and the Use of Load Tap Change RA-18-0016, Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses2018-06-0505 June 2018 Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses RNP-RA/18-0036, Response to Request for Additional Information (RAI) Regarding Amendment Request Proposing to Add a Qualified Offsite Circuit to Technical Specification 3.8.1, AC Sources - Operating and the Use ...2018-05-16016 May 2018 Response to Request for Additional Information (RAI) Regarding Amendment Request Proposing to Add a Qualified Offsite Circuit to Technical Specification 3.8.1, AC Sources - Operating and the Use ... RNP-RA/17-0085, Request for Additional Information Regarding Application to Revise Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles2018-01-0808 January 2018 Request for Additional Information Regarding Application to Revise Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles RA-17-0055, Response to Second Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Revision 02017-12-19019 December 2017 Response to Second Request for Additional Information (RAI) Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Revision 0 RA-17-0048, Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2)2017-10-30030 October 2017 Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Rev. 0 (Part 2) RA-17-0043, Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Revision 02017-10-0909 October 2017 Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3009, Revision 0 RNP-RA/17-0068, Response to NRC Request for Additional Information Related to License Amendment Request Regarding Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles2017-09-28028 September 2017 Response to NRC Request for Additional Information Related to License Amendment Request Regarding Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles RA-17-0039, Response to Request for Additional Information (RAI) Regarding 10 CFR 50.55a(z)(1)Proposed Alternative to ASME Section XI Threads in Flange Examination (17-GO-001)2017-08-0909 August 2017 Response to Request for Additional Information (RAI) Regarding 10 CFR 50.55a(z)(1)Proposed Alternative to ASME Section XI Threads in Flange Examination (17-GO-001) RA-16-0042, Response to Request for Additional Information Application to Revise Technical Specifications for Methodology Reports DPC-NF-2010, Revision 3 & DPC-NE-2011-P, Revision 2. Redacted Version Enclosed2016-11-17017 November 2016 Response to Request for Additional Information Application to Revise Technical Specifications for Methodology Reports DPC-NF-2010, Revision 3 & DPC-NE-2011-P, Revision 2. Redacted Version Enclosed ML16315A2722016-11-10010 November 2016 Duke Energy Transmittal of Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-3008, Revision 0 RNP-RA/16-0079, Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Light Water Reactor Electric Generating Plants.2016-10-0505 October 2016 Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Light Water Reactor Electric Generating Plants. ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RA-16-0036, Regarding Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-1008, Rev 02016-10-0303 October 2016 Regarding Response to Request for Additional Information Regarding Application to Revise Technical Specifications for Methodology Report DPC-NE-1008, Rev 0 RA-16-0035, Response to Request for Additional Information (RAI) Regarding Application for Emergency Operations Facility (EOF) Consolidation2016-10-0303 October 2016 Response to Request for Additional Information (RAI) Regarding Application for Emergency Operations Facility (EOF) Consolidation RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RNP-RA/16-0072, Response to Request for Additional Information Regarding Application for Technical Specification Change to Adopt Technical Specifications Task Force (TSTF)-339, Relocate TS Parameters to the COLR Consistent with WCAP-14483, Rev. 22016-09-14014 September 2016 Response to Request for Additional Information Regarding Application for Technical Specification Change to Adopt Technical Specifications Task Force (TSTF)-339, Relocate TS Parameters to the COLR Consistent with WCAP-14483, Rev. 2 ML16230A2342016-07-25025 July 2016 Response to Request for Additional Information License Amendment Request to Adopt National Fire Protection Association Standard 805 RA-16-0027, Response to NRC Regulatory Issue Summary 2016-09 Preparation and Scheduling of Operator Licensing Examinations2016-07-14014 July 2016 Response to NRC Regulatory Issue Summary 2016-09 Preparation and Scheduling of Operator Licensing Examinations 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 ML16158A0062016-05-25025 May 2016 Response to Request for Additional Information Re License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Light Water Reactor Electric Generating Plants. Pages 1-21 RNP-RA/16-0031, Supplemental Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-05-0909 May 2016 Supplemental Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RNP-RA/16-0024, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-03-31031 March 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RNP-RA/16-0017, Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Light Water Reactor Electric Generating Plants.2016-03-16016 March 2016 Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Light Water Reactor Electric Generating Plants. RNP-RA/16-0011, Response to NRC Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals2016-02-18018 February 2016 Response to NRC Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals 2024-07-26
[Table view] |
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ENERGY. Serial: RNP-RA/16-0073 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. 8. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I RENEWED LICENSE NO. DPR-23 R. Michael Glover H. B. Robinson Steam Electric Plant Unit 2 Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsvllle, SC 29550 0: 843 857 1704 F: 843 8571319 Mike.Glover@duke-energy.com RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO REVISE REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS
Dear Sir/Madam:
By letter dated November 2, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15307 A069) Duke Energy Progress, Inc. (DEP) submitted a license amendment request (LAR) for H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). This LAR would revise the reactor coolant system (RCS) pressure and temperature (P-T) limits in the Technical Specifications (TSs) of the HBRSEP2. The proposed revision would extend the HBRSEP2 P-T limits applicability from the current 35 effective full power years (EFPY) up to 50 EFPY. The 50 EFPY P-T limits are based on the P-T limit curves developed in Westinghouse report, WCAP-15827, Revision 0, "H. 8. Robinson Unit 2, Heatup and Cooldown Limit Curves for Normal Operation," March 2003, which was included as Attachment 4 to the submittal. The Nuclear Regulatory Commission (NRG) staff determined that additional information is needed to complete its LAA review. A draft of that information request was received by DEP via electronic mail message dated March 2, 2016, which provided four (4) requests for additional information (RAls). An RAI clarification call was held on March 15, 2016 between NRG staff and DEP during which DEP requested a revised RAI response date for RAI No. 3 to accommodate outstanding vendor deliverables necessary for DEP's response. The DEP responses to RAls 1, 2, and 4 were provided to NRG via letter dated March 31, 2016. The attachment to this letter provides Westinghouse report, MCOE-L TR-16-33, Rev. 0, which is intended to address RAI No. 3. Please address any comments or questions regarding this matter to Mr. Tony Pilo, Acting Manager -Nuclear Regulatory Affairs at (843) 857-1409. There are no new regulatory commitments made in this letter. I declare under penalty of perjury that the foregoing is true and correct. Executed on \'-\ ' 2016. Sincerely, R. Michael Glover Site Vice President U. S. Nuclear Regulatory Commission Serial: RNP-RA/16-0073 Page2 RMG/jmw
Attachment:
Westinghouse Report MCOE-L TR-16-33, Rev. O cc: Region Administrator, NRC, Region II Mr. Dennis Galvin, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)
U. S. Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/16-0073 23 Pages (including this cover page)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO REVISE REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS
REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST REVISION OF REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS DUKE ENERGY PROGRESS, INC. H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT 2 DOCKET NO. 50-261 RAI-3 Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix G, requires that P-T limits be developed to bound all ferritic materials in the reactor pressure vessel (RPV). Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," dated October 14, 2014 (ADAMS Accession No. ML14149A165) clarifies that P-T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may define P-T curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure. The staff noted that the licensee addressed the fluence levels of the RPV inlet and outlet nozzles for the 50 EFPY in WCAP-15827, Table 6, "Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 30, 35, 40, 45 and 50 EFPY Heatup/Cooldown Curves," and reported the adjusted reference temperatures (ART) for the RPV inlet and outlet nozzles in Table 16, "Calculation of the ART Values for the 1/4T Location @ 50 EFPY" and Table 17, "Calculation of the ART Values for the 3/4T Location @ 50 EFPY." However, WCAP-15827 does not have P-T limit calculations for the RPV inlet and outlet nozzles, and therefore, does not demonstrate how the P-T limit curves developed for 50 EFPY bound all ferritic pressure boundary components of the RPV.
Therefore, the staff requests the licensee to provide P-T limit calculations for the Robinson RPV inlet and outlet nozzles or otherwise demonstrate how the P-T limit curves developed for 50 EFPY in WCAP-15827 bound all ferritic pressure boundary components of the RPV. In the P-T limit calculations for the Robinson RPV inlet and outlet nozzles, the staff requests the following to be used: 1) the ART values of the Robinson RPV inlet nozzle, outlet nozzle, and "Nozzle Welds" in Tables 16 and 17 of WCAP-15827, and 2) consideration of the stress levels in the welds that attach the Robinson RPV inlet and outlet nozzles to the RPV. Lastly, the staff requests the licensee to confirm that here are no other ferritic pressure boundary components of the Robinson RPV that need to be considered for P-T limit evaluation for the period of extended operation. DEP Response: See following pages containing Westinghouse Report, MCOE-LTR-16-33, Rev. 0.
Westinghouse Non-Proprietary Class 3 Page 1 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 NP-Attachment (Non-Proprietary) H.B. Robinson Unit 2 Cylindrical Shell Pressure-Temperature Limit Curve Applicability Check and Reactor Vessel Nozzle Pressure-Temperature Limit Curve Development for Extended Plant Operation Westinghouse Non-Proprietary Class 3 Page 2 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 H.B. Robinson Unit 2 Cylindrical Shell Pressure-Temperature Limit Curve Applicability Check and Reactor Vessel Nozzle Pressure-Temperature Limit Curve Development for Extended Plant Operation Introduction For Westinghouse nuclear steam supply systems such as H.B. Robinson Unit 2, the Topical Report WCAP-14040-NP-A, Revision 2 [Ref. 1] describes the methodology that is used to comply with the requirements of 10 CFR 50 Appendix G, "Fracture Toughness Requirements" [Ref. 2]. Since the reactor vessel materials surrounding the core region receive significant neutron fluence and undergo neutron embrittlement, the reactor vessel beltline region is considered to be the limiting reactor coolant system (RCS) component for P-T limits. The methodology in WCAP-14040-NP-A, Revision 2 only addresses the reactor vessel beltline region of the RCS as the most limiting for the P-T limits. The original Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) for this topical report states, "We find the report to be acceptable for referencing in the administrative controls section of technical specifications for license amendment applications to the extent specified and under the limitations delineated in the report and the associated NRC safety evaluation, which is enclosed. The safety evaluation defines the basis for acceptance of the report." The SE further states, "The staff finds the WCAP-14040 methodology consistent with Appendix G to Section III of the ASME Code and SRP Section 5.3.2." and "T is the metal temperature and RTNDT is the ART value of the limiting vessel material" thereby confirming that the reactor vessel is the limiting component evaluated in the development of the P-T limits. Table 1 of the NRC SE provides requirements regarding the fluence methodology, surveillance capsule program requirements, low-temperature overpressure protection system (LTOPS) requirements, adjusted reference temperature (ART) calculation, and 10 CFR 50 Appendix G temperature requirements, which have all been addressed for H.B. Robinson Unit 2 in WCAP-15827, Revision 0 [Ref. 3], consistent with the NRC SE.
Before implementation, the applicability of the 50 effective full-power years (EFPY) P-T limit curves for H.B. Robinson Unit 2 documented in WCAP-15827, Revision 0 [Ref. 3] must be confirmed. An updated fluence analysis, considering all currently completed fuel cycles and the most up-to-date prediction of future fuel cycles was completed and documented in WCAP-18100-NP, Revision 0 [Ref. 4]. The fluence values documented in WCAP-18100-NP, Revision 0 [Ref. 4] and all relevant sister-plant data are used herein to confirm that the 50 EFPY P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable through 50 EFPY for H.B. Robinson Unit 2.
Additionally, the H.B. Robinson reactor vessel inlet and outlet nozzles must now be considered during development of pressure-temperature (P-T) limit curves due to the recent issuance of the Regulatory Issue Summary (RIS) 2014-11 [Ref. 5]. Although one quarter-thickness (1/4T) and three quarter-thickness (3/4T) Adjusted Reference Temperature (ART) values were calculated in WCAP-15827, Revision 0 [Ref. 3], nozzle P-T limit curves based on nozzle-specific stresses have not been previously analyzed for H.B. Robinson Unit 2. The reactor vessel inlet and outlet nozzle corner region is more stressed than the cylindrical vessel regions due to the geometric discontinuity at the nozzle inner radius. As a result, the Westinghouse Non-Proprietary Class 3 Page 3 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 inlet and outlet nozzle corner regions, along with the vessel flange region, are considered to be the bounding stress concentration within the reactor vessel. Although the nozzle welds are subjected to primary (pressure) and secondary (cooldown transient) stresses, the nozzle welds are bounded by the larger nozzle corner region stresses because of the discontinuities at the nozzle inner radius. Also, the nozzle welds are included in the cylindrical shell P-T limits.
Note that the 10 CFR 50, Appendix G [Ref. 2] flange requirements have previously been incorporated into the 50 EFPY cylindrical shell P-T limits per WCAP-15827, Revision 0 [Ref. 3]. Therefore, the 50 EFPY cylindrical shell P-T limit curves previously generated for H.B. Robinson Unit 2 must now be checked against the nozzle P-T limit curves to ensure that the 50 EFPY cylindrical shell curves remain the most limiting.
50 EFPY Cylindrical Shell P-T Limit Curves Applicability Check The H.B. Robinson Unit 2 calculated maximum neutron fluence projections at 50 EFPY for the reactor vessel materials were originally documented in Table 6 of WCAP-15827, Revision 0 [Ref. 3]. However, these values were recently updated in an Ex-Vessel Neutron Dosimetry (EVND) analysis, WCAP-18100-NP [Ref. 4]. Table 1 compares the beltline fluence projections from the original 50 EFPY cylindrical shell P-T limit curves report [Ref. 3] with those in the most recent EVND fluence analysis [Ref. 4]. Per RIS 2014-11 [Ref. 5] guidance, beltline materials were determined to be the reactor vessel materials that will be exposed to a neutron fluence greater than or equal to 1 x 1017 n/cm2 (E > 1.0 MeV) at end of license (EOL, 50 EFPY). Table 1 H.B. Robinson Unit 2 Calculated Neutron Fluence Projections on the Reactor Vessel Beltline Materials at 50 EFPY Reactor Vessel Material 50 EFPY Fluence (n/cm2, E > 1.0 MeV) WCAP-15827 WCAP-18100-NP Inlet Nozzle to Upper Shell Weld - Lowest Extent(a) 3.93 x 1017 1.14 x 1017 Outlet Nozzle to Upper Shell Weld - Lowest Extent(a) 2.53 x 1017 1.35 x 1017 Upper Shell Plates 2.50 x 1019 2.45 x 1019 Upper Shell Longitudinal Welds 2.50 x 1019 1.80 x 1019 Upper Shell to Intermediate Shell Circumferential Weld 2.50 x 1019 2.45 x 1019 Intermediate Shell Plates 6.01 x 1019 5.69 x 1019 Intermediate Shell Longitudinal Welds 4.46 x 1019 4.18 x 1019 Intermediate to Lower Shell Circumferential Weld 2.05 x 1019 2.03 x 1019 Lower Shell Plates 2.05 x 1019 2.03 x 1019 Lower Shell Longitudinal Welds 2.05 x 1019 2.03 x 1019 Note for Table 1: (a) The fluence for the inlet and outlet nozzle to upper shell welds was also conservatively used as the fluence for the inlet and outlet nozzle materials. The actual nozzle forging fluence values, at the location of a postulated 1/4T flaw along the nozzle corner region, are expected to be lower since they are further away from the active core.
Westinghouse Non-Proprietary Class 3 Page 4 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Per Table 1, the fluence values reported in WCAP-15827, Revision 0 [Ref. 3] are conservative when compared to those from the revised analysis, WCAP-18100-NP, Revision 0 [Ref. 4]. Thus, the ART values calculated in development of the P-T limit curves in WCAP-15827, Revision 0 [Ref. 3] are conservative for all materials analyzed, except potentially the upper to intermediate shell circumferential weld material (Heat # W5214). The ART value for this material must be re-calculated, because additional sister plant data now exists for this weld heat that was not available for incorporation in WCAP-15827, Revision 0 [Ref. 3]. Thus, the upper to intermediate shell circumferential weld material is the only cylindrical shell material re-analyzed as a part of this P-T limits applicability evaluation.
The H.B. Robinson Unit 2 upper to intermediate shell circumferential weld material was fabricated using weld Heat # W5214. In addition to being contained in the H.B. Robinson Unit 2 surveillance program, weld Heat # W5214 is contained in the Palisades supplemental surveillance program, as well as in the Indian Point Units 2 and 3 surveillance programs. Thus, the H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 data will be used in the calculation of the Position 2.1 chemistry factor (CF) value for the H.B. Robinson Unit 2 weld Heat # W5214. Table 2 summarizes the applicable surveillance capsule data pertaining to weld Heat # W5214, and Table 3 shows the calculation of the Position 2.1 CF. The combined surveillance data is considered not fully credible per SIA Report 0901132.401, Revision 0 [Ref. 6]. Thus, the Position 2.1 CF can be used for determining ART values, but a full margin term must be included in the calculation as established in Reference 6. This calculation methodology is consistent with the NRC-approved conclusions of ML112870050 [Ref. 7] and ML113480303 [Ref. 8].
H.B. Robinson Unit 2 upper to intermediate shell circumferential weld seam (Heat # W5214) 1/4T and 3/4T ART values were calculated per Regulatory Guide 1.99, Revision 2 [Ref. 9]. These ART calculations are summarized in Tables 4 and 5.
Westinghouse Non-Proprietary Class 3 Page 5 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 2 Surveillance Capsule Data for Weld Heat # W5214(a) Weld Metal Heat # W5214 Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV) Measured RTNDT (°F) Tcapsule (°F) Temperature Adjustment(b) (°F) Wt. % Cu Wt. % Ni Position 1.1 CF(c) (°F) H.B. Robinson Unit 2 Data V 0.530(d) 208.8 547 0.0 0.34 0.66 217.7 T 3.87(d) 289.1 547 0.0 X 4.49(d) 265.6 547 0.0 Palisades Data SA-60-1 1.50 259 535.0 -12.0 0.307 1.045 266.5 SA-240-1 2.38 280.1 535.7 -11.3 Indian Point Unit 2 Data Y 0.455 193.9 529.1 -17.9 0.20 1.03 226.3 V 0.492 197.5 524 -23.0 Indian Point Unit 3 Data T 0.263 149.8 539.4 -7.6 0.16 1.12 206.2 Y 0.692 171.1 539.5 -7.5 X 0.874 192.5 539.7 -7.3 Z 1.04 228.3 538.9 -8.1 Notes for Table 2: (a) All data contained here taken from SIA Report 0901132.401, Revision 0 [Ref. 6], unless otherwise noted. (b) Temperature adjustment = 1.0*(Tcapsule - Tplant), where Tplant = 547°F for H.B. Robinson Unit 2 (applied to the weld RTNDT data for each of the Palisades, Indian Point Unit 2, and Indian Point Unit 3 capsules in the Position 2.1 chemistry factor calculation). (c) CF = Chemistry factor. Calculated using Table 1 of Regulatory Guide 1.99, Revision 2 [Ref. 9]. Note that per WCAP-15827, Revision 0 [Ref. 3] the Position 1.1 Chemistry Factor (CF) of the H.B. Robinson Unit 2 upper to intermediate shell circumferential weld seam is 230.2°F. (d) The fluence values assigned to the H.B. Robinson Unit 2 capsules are consistent with the values documented in WCAP-15827, Revision 0 [Ref. 3] and WCAP-18100-NP, Revision [Ref. 4].
Westinghouse Non-Proprietary Class 3 Page 6 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 3 Calculation of H.B. Robinson Unit 2 Position 2.1 Chemistry Factor Value for Weld Heat # W5214 Using Surveillance Capsule Data Weld Metal Heat # W5214 Capsule Capsule Fluence(a) (x 1019 n/cm2, E > 1.0 MeV)FF(b) RTNDT(c) (°F) FF*RTNDT(°F) FF2 H.B. Robinson Unit 2 Data V 0.530 0.823 221.33 (208.8) 182.07 0.677 T 3.87 1.349 306.45 (289.1) 413.40 1.820 X 4.49 1.381 281.54 (265.6) 388.71 1.906 Palisades Data SA-60-1 1.50 1.112 212.42 (259) 236.27 1.237 SA-240-1 2.38 1.234 231.17 (280.1) 285.23 1.522 Indian Point Unit 2 Data Y 0.455 0.781 179.52 (193.9) 140.17 0.610 V 0.492 0.802 177.99 (197.5) 142.78 0.643 Indian Point Unit 3 Data T 0.263 0.637 159.26 (149.8) 101.41 0.405 Y 0.692 0.897 183.23 (171.1) 164.31 0.804 X 0.874 0.962 207.42 (192.5) 199.59 0.926 Z 1.04 1.011 246.62 (228.3) 249.33 1.022 SUM : 2503.25 11.573 CFHeat # W5214 = (FF
- RTNDT) ÷ (FF2) = (2503.25) ÷ (11.573) = 216.3ûF Notes for Table 3: (a) Taken from Table 2 herein. (b) FF = fluence factor = f(0.28 - 0.10*log f). (c) RTNDT values are the measured 30 ft-lb shift values. The RTNDT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 2 of this report). The temperature adjustments are listed in Table 2 herein. Ratio applied to the H.B. Robinson Unit 2 surveillance data = CFVessel Weld / CFSurv. Weld = 230.2°F / 217.7°F = 1.06. Ratio applied to the Palisades surveillance data = CFVessel Weld / CFSurv. Weld = 230.2°F / 266.5°F = 0.86. Ratio applied to the Indian Point Unit 2 surveillance data = CFVessel Weld / CFSurv. Weld = 230.2°F / 226.3°F = 1.02. Ratio applied to the Indian Point Unit 3 surveillance data = CFVessel Weld / CFSurv. Weld = 230.2°F / 206.2°F = 1.12. The Position 1.1 CF for the H.B. Robinson Unit 2 reactor vessel weld of 230.2°F was taken from WCAP-15827, Revision 0 [Ref. 3].
Westinghouse Non-Proprietary Class 3 Page 7 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 4 ART Calculations at the 1/4T Location for the H.B. Robinson Unit 2 Upper to Intermediate Shell Circumferential Weld Seam 10-273(a) Reactor Vessel Material and ID Number Heat Number CF (°F) 1/4T Fluence(d) (n/cm2, E > 1.0 MeV) 1/4T FF(d) RTNDT(U)(e) (°F) RTNDT (°F) I (°F) (f) (°F) Margin (°F) ART (°F) Upper to Intermediate Shell Circumferential Weld 10-273 W5214 230.2(b) 1.40 x 1019 1.0936 -56 251.8 17 28 65.5 261 Using surveillance data W5214 216.3(c) 1.40 x 1019 1.0936 -56 236.6 17 28 65.5 246 Notes for Table 4: (a) The Regulatory Guide 1.99, Revision 2 [Ref. 9] methodology was utilized in the calculation of the ART values. (b) This Position 1.1 CF was taken from WCAP-15827, Revision 0 [Ref. 3] (c) This Position 2.1 CF was taken from Table 3. (d) 1/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2 [Ref. 9], the H.B. Robinson Unit 2 reactor vessel beltline thickness of 9.313 inches, and the surface fluence value reported in Table 1. (e) Taken from Table 1 WCAP-15827, Revision 0 [Ref. 3]. (f) As documented previously, the surveillance data for weld Heat # W5214 was deemed not fully credible. Thus, a full margin term ( = 28°F) must be used for both Position 1.1 and 2.1; however, the lower of the two values may be taken as the ART value (consistent with NRC-approved conclusions of References 7 and 8).
Westinghouse Non-Proprietary Class 3 Page 8 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 5 ART Calculations at the 3/4T Location for the H.B. Robinson Unit 2 Upper to Intermediate Shell Circumferential Weld Seam 10-273(a) Reactor Vessel Material and ID Number Heat Number CF (°F) 3/4T Fluence(d) (n/cm2, E > 1.0 MeV) 3/4T FF(d) RTNDT(U)(e) (°F) RTNDT (°F) I (°F) (f) (°F) Margin (°F) ART (°F) Upper to Intermediate Shell Circumferential Weld 10-273 W5214 230.2(b) 4.58 x 1018 0.7828 -56 180.2 17 28 65.5 190 Using surveillance data W5214 216.3(c) 4.58 x 1018 0.7828 -56 169.3 17 28 65.5 179 Notes for Table 5: (a) The Regulatory Guide 1.99, Revision 2 [Ref. 9] methodology was utilized in the calculation of the ART values. (b) This Position 1.1 CF was taken from WCAP-15827, Revision 0 [Ref. 3] (c) This Position 2.1 CF was taken from Table 3. (d) 3/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2 [Ref. 9], the H.B. Robinson Unit 2 reactor vessel beltline thickness of 9.313 inches, and the surface fluence value reported in Table 1. (e) Taken from Table 1 WCAP-15827, Revision 0 [Ref. 3]. (f) As documented previously, the surveillance data for weld Heat # W5214 was deemed not fully credible. Thus, a full margin term ( = 28°F) must be used for both Position 1.1 and 2.1; however, the lower of the two values may be taken as the ART value (consistent with NRC-approved conclusions of References 7 and 8).
Westinghouse Non-Proprietary Class 3 Page 9 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 In order to check the applicability of the 50 EFPY cylindrical shell curves documented in WCAP-15827, Revision 0 [Ref. 3], the limiting 1/4T and 3/4T ART values used to develop the curves must be compared to the revised upper to intermediate shell circumferential weld 1/4T and 3/4T ART values documented in Tables 4 and 5, respectively. Table 6 compares the revised ART values for the H.B. Robinson Unit 2 upper to intermediate circumferential weld at 50 EFPY with those used in the development of the P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3]. The values calculated in WCAP-15827, Revision 0 [Ref. 3] for the cylindrical shell remain bounding; thus, the 50 EFPY cylindrical shell P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable through 50 EFPY. Table 6 H.B. Robinson Unit 2 Revised and Limiting Cylindrical Shell ART Values at 50 EFPY(a) ART Value (°F) 1/4T Location 3/4T Location Upper to Intermediate Shell Circumferential Weld 10-273 (Heat # W5214) Limiting(a) 263 191 Re-calculated, herein(b) 246 179 Notes for Table 6: (a) Values were taken from WCAP-15827, Revision 0 [Ref. 3] and represent circumferential flaw ART values. These values are conservative when compared to the revised ART values based on updated fluence and all applicable capsule results. Thus, the 50 EFPY cylindrical shell P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable through 50 EFPY. (b) Values were taken from Tables 4 and 5 and represent circumferential flaw ART values. Since the data was not fully credible, a full margin term was used; however, the Position 2.1 ART value was assigned to this material, since the Position 2.1 value was lower (consistent with NRC-approved conclusions of References 7 and 8).
Westinghouse Non-Proprietary Class 3 Page 10 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Nozzle Initial RTNDT and ART Calculations H.B. Robinson reactor vessel inlet and outlet nozzle material properties were used to determine initial RTNDT and the ART values per Regulatory Guide 1.99, Revision 2 [Ref. 9]. The initial RTNDT values for the reactor vessel inlet nozzle forging materials were determined using the Branch Technical Position (BTP) 5-3 Position 1.1(4) methodology [Ref. 10] since limited Charpy V-notch tests were performed at a single temperature for the inlet nozzles. The initial RTNDT values for the reactor vessel outlet nozzle forging materials were determined using the BWRVIP-173-A, Appendix B, Alternative Approach 2 methodology [Ref. 11] since no drop-weight test data was available for the outlet nozzles. These updated initial RTNDT values are summarized in Table 7. The copper (Cu) and nickel (Ni) chemistry information and CF values (determined per Reference 9) for each of the H.B. Robinson Unit 2 reactor vessel nozzle materials are also summarized in Table 7.
Table 7 Summary of the H.B. Robinson Unit 2 Reactor Vessel Nozzle Material Properties(a) Reactor Vessel Material Material Heat ID # Wt. % Cu Wt. % Ni Position 1.1 CF(b) (°F) RTNDT(u)(c) (°F) Inlet Nozzle W-10207-1 X15156/X53163 0.02 0.90 20 10 Inlet Nozzle W-10207-2 X15156/X53163 0.02 0.90 20 10 Inlet Nozzle W-10207-3 X15156/X53163 0.02 0.90 20 10 Outlet Nozzle B-3201-1 BT2305 [[ {E}]](d) 0.71 137.9 -7.8 Outlet Nozzle B-3201-2 BT2305 [[ {E}]](d) 0.71 137.9 1.6 Outlet Nozzle B-3201-3 BT2305 [[ {E}]](d) 0.71 137.9 7.2 Notes for Table 7: (a) Nozzle information was determined per the H.B. Robinson Unit 2 Certified Material Test Reports (CMTRs). Nozzle chemistry values are the maximum of the reported values in the CMTRs, unless otherwise noted. (b) The Position 1.1 CF values for the nozzle materials are calculated based on Table 2 of Regulatory Guide 1.99, Revision 2 [Ref. 9] using the Cu and Ni weight percent (wt. %) values summarized in this table. (c) Inlet nozzle forging initial RTNDT values were determined using the Branch Technical Position (BTP) 5-3 Position 1.1(4) methodology [Ref. 10]. Outlet nozzle forging initial RTNDT values for the reactor vessel outlet nozzle forging materials were determined using the BWRVIP-173-A, Appendix B, Alternative Approach 2 methodology [Ref. 11]. (d) Since no Cu wt. % data exists for these materials, Cu wt. % values are the best-estimate values for SA-508, Class 2 low allow steels as documented in BWRVIP-173-A [Ref. 11]. This is Electric Power Research Institute (EPRI) proprietary information.
Westinghouse Non-Proprietary Class 3 Page 11 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 8 shows the calculation of the maximum ART values for the H.B. Robinson Unit 2 nozzle materials. The fast neutron fluences (E > 1.0 MeV) for the inlet and outlet nozzle to upper shell welds (lowest extent) were calculated conservatively in Reference 4. The fluences for the inlet and outlet nozzle to upper shell welds (lowest extent) were taken from an elevation several centimeters closer to the core than the actual location of the welds. In addition, the methodology that was used to calculate the fast neutron fluences (E > 1.0 MeV) utilized the two-dimensional/one-dimensional fluence rate synthesis technique to obtain three-dimensional (3D) synthesized fluence rates. The fluence rate synthesis technique provides more conservative results (more than 50% in relative difference) near nozzle elevations, compared to full 3D radiation transport solutions. Since the calculated values of RTNDT are less than 25°F for each nozzle material, embrittlement effects may be neglected per the following conclusions of Section 4 of TLR-RES/DE/CIB-2013-01 [Ref. 12]: Based on the results of the studies documented in this report, the RES staff recommends the following definition for the RPV [Reactor Pressure Vessel] beltline region: 1. The beltline is defined as the region of the RPV adjacent to the reactor core that is projected to receive a neutron fluence level of 1x1017 n/cm2 (E > 1.0 MeV) or higher at the end of the licensed operating period. 2. Embrittlement effects may be neglected for any region of the RPV if either of the following conditions are met: (1) neutron fluence is less than 1x1017 n/cm2 (E > 1.0 MeV) at EOL, or (2) the mean value of T30 [equal to RTNDT] estimated using an ETC [Embrittlement Trend Correlation] acceptable to the staff is less than 25°F at EOL. The estimate of T30 at EOL shall be made using best-estimate chemistry values.
Westinghouse Non-Proprietary Class 3 Page 12 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Table 8 Surface ART Calculations for the H.B. Robinson Unit 2 Nozzle Forging Materials(a) Reactor Vessel Material and ID Number Heat Number CF(b) (°F) Surface Fluence(c) (n/cm2, E > 1.0 MeV) FF(c) RTNDT(U)(b) (°F) RTNDT(d) (°F) I (°F) (e) (°F) Margin (°F) ART (°F) Inlet Nozzle W-10207-1 X15156/X53163 20 1.14 x 1017 0.1198 10 0 (2.4) 0 0 0.0 10 Inlet Nozzle W-10207-2 X15156/X53163 20 1.14 x 1017 0.1198 10 0 (2.4) 0 0 0.0 10 Inlet Nozzle W-10207-3 X15156/X53163 20 1.14 x 1017 0.1198 10 0 (2.4) 0 0 0.0 10 Outlet Nozzle B-3201-1 BT2305 137.9 1.35 x 1017 0.1339 -7.8 0 (18.5) 0 0 0.0 -7.8 Outlet Nozzle B-3201-2 BT2305 137.9 1.35 x 1017 0.1339 1.6 0 (18.5) 0 0 0.0 1.6 Outlet Nozzle B-3201-3 BT2305 137.9 1.35 x 1017 0.1339 7.2 0 (18.5) 0 0 0.0 7.2 Notes for Table 8: (a) The Regulatory Guide 1.99, Revision 2 [Ref. 9] methodology was utilized in the calculation of the ART values. (b) Values taken from Table 7. (c) Surface fluence was taken from Table 1. FF values were calculated using Regulatory Guide 1.99, Revision 2 [Ref. 9]. (d) Calculated RTNDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 [Ref. 12]. Actual calculated RTNDT values are listed in parentheses for these materials for information purposes. (e) Per Regulatory Guide 1.99, Revision 2 [Ref. 9], need not exceed 0.5*RTNDT.
Westinghouse Non-Proprietary Class 3 Page 13 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Note that the nozzle corner fracture mechanics calculations have utilized the limiting inlet and outlet nozzle ART values as input to the nozzle P-T limit curves. If the projected fluence at the H.B. Robinson Unit 2 nozzle materials should increase in the future (e.g., due to a power uprate), the embrittlement calculations can be checked to ensure that the RTNDT for these materials remains less than 25°F and can subsequently be set equal to 0°F per TLR-RES/DE/CIB-2013-01 [Ref. 12], thus allowing the conclusions of this letter to remain unchanged. Based on the nozzle material properties documented in Table 7, the inlet nozzle RTNDT values will remain below 25°F until they reach a fluence value of 2.54 x 1019 n/cm2 (E > 1.0 MeV). The outlet nozzle RTNDT values will remain below 25°F until they reach a fluence value of 2.18 x 1017 n/cm2 (E > 1.0 MeV). The reactor vessel inlet and outlet nozzle forging limiting ART values for H.B. Robinson Unit 2 are summarized in Table 9.
Table 9 Summary of Limiting ART Values for the Inlet and Outlet Nozzle Materials for H.B. Robinson Unit 2 EFPY Nozzle Material and ID Number Limiting ART Value (°F) 50 Inlet Nozzles W-10207-1, W-10207-2, and W-10207-3 10 Outlet Nozzle B-3201-3 7.2 Westinghouse Non-Proprietary Class 3 Page 14 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Nozzle P-T Limits A calculation of the H.B. Robinson Unit 2 nozzle cooldown P-T limits was completed using the inlet and outlet nozzle ART values from Table 9. The stress intensity factor correlations used for the nozzle corners are consistent with the ASME PVP2011-57015 [Ref. 13] and Oak Ridge National Laboratory (ORNL) study, ORNL/TM-2010/246 [Ref. 14]. The methodology used included postulating an inside surface 1/4T nozzle corner flaw, along with calculating through-wall nozzle corner stresses for a cooldown rate of 100ºF/hour.
The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form:
=A0+A1x+A2x2+A3x3 where,
= through-wall stress distribution x = through-wall distance from inside surface A0, A1, A2, A3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below The stress intensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010/246. The stress intensity factor expression for a rounded corner is: KI= a0.706A0+ 0.5372aA1+ 0.448a22A2+ 0.3934a33A3 where, KI = stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius corner a = crack depth at the nozzle corner, for use with 1/4T (25% of the wall thickness) The H.B. Robinson Unit 2 inlet and outlet nozzle P-T limit curves shown in Figures 1 and 2 are based on the stress intensity factor expression discussed above; also shown in these figures are the traditional beltline P-T limits from WCAP-15827, Revision 0 [Ref. 3]. The nozzle P-T limits and the traditional beltline P-T limits are applicable to 50 EFPY. The nozzle P-T limits are provided for a cooldown rate of 100°F/hr, as well as at steady-state.
It should be noted that an outside surface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve Westinghouse Non-Proprietary Class 3 Page 15 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 is not provided, since it would be less limiting than the cooldown nozzle P-T limit curve in Figures 1 and 2 for an inside surface flaw.
Based on the results shown in Figures 1 and 2, it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the 50 EFPY P-T limits contained in WCAP-15827, Revision 0 [Ref. 3] are applicable for the beltline and non-beltline reactor vessel components of H.B. Robinson Unit 2 through 50 EFPY based on the technical evaluations contained herein.
Westinghouse Non-Proprietary Class 3 Page 16 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Figure 1: Comparison of H.B. Robinson Unit 2 50 EFPY Inlet Nozzle P-T Limits and Cylindrical Shell P-T Limits 0250 5007501000 1250 1500 1750 2000 22502500050100150200250300350400450500550Reactor Coolant System Pressure (psig)Average Reactor Coolant System Temperature (°F)Inlet Nozzle Cooldown -100 °F/hr Inlet Nozzle Steady StateBoltup TemperatureCooldown Rates
ºF/Hr Steady-State-20-40-60
-100 Westinghouse Non-Proprietary Class 3 Page 17 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Figure 2: Comparison of H.B. Robinson Unit 2 50 EFPY Outlet Nozzle P-T Limits and Cylindrical Shell P-T Limits 0250 500 7501000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Reactor Coolant System Pressure (psig)Average Reactor Coolant System Temperature (°F)BoltupTemperatureOutlet Nozzle Steady StateOutlet Nozzle Cooldown -100 °F/hr Cooldown RatesºF/HrSteady-State 40-60-100 Westinghouse Non-Proprietary Class 3 Page 18 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Other Ferritic Components in the Reactor Coolant Pressure Boundary (RCPB) 10 CFR Part 50, Appendix G [Ref. 2], requires that all RCPB components meet the requirements of Section III of the ASME Code. The ferritic RCPB components that are not part of the reactor vessel consist of the pressurizer, replacement reactor vessel head, and replacement steam generators. The H.B. Robinson Unit 2 pressurizer was built in approximately 1967 and therefore was analysed to the ASME Code Section III 1965 Edition. The H.B. Robinson Unit 2 replacement reactor vessel head was analysed to the ASME Section III 1998 Edition with 2000 Addendum. The H.B. Robinson Unit 2 replacement steam generators were analysed to the ASME Code Section III 1965 Edition with Summer 1966 Addendum. These components met all applicable requirements at the time of construction; therefore, no further consideration is necessary for these components with regards to P-T limits.
The lowest service temperature (LST) requirement of NB-2332(b) of ASME Section III is applicable to material for ferritic piping, pumps, and valves with a nominal wall thickness greater than 2 1/2 inches [Ref. 15]. Note that the H.B. Robinson Unit 2 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps, or valves. Therefore, the LST requirements of NB-2332(b) are not applicable to the H.B. Robinson Unit 2 P-T limits.
Westinghouse Non-Proprietary Class 3 Page 19 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 Conclusions Based on a comparison of the fluence values assigned to the H.B. Robinson Unit 2 materials in the P-T limits report containing the 50 EFPY cylindrical shell curves [Ref. 3] to the fluence values assigned to the same materials in the most recent EVND fluence analysis [Ref. 4] in Table 1, it was concluded that the 50 EFPY cylindrical shell curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable for all materials analyzed, except potentially those materials using sister plant data. The only H.B. Robinson Unit 2 material that uses sister plant data is the upper to intermediate shell circumferential weld (Heat # W5214). The Position 2.1 CF was calculated for this material in Table 3, and this Position 2.1 CF was used to recalculate the 1/4T and 3/4T ART values for the upper to intermediate shell circumferential weld in Tables 4 and 5. As determined in Table 6, the limiting ART values used for the development of the 50 EFPY P-T limit curves in WCAP-15827, Revision 0 [Ref. 3] are bounding when compared to the upper to intermediate shell circumferential weld ART values calculated using updated fluence and all applicable capsule results. Thus, the H.B. Robinson Unit 2 cylindrical shell 50 EFPY P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable through 50 EFPY. The chemistry, CF, and initial RTNDT values of the H.B. Robinson Unit 2 inlet nozzle and outlet nozzle materials are presented in Table 7. Chemistry information and original test data were taken from the H.B. Robinson Unit 2 CMTRs. Inlet nozzle forging initial RTNDT values were determined using the Branch Technical Position (BTP) 5-3 Position 1.1(4) methodology [Ref. 10], while outlet nozzle forging initial RTNDT values for the reactor vessel outlet nozzle forging materials were determined using the BWRVIP-173-A, Appendix B, Alternative Approach 2 methodology [Ref. 11].
The calculated reactor vessel inlet and outlet nozzle forging ART values for H.B. Robinson Unit 2 at 50 EFPY are presented in Table 8. Since the 50 EFPY nozzle material RTNDT values are below 25°F, embrittlement effects are considered negligible per TLR-RES/DE/CIB-2013-01 [Ref. 12]. The limiting inlet and outlet nozzle ART values for H.B. Robinson Unit 2 at 50 EFPY are summarized in Table 9.
Although neutron embrittlement does not need to be considered for the H.B. Robinson Unit 2 reactor vessel inlet and outlet nozzles through 50 EFPY, the inside corner regions of the reactor vessel nozzles are considered to be highly stressed. P-T limit curves were developed for the reactor vessel nozzles at 50 EFPY using the ART values from Table 9. Figures 1 and 2 display the inlet and outlet nozzle P-T limit curves, and demonstrate that the 50 EFPY P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] bound the reactor vessel nozzle P-T limit curves through 50 EFPY for H.B. Robinson Unit 2.
Additionally, other ferritic components in the RCPB were considered. The H.B. Robinson Unit 2 pressurizer, replacement reactor vessel head, and replacement steam generators met all applicable ASME Code Section III requirements at the time of construction; therefore, no further consideration is necessary for these components with regards to P-T limits. Furthermore, the LST requirements of NB-2332(b) are not applicable to the P-T limits since the H.B. Robinson Unit 2 reactor coolant systems do not have ferritic materials in the Class 1 piping, pumps, or valves. In conclusion, the H.B. Robinson Unit 2 50 EFPY P-T limit curves documented in WCAP-15827, Revision 0 [Ref. 3] remain applicable through 50 EFPY and no further changes are necessary.
Westinghouse Non-Proprietary Class 3 Page 20 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016
References:
1. Westinghouse Report WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996. 2. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995. 3. Westinghouse Report WCAP-15827, Revision 0, "H.B. Robinson Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," March 2003. 4. Westinghouse Report WCAP-18100-NP, Revision 0, "Ex-Vessel Neutron Dosimetry Program for H. B. Robinson Unit 2 Cycles 16 through 29," March 2016. 5. U. S. NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," dated October 14, 2014. [Agencywide Document Access and Management System (ADAMS) Accession Number ML14149A165] 6. Structural Integrity Associates, Inc. Report No. 0901132.401, Revision 0, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," April 2010. [ADAMS Accession Number ML110060693] 7. "Updated Reactor Pressure Vessel Pressurized Thermal Shock Evaluation for Palisades Nuclear Plant (TAC No. ME5263)," dated December 7, 2011. [ADAMS Accession Number ML112870050] 8. "Palisades Nuclear Plant - Issuance of Amendment RE: Primary Coolant System Pressure-Temperature Limits (TAC No. ME5806)," dated January 19, 2012. [ADAMS Accession Number ML113480303] 9. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. 10. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 of LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007. 11. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835. 12. U.S. NRC Technical Letter Report TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research [RES], dated November 14, 2014. [ADAMS Accession Number ML14318A177]. 13. ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.
Westinghouse Non-Proprietary Class 3 Page 21 of 21 NP-Attachment MCOE-LTR-16-33, Rev. 0 August 22, 2016 14. Oak Ridge National Laboratory Report, ORNL/TM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles - Revision 1," June 2012. 15. ASME B&PV Code,Section III, Division I, Subarticle NB-2332, "Material for Piping, Pumps, and Valves, Excluding Bolting Material."