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Category:Calculation
MONTHYEARRA-20-0328, Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals2021-01-19019 January 2021 Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals RA-20-0280, Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw2020-09-14014 September 2020 Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw RA-19-0211, 2018 Annual Radioactive Effluent Release Report (ARERR) and Offsite Dose Calculation Manual (ODCM)2019-04-30030 April 2019 2018 Annual Radioactive Effluent Release Report (ARERR) and Offsite Dose Calculation Manual (ODCM) ONS-2018-016, Calculation 0079-0191-CALC-002, Induced Differential Voltage in Control Cables.2018-02-12012 February 2018 Calculation 0079-0191-CALC-002, Induced Differential Voltage in Control Cables. ONS-2016-039, Offsite Dose Calculation Manual, Revision 272016-03-0707 March 2016 Offsite Dose Calculation Manual, Revision 27 ML15138A1252015-03-16016 March 2015 Offsite Dose Calculation Manual (Odcm), Revision 56 ONS-2015-043, Oconee, Units 1, 2, and 3 - Offsite Dose Calculation Manual (Odcm), Revision 562015-03-16016 March 2015 Oconee, Units 1, 2, and 3 - Offsite Dose Calculation Manual (Odcm), Revision 56 ML14346A0092014-08-13013 August 2014 Calculation 13922-0402, Revision 0, Helium Leak Rate Evaluation for 24PHB Dsc. ML14346A0102014-04-0202 April 2014 Calculation 1098-8, Revision 3, Helium Leak Testing Vent and Siphon Ports for Leak Tight Nuhoms Dscs. ML13305A1222013-11-25025 November 2013 Enclosure 3 - Areva Document No. 32-9212003-000, Lst Considering RCS Piping and RVCH Limit for B&W Plant Designs, September 2012 ONS-2014-044, Revision 54 to the Offsite Dose Calculation Manual (OCDM)2013-10-10010 October 2013 Revision 54 to the Offsite Dose Calculation Manual (OCDM) ML13214A0622013-03-14014 March 2013 Offsite Dose Calculation Manual, Revision 53 ML12125A3252012-03-31031 March 2012 2011 Offsite Dose Calculation Manual Compact Disc ML12080A1992012-03-0101 March 2012 Tornado and High Energy Line Break Mitigation License Amendment Requests - Supplemental Responses to Request for Additional Information Nos. 70, 76, and 106 ML12080A2002012-02-29029 February 2012 Calculation OSC-10008, Revision 1, Calculation Impact Assessment ML12080A2022012-02-21021 February 2012 Calculation OSC-9510, Revision 1, FMEA for Ssf 4.16kV Alternate Power Feed from Psw EC91876 (OD500928) ML11124A1262011-04-28028 April 2011 Response to RAI, License Amendment Request to Change Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles License Amendment Request (LAR) No. 2010-001, Supplement 2 ML11129A2962011-01-31031 January 2011 Offsite Dose Calculation Manual 2011 ML1012604742010-04-30030 April 2010 Offsite Dose Calculation Manual, Revision 50 ML11223A0122009-09-22022 September 2009 Calculation OSC-9863, Oconee, Units 1, 2 and 3 - License Exemption Using BAW-2308 ML0912806652009-01-31031 January 2009 Offsite Dose Calculation Manual, Revision 49 ML0910402142008-10-28028 October 2008 Calculation No. OSC-9314, Rev. 0, NFPA 805 Transition Risk-Informed, Performance-Based Change Evaluation Methodology. ML0910402062008-05-29029 May 2008 Calculation No. OSC-9291, Rev. 0, NFPA 805 Transition B-2 Table. ML0910402082008-05-28028 May 2008 Calculation No. OSC-9295, Rev. 1, NFPA 805 Transition B-1 Table/Report. ML0910402072008-05-23023 May 2008 Calculation No. OSC-9293, Rev. 0, NFPA 805 Transition Radioactive Release G-1 Table. ML0814305062008-01-31031 January 2008 Units, 1, 2 and 3 - 2008 Oconee Offsite Dose Calculation Manual, Revision 48 ML0715810232007-01-31031 January 2007 Rev. 47 to Offsite Dose Calculation Manual (ODCM) ML0611800782006-03-27027 March 2006 Unit 3 Calculations for Oconee TAC MC8127 ML0612104022006-01-31031 January 2006 Offsite Dose Calculation Manual (Odcm), Revision No. 46 ML0910503332005-03-0808 March 2005 Calculation OSC-8623, Rev. 0, RPS & ESFAS System Functional Description for Areva Teleperm Xs. ML0612206472005-02-0404 February 2005 Offsite Dose Calculation Manual, Revision 17 ML0412404352004-01-31031 January 2004 Units 1, 2, and 3 - ODCM Offsite Dose Calculation Manual 2004 2021-01-19
[Table view] Category:Letter type:RA
MONTHYEARRA-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-23-0320, 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-23-0318, 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-23-0182, 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-23-0284, 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-23-0304, 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-23-0281, 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-23-0268, 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-23-0225, 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-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-23-0211, 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-23-0186, 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-23-0184, 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-23-0146, 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-23-0135, 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-23-0041, 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-23-0111, 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 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-23-0047, Duke Energy Annual Radiological Environmental Operating Report - 20222023-04-26026 April 2023 Duke Energy Annual Radiological Environmental Operating Report - 2022 RA-23-0044, Aduke Energy Annual Report of Changes Pursuant to 10 CFR 50.462023-04-26026 April 2023 Aduke Energy Annual Report of Changes Pursuant to 10 CFR 50.46 RA-23-0099, Subsequent License Renewal Application Second Annual Amendment to the License Renewal Application2023-04-25025 April 2023 Subsequent License Renewal Application Second Annual Amendment to the License Renewal Application RA-23-0046, Annual Radioactive Effluent Release Report - 20222023-04-24024 April 2023 Annual Radioactive Effluent Release Report - 2022 RA-22-0292, 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 Frequencies2023-04-0404 April 2023 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 RA-23-0036, Biennial Decommissioning Financial Assurance Reports2023-03-30030 March 2023 Biennial Decommissioning Financial Assurance Reports RA-23-0040, Onsite Property Insurance Coverage2023-03-30030 March 2023 Onsite Property Insurance Coverage RA-23-0039, 10 CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums2023-03-30030 March 2023 10 CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums RA-23-0037, Duke Energy - Decommissioning Funding Plan for Independent Spent Fuel Storage Installations (Isfsis)2023-03-30030 March 2023 Duke Energy - Decommissioning Funding Plan for Independent Spent Fuel Storage Installations (Isfsis) 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-23-0052, Refuel 32 (O1R32) Inservice Inspection (ISI) Report, Fifth 10-Year ISI Interval2023-02-22022 February 2023 Refuel 32 (O1R32) Inservice Inspection (ISI) Report, Fifth 10-Year ISI Interval 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) RA-22-0091, Application to Revise Technical Specifications to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements2023-02-16016 February 2023 Application to Revise Technical Specifications to Adopt TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements RA-23-0001, Request to Use a Provision of a Later Edition and Addenda of the ASME Boiler and Pressure Vessel Code, Section XI2023-02-0202 February 2023 Request to Use a Provision of a Later Edition and Addenda of the ASME Boiler and Pressure Vessel Code, Section XI RA-22-0118, License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies2023-02-0101 February 2023 License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies RA-22-0256, Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)2023-01-23023 January 2023 Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) RA-22-0335, Submittal of 30-Day Report Per 10CFR26.719(c)(1) - Unsatisfactory Performance of a Health & Human Services Certified Lab2022-12-0505 December 2022 Submittal of 30-Day Report Per 10CFR26.719(c)(1) - Unsatisfactory Performance of a Health & Human Services Certified Lab RA-22-0334, Refuel 31 (O3R31) Steam Generator Tube Inspection Report2022-11-21021 November 2022 Refuel 31 (O3R31) Steam Generator Tube Inspection Report RA-22-0329, Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 1 Cycle 33 (Revision 0 and Revision 1)2022-11-11011 November 2022 Transmittal of Core Operating Limits Report (COLR) for Oconee Unit 1 Cycle 33 (Revision 0 and Revision 1) RA-22-0285, Subsequent License Renewal - Appendix E Environmental Report Supplement 22022-11-0707 November 2022 Subsequent License Renewal - Appendix E Environmental Report Supplement 2 RA-22-0313, Emergency Action Level (EAL) Technical Basis Document, Revision 4, EAL Wallchart, Revision 3 and Alternate TSC and OSC Renovations2022-10-27027 October 2022 Emergency Action Level (EAL) Technical Basis Document, Revision 4, EAL Wallchart, Revision 3 and Alternate TSC and OSC Renovations 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-22-0192, 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 2024-01-08
[Table view] Category:Report
MONTHYEARML23270B8362023-09-26026 September 2023 Code Case N-752 Audit September 26, 2023, E-mail Providing Additional Information Regarding the Use of Owner'S Requirements and Engineering Judgment in Lieu of Code and Standards ML22269A3882022-09-0707 September 2022 02-Gamma Spectroscopy of Concrete Pucks RA-20-0280, Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw2020-09-14014 September 2020 Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw ML19242C7282019-08-28028 August 2019 Attachments 5 - 12: Thermal-Hydraulic Models for High Energy Line Break Transient Analysis, Duke Energy/Framatome Affidavits, Regulatory Requirements, Definitions, Time Critical Operator Actions and Feasibility Assessment for New Proposed T RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 ML17031A4312017-01-26026 January 2017 Notification of Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events and FLEX Final Integrated Plan ML16230A2822016-08-17017 August 2016 Duke Power Company Oconee Nuclear Station Unit 2 Integrated Leak Rate Test of the Reactor Containment Building ML16152A0522016-05-31031 May 2016 ANP-3477NP, Rev. O, MRP-227-A Applicant/Licensee Action Item 6 Analysis ONS-2016-017, Request for Alternative to Codes and Standards Requirements Pursuant to 10 CFR 50.55a(z) to Satisfy 10 CFR 50.55a(h)(2)2016-02-15015 February 2016 Request for Alternative to Codes and Standards Requirements Pursuant to 10 CFR 50.55a(z) to Satisfy 10 CFR 50.55a(h)(2) ML16034A3252015-12-31031 December 2015 Revision 0 to Summary of Methodology Used for Seismic Capacity Evaluation of GMRS Mitigation Components ML15218A2972015-08-0606 August 2015 Historic Temperature Data Including Record Highs for the Seneca ML15201A0082015-07-22022 July 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review ML15161A4202015-06-15015 June 2015 Review of the Spring 2014 Steam Generator Tube Inservice Inspection Report During Refueling Outage 27 ONS-2015-065, TIA 2014-04, Request for Technical Assistance Regarding the Adequacy of the Station Design and Licensing Bases for the Degraded Voltage Relay2015-05-22022 May 2015 TIA 2014-04, Request for Technical Assistance Regarding the Adequacy of the Station Design and Licensing Bases for the Degraded Voltage Relay ONS-2015-058, Engineering Report 555, Oconee, Units 1, 2, and 3 - Ul 1569 Impact and Crush Tests on Keowee Underground Trench Power Cables for Duke Energy Carolinas, Llc.2015-03-27027 March 2015 Engineering Report 555, Oconee, Units 1, 2, and 3 - Ul 1569 Impact and Crush Tests on Keowee Underground Trench Power Cables for Duke Energy Carolinas, Llc. ML16272A2172015-03-0606 March 2015 Revision 1 to Flood Hazard Reevaluation Report on Oconee Nuclear Station, Pp. 1-70, Dated March 6, 2015 (Redacted) ML16272A2182015-03-0606 March 2015 Revision 1 to Flood Hazard Reevaluation Report on Oconee Nuclear Station, Appendices, Dated March 6, 2015 (Redacted) ML15072A2172015-02-25025 February 2015 Form OAR-1 Owners Activity Report (CC N-532-4) for Oconee Unit 1, Refueling Outage EOC 28 ONS-2014-161, Submittal of the Expedited Seismic Evaluation Process Report (CEUS Sites)2014-12-19019 December 2014 Submittal of the Expedited Seismic Evaluation Process Report (CEUS Sites) ML14218A8072014-08-27027 August 2014 NRC Staff Review of the Documentation Provided by Duke Energy for the Oconee Nuclear Station Concerning Resolution of GL 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Wat ONS-2014-074, Special Report Per Selected Licensee Commitment 16.9.9 Auxiliary Service Water (Asw) System and Main Steam Atmospheric Dump Valves and Oconee Transition from the Station Asw System to the Protected Service Water System2014-08-15015 August 2014 Special Report Per Selected Licensee Commitment 16.9.9 Auxiliary Service Water (Asw) System and Main Steam Atmospheric Dump Valves and Oconee Transition from the Station Asw System to the Protected Service Water System ML16272A2162014-04-30030 April 2014 Validation of Hrr Breach Hydrograph for Jocassee Dam, Dated April 30, 2014 (Redacted) ML15065A3372014-04-23023 April 2014 Response to NRC Request for Additional Information (RAI) Related to the MRP-227-A Applicant/Licensee Action Item #7 Analysis Supporting the Reactor Vessel Internals Inspection Plan Amendment Request (ANP-3267Q1NP (R1)) ML14076A0002014-04-0202 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident ONS-2014-040, 2013 Annual Commitment Change Report2014-03-31031 March 2014 2013 Annual Commitment Change Report ONS-2014-046, Seismic Hazard and Screening Report (CEUS Sites), 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..2014-03-31031 March 2014 Seismic Hazard and Screening Report (CEUS Sites), 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.. ML14087A1752014-03-20020 March 2014 ANP-3267Q1NP, Rev. 0, Response to NRC Request for Additional Information (RAI) Related to the MRP-227-A Applicant/Licensee Action Item #7 Analysis Supporting the Reactor Vessel Internals Inspection Plan Amendment Request for Oconee, Units 1 ML13365A2582014-02-10010 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14028A5122014-02-0404 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Oconee Nuclear Station, Units 1, 2, and 3, TAC Nos.: MF0782, MF0783, and MF0784 ONS-2014-002, Special Report Per Technical Specification 5.6.6, Problem Investigation Process No. O-13-144472014-01-16016 January 2014 Special Report Per Technical Specification 5.6.6, Problem Investigation Process No. O-13-14447 ML13339A3482013-11-30030 November 2013 ANP-3267NP, Revision 0,MRP-227-A Applicant/Licensee Action Item #7 Analysis for the Oconee Nuclear Station Units, Licensing Report ML13275A3002013-08-31031 August 2013 ANP-3208NP, Rev 1, Confirmation of Stress Relief for ONS-1, ONS-2, ONS-3, and CR-3 Core Support Shield Upper Flange Welds ML13275A2992013-08-31031 August 2013 ANP-3186NP, Rev 2, ONS License Renewal Scope and MRP-189, Rev. 1 Comparison ML13196A2482013-07-11011 July 2013 Duke Energy Carolinas, LLC (Duke Energy): Report Pursuant to 10 CFR 50.46, Changes to or Errors in an Evaluation Model ML13190A0152013-07-0101 July 2013 Aging Management Reviews for Newly Identified Systems, Structures and Components as of December 31, 2012 ML13192A1572013-05-15015 May 2013 Stone & Webster, Inc., Technical Report 1457690202-R-M-00005-0, NTTF 2.3 Seismic Peer Review Supplementary Report, Oconee Nuclear Station Unit 1. ML13119A0412013-03-31031 March 2013 Attachment 1, ANP-3186NP, Rev. 1, ONS License Renewal Scope and MRP-189, Rev. 1 Comparison. ML13119A0422013-03-31031 March 2013 Attachment 2, ANP-3208NP, Rev. 0, Confirmation and Stress Relief for ONS-1, ONS-2, ONS-3, and CR-3 Core Support Shield Upper Flange Welds. ML12347A2522012-11-27027 November 2012 Seismic Walkdown Information Request for Information Pursuant to Title 10 of Code of Federal Regulations Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident 3/12/12 ML16272A2202012-11-27027 November 2012 Enclosure 1: Flood Walkdown Report (NRC 50.54(f) NTTF Recommendation 2.3), Dated November 27, 2012 (Redacted) ML12347A2532012-11-16016 November 2012 Attachment 6: Nttp 2.3 Seismic Peer Review Report Oconee Nuclear Station Units 1,2 and 3 ML14058A0452012-09-20020 September 2012 History of Oconee Flood Concerns ML12251A0312012-09-10010 September 2012 Review of the Spring 2011 Steam Generator Tube Inservice Inspections During End of Cycle 26 Refueling Outage ML12251A0042012-09-0505 September 2012 Unit 3 Cycle 27 Startup Testing Report ML12187A2152012-03-31031 March 2012 Attachment 1, ANP-2951, Rev. 002, Inspection Plan for the Oconee Nuclear Station, Units 1, 2, and 3, Reactor Vessel Internals, Amp/Application to Implementation MRP-227-A ML11180A0112011-06-23023 June 2011 Closure of 60-Day Interim Report Notification: Dubose National Energy Services, Inc ML1108208652011-03-31031 March 2011 Evaluation of 2010 (Cycle 24) Steam Generator (SG) Tube Inspections ML1100404092010-12-23023 December 2010 Draft, Regarding Review of Root Cause Analysis and Actions Addressing the Underground Pipe Damage to the Condensate Storage Tank Return Line ML12053A3332010-10-27027 October 2010 Areva Document No. 47-9048125-002, Update of Irradiation Embrittlement in BAW-10008 Part 1 Rev 1, Non-Proprietary ML1029900642010-10-26026 October 2010 NRC Staff Assessment of Duke Energy Carolinas, Llc. Oconee External Flooding Issue 2023-09-26
[Table view] Category:Technical
MONTHYEARML22269A3882022-09-0707 September 2022 02-Gamma Spectroscopy of Concrete Pucks RA-20-0280, Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw2020-09-14014 September 2020 Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw ML16230A2822016-08-17017 August 2016 Duke Power Company Oconee Nuclear Station Unit 2 Integrated Leak Rate Test of the Reactor Containment Building ML16034A3252015-12-31031 December 2015 Revision 0 to Summary of Methodology Used for Seismic Capacity Evaluation of GMRS Mitigation Components ONS-2015-065, TIA 2014-04, Request for Technical Assistance Regarding the Adequacy of the Station Design and Licensing Bases for the Degraded Voltage Relay2015-05-22022 May 2015 TIA 2014-04, Request for Technical Assistance Regarding the Adequacy of the Station Design and Licensing Bases for the Degraded Voltage Relay ML16272A2182015-03-0606 March 2015 Revision 1 to Flood Hazard Reevaluation Report on Oconee Nuclear Station, Appendices, Dated March 6, 2015 (Redacted) ML16272A2172015-03-0606 March 2015 Revision 1 to Flood Hazard Reevaluation Report on Oconee Nuclear Station, Pp. 1-70, Dated March 6, 2015 (Redacted) ML16272A2162014-04-30030 April 2014 Validation of Hrr Breach Hydrograph for Jocassee Dam, Dated April 30, 2014 (Redacted) ML15065A3372014-04-23023 April 2014 Response to NRC Request for Additional Information (RAI) Related to the MRP-227-A Applicant/Licensee Action Item #7 Analysis Supporting the Reactor Vessel Internals Inspection Plan Amendment Request (ANP-3267Q1NP (R1)) ONS-2014-046, Seismic Hazard and Screening Report (CEUS Sites), 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..2014-03-31031 March 2014 Seismic Hazard and Screening Report (CEUS Sites), 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.. ML14087A1752014-03-20020 March 2014 ANP-3267Q1NP, Rev. 0, Response to NRC Request for Additional Information (RAI) Related to the MRP-227-A Applicant/Licensee Action Item #7 Analysis Supporting the Reactor Vessel Internals Inspection Plan Amendment Request for Oconee, Units 1 ML13365A2582014-02-10010 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14028A5122014-02-0404 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Oconee Nuclear Station, Units 1, 2, and 3, TAC Nos.: MF0782, MF0783, and MF0784 ML13339A3482013-11-30030 November 2013 ANP-3267NP, Revision 0,MRP-227-A Applicant/Licensee Action Item #7 Analysis for the Oconee Nuclear Station Units, Licensing Report ML13275A3002013-08-31031 August 2013 ANP-3208NP, Rev 1, Confirmation of Stress Relief for ONS-1, ONS-2, ONS-3, and CR-3 Core Support Shield Upper Flange Welds ML13275A2992013-08-31031 August 2013 ANP-3186NP, Rev 2, ONS License Renewal Scope and MRP-189, Rev. 1 Comparison ML13192A1572013-05-15015 May 2013 Stone & Webster, Inc., Technical Report 1457690202-R-M-00005-0, NTTF 2.3 Seismic Peer Review Supplementary Report, Oconee Nuclear Station Unit 1. ML13119A0422013-03-31031 March 2013 Attachment 2, ANP-3208NP, Rev. 0, Confirmation and Stress Relief for ONS-1, ONS-2, ONS-3, and CR-3 Core Support Shield Upper Flange Welds. ML13119A0412013-03-31031 March 2013 Attachment 1, ANP-3186NP, Rev. 1, ONS License Renewal Scope and MRP-189, Rev. 1 Comparison. ML12347A2522012-11-27027 November 2012 Seismic Walkdown Information Request for Information Pursuant to Title 10 of Code of Federal Regulations Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident 3/12/12 ML16272A2202012-11-27027 November 2012 Enclosure 1: Flood Walkdown Report (NRC 50.54(f) NTTF Recommendation 2.3), Dated November 27, 2012 (Redacted) ML12347A2532012-11-16016 November 2012 Attachment 6: Nttp 2.3 Seismic Peer Review Report Oconee Nuclear Station Units 1,2 and 3 ML12251A0042012-09-0505 September 2012 Unit 3 Cycle 27 Startup Testing Report ML12053A3332010-10-27027 October 2010 Areva Document No. 47-9048125-002, Update of Irradiation Embrittlement in BAW-10008 Part 1 Rev 1, Non-Proprietary ML1029900642010-10-26026 October 2010 NRC Staff Assessment of Duke Energy Carolinas, Llc. Oconee External Flooding Issue ML0910503322009-04-0303 April 2009 Encl. 5 to Supplemental Request for Additional Information for License Amendment Request for Reactor Protective System/Engineered Safeguards Protective System Digital Upgrade, Technical Specification Change Number 2007-09 ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 ML0809406392008-04-10010 April 2008 Preliminary Results of the NRC Staff Review of the Fire Probabilistic Risk Assessment Model to Support Implementation of National Fire Protection Association Standard NFPA-805, Performance Based Standard for Fire.. ML0516403072005-07-22022 July 2005 Third 10-Year Interval Inservice Inspection Program Plan Requests for Relief 04-ON-002 and 04-ON-003 ML0505600372005-02-0808 February 2005 12-5023272-01, Risk Assessment for Alternate STS End States. ML0429603912004-10-15015 October 2004 ORNL/NRC/LTR-04/18, Electronic Archival of the Results of Pressurized Thermal Shock Analyses for Beaver Valley, Oconee, and Palisades Reactor Pressure Vessels Generated with the 04.1 Version of Favor. ML0435502712004-03-31031 March 2004 CTRS-Report, Thermal - Hydraulic Uncertainty Analysis in Pressurized Thermal Shock Risk Assessment. ML0415507862004-02-29029 February 2004 FRA-ANP Document No. 77-5038568-00, LBB Evaluation of the Core Flood and Low Pressure Injection/Decay Heat Removal Piping Systems for Oconee Unit 3. ML0327315572003-09-29029 September 2003 Technical Letter Report on the Third 10-Year Interval Inservice Inspection Request for Relief Nos. 02-004 and 02-005 for Oconee Nuclear Station, Unit 1 ML0306507162003-02-25025 February 2003 Special Report Per Technical Specification 5.6.6., Problem Investigation Process No.: 0-03-0154, 0-03-0415 ML0234703082002-12-0505 December 2002 Steam Generator Inservice Inspection, Steam Generator Tube Plugging & Repair 30-Day Report ML0716501681999-04-0101 April 1999 NEI Report: RCP Lube Oil Collection System Current Trend Abstract, April 1999 (Revision 1) NRC Generic Letter 1979-451979-09-25025 September 1979 NRC Generic Letter 1979-045: Transmittal of Reports Regarding Foreign Reactor Operation Experiences 2022-09-07
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Text
Steve Snider
( ., DUKE Vice President ENERGY Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 Steve.Snider@duke-energy.com RA-20-0280 September 14, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oconee Nuclear Station, Unit 3 Docket No. 50-287 Renewed License No. DPR-55
Subject:
Oconee Unit 3, Submittal of IWB-3640 Analytical Evaluation Performed to Accept an ASME Section XI Code Rejected Sub-Surface Flaw Pursuant to the 2007 Edition through the 2008 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, IWB-3134, Duke Energy is submitting an Analytical Evaluation performed to accept an ASME code rejected sub-surface flaw identified on the 3B1 high pressure injection (HPI) nozzle-to-safe end weld during the 28 th refueling outage (O3R28) for Oconee Nuclear Station (ONS), Unit 3 in the Spring 2016. The May 2016 Analytical Evaluation for the code rejected flaw was performed in accordance with the 2007 Edition through the 2008 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, IWB-3640 and is being submitted for information only.
The Enclosure contains the HPI nozzle weld flaw evaluation per ASME Code,Section XI, IWB-3640 requirements.
This submittal contains no regulatory commitments. Please refer any questions regarding this submittal to Art Zaremba, Manager - Nuclear Fleet Licensing, at 980-373-2062.
Sincerely, Steve Snider Vice President - Nuclear Engineering
Enclosure:
High Pressure Injection (HPI) Nozzle Weld 3-RC-212-53V Flaw Evaluation
U.S. Nuclear Regulatory Commission Page 2 RA-20-0280 cc: (w/ Enclosure)
Ms. Laura Dudes Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Shawn Williams NRC Project Manager Oconee Nuclear Station Mr. Jared Nadel NRC Senior Resident Inspector Oconee Nuclear Station
Enclosure to RA-20-0280 Enclosure High Pressure Injection (HPI) Nozzle Weld 3-RC-212-53V Flaw Evaluation
I; Structural Integrity Associates, Inc. File No.: 1600492.301 Project No.: 1600492 CALCULATION PACKAGE Quality Program Type: 1:8] Nuclear
Oconee Unit 3 HPI IWB-3600 Flaw Evaluation CONTRACT NO.:
03021365 00001 CLIENT: PLANT:
Duke Energy Oconee Nuclear Generating Station, Unit 3 CALCULATION TITLE:
HPI Nozzle Weld 3-RC-212-53V Flaw Evaluation per ASME Code,Section XI, IWB-3640 Requirements Project Manager Preparer(s) &
Document Affected Revision Description Approval Checker(s)
Revision Pages Signature & Date Signatures & Date 0 1 - 12 Initial Issue Preparers:
A A-2 Chris Lohse Wilson Wong 5/9/16 5/9/16 Checkers:
Kevin L. Wong 5/9/16 Page 1 of 12 F0306-01R2
SJ Structural Integrity Associates, lnc.,s; Table of Contents
1.0 INTRODUCTION
.........................................................................................................3 2.0 TECHNICAL APPROACH ..........................................................................................3 3.0 ASSUMPTIONS AND DESIGN INPUTS ...................................................................4 3.1 Piping Loads and Load Combinations...............................................................4 3.2 Thermal Transients ............................................................................................5 4.0 EVALUATION .............................................................................................................5 4.1 Allowable Flaw Size Determination..................................................................5 4.2 Crack Growth Consideration .............................................................................7
5.0 CONCLUSION
..............................................................................................................8
6.0 REFERENCES
..............................................................................................................9 APPENDIX A PHASED ARRAY ULTRASONIC EXAMINATION RESULTS [3, PDF PG. 4].................................................................................................... A-1 List of Tables Table 1: Loads at Flaw Location [9].......................................................................................10 Table 2: Stress Determination at Indications Location...........................................................10 Table 3: Stress Ratios and Allowable Flaw Depth-to-Thickness Ratios for a Non-Flux Weld Using Limit Load Analysis ..........................................................................11 Table 4: Material Properties for Alloy 600 (N06600) [10] .....................................................11 Table 5: Fatigue Crack Growth Transients and Cycles ..........................................................12 File No.: 1600492.301 Page 2 of 12 Revision: 0 F0306-01R2
SJ Structural Integrity Associates, lnc.,s;
1.0 INTRODUCTION
During phased array ultrasonic testing (PAUT) of Weld 3-RC-212-53V on the 3B1 HPI nozzle at Oconee Unit 3, a planar sub-surface flaw was identified in the Alloy 82 nozzle-to-safe end weld [3].
This indication was previously identified in 2014 and found to be acceptable [2]. However, using a higher resolution technique in 2016, the indication now measures larger than the acceptance standards as defined by ASME Code Section XI, IWB-3500 [1].
The indication was measured with several different angles with varying sizing dimensions. As such, the maximum flaw length (1.47) and maximum flaw through-wall depth (0.19) measured from any angle will be used for the flaw evaluation [3].
The objective of this calculation is to perform a flaw evaluation to the requirements of ASME Code,Section XI, IWB-3640 [1] considering the stresses and material properties at the location of the indication to determine the maximum Code allowable flaw size and acceptability for continued operation due to crack growth. The Section XI Code-of-record for Oconee is the 2007 Edition with 2008 Addenda [4].
2.0 TECHNICAL APPROACH The following technical approach will be used to perform the flaw evaluation.
- 1. Determine the stresses at the flaw location. Since the indications are in the circumferential direction [3], the stresses required are the axial stresses.
- 2. Determine the allowable flaw size using ASME Code,Section XI rules for austenitic stainless steel piping in IWB-3640 and Appendix C. The 2007 Edition of the ASME Code,Section XI states that the evaluation procedures and acceptance criteria in Nonmandatory Appendix C are applicable to piping 4 inch NPS or larger [1]. The 2013 Edition of ASME Code,Section XI [5]
states that the evaluation procedures and acceptance criteria in Nonmandatory Appendix C are applicable to piping 1 inch NPS or larger. There are no differences in the formulas, margins, or acceptance criteria in Nonmandatory Appendix C between the two versions, and as such, the 2007 Edition with 2008 Addenda can be used for the evaluation of the HPI Nozzle-to-safe end weld.
- 3. Assess the crack growth to ensure the as-found flaw will not reach the allowable flaw size determined in Step 2 before the end of the 10 year evaluation period. Since the flaw is a subsurface flaw as defined below in Section 4.0, there is no stress corrosion crack growth and only fatigue crack growth (FCG) is considered.
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SJ Structural Integrity Associates, lnc.,s; 3.0 ASSUMPTIONS AND DESIGN INPUTS Geometry (Dimensions taken from nozzle to safe end weld)
Pipe outside diameter (OD) = 3.5 inches [6, Figure 1]
Pipe inside diameter (ID) = 2 inches [6, Figure 1, with cladding]
Pipe thickness = 0.75 inches [6, Figure 1, with cladding]
Note that the PAUT lists a component thickness of 0.77 inches, but the use of 0.75 inches for the thickness is conservative for calculation of stresses.
Materials The nozzle to safe end weld is Alloy 82 weld metal [7] with elastic material properties used for finite element analysis shown in Table 4.
Operating Conditions Max Transient Internal Pressure = 3,192 psig [14, Table 4]
Design Temperature = 650°F [8, PDF pg36]
3.1 Piping Loads and Load Combinations The nozzle piping loads used in the evaluation are obtained from Reference [9] at the Alloy 600 weld.
These loads are summarized in Table 1. Safe Shutdown Earthquake is assumed to be 2X the Operating Basis Earthquake. The following Service Level load combinations are assumed to perform the flaw evaluation:
Primary Loads:
Service Level A/B (Normal/Upset): Deadweight (DW) + Operating Basis Earthquake (OBE)
Service Level C/D (Emergency/Faulted): DW + Safe Shutdown Earthquake (SSE)
From Table 1, the resultant moments for the primary and thermal loads for the various Service Levels are shown below. Even though only axial stress is needed, the moments are conservatively calculated using the SRSS of all three directions. Service Level B loads are conservatively used for both Service Level A and B.
Service Level A/B (Normal/Upset Condition, N/U) 7,182 in-lbs.
Service Level C/D (Emergency/Faulted Condition, E/F) 10,420 in-lbs.
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SJ Structural Integrity Associates, lnc.,s; 3.2 Thermal Transients In order to perform FCG analysis, stresses from thermal transients must be obtained. A previous SI calculation package [11] analyzed thermal transients on the Oconee HPI nozzle, but used material properties for Alloy 690 on the nozzle to safe end weld since the weld was changed to Alloy 52. Since the nozzle to safe end weld in the 3B1 nozzle in Unit 3 is Alloy 82, the material properties were modified to the values shown in Table 4. The model was regenerated with the new material properties using the HPI_GEOM.INP input file [11] in ANSYS 8.1 [12]. All thermal transient analysis (transient input files taken directly from [11]) and post processing was performed in ANSYS 14.5 [13] to take advantage of faster solving and post processing capabilities. First order stress coefficients (linear) are generated from mapped stresses through the weld (Path 2 in Reference [11]) using the GenStress.mac SI macro for use in FCG analysis. The GETPATH.TXT file contains the path information for stress extraction. The stress coefficients are output to *_STRS_COE_P1.CSV files, where
- is the transient number. The number of cycles for each transient for 60 years of operation are obtained from Reference
[14, Table 4], and shown in Table 5. These are scaled to 10 years by dividing the values by 6.
4.0 EVALUATION Flaw Characterization Based on the exam results in Appendix A, the following bounding flaw dimensions will be used for Flaw No. 1:
Maximum flaw length (l) = 1.47 inches Maximum flaw depth (2a) = 0.19 inch (a = 0.095 inch)
Flaw depth-to-thickness ratio (a/t) = 0.095/0.75 = 0.1266 or 12.67%
Flaw aspect ratio (all) = 0.095/1.47 = 0.065 Bottom of flaw to inside surface (S) = 0.06 Type of flaw: 0.06 (S) > 0.038 (0.4a), subsurface flaw Determination of Stresses The bending stresses were calculated from the bending moments shown in Table 1 using the equation from C-2500 [1]. Pressure stresses were also calculated based on the maximum pressure in Section 3.0 and using the equation from C-2500 [1]. The resulting stresses for the various load combinations are shown in Table 2.
4.1 Allowable Flaw Size Determination For a non-flux weld [7], Figure C-4210-1 of ASME Code,Section XI, Appendix C, allows the use of limit load techniques of C-5000 to determine the allowable flaw size [1].
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SJ Structural Integrity Associates, lnc.,s; For a limit load evaluation of a circumferential flaw, Article C-5000 points to Tables C-5310-1 through C-5310-5, which provide the allowable flaw depth-to-thickness ratios (a/t) as a function of the stress ratios and ratio of flaw length-to-pipe circumference ( f/ D). These tables are only valid if primary membrane stress is less than a factor times the material flow stress.
The stress ratio (SR) for limit load is defined as:
SR = (crm +crb)/crf where:
O'm = primary membrane stress O'b = primary bending stress O'f = flow stress = 54.85 ksi = average of yield (cry) and ultimate tensile stress (cru) [1, C-8200]
O'y = yield stress at operating temperature, 29.7 ksi [10]
O'u = ultimate tensile stress at operating temperature, 80.0 ksi [10]
OD = 3.5 inches [6]
From Table 2, it can be seen that the primary membrane stress is less than 0.2 times the flow stress, or 10.97 ksi. Since the same pressure is used for all service levels and 0.2 is the most limiting factor, all service level tables are applicable. The determination of the stress ratios at the location of the flaw is shown in Table 3 for Service Levels A/B and C/D.
For Service Levels A and B O'm = pressure axial stress O'b = bending stress due to DW + seismic OBE For Service Level C and D O'm = pressure axial stress O'b = bending stress due to DW + seismic SSE Using the maximum flaw length of 1.47 inches, the ratio of flaw length to pipe circumference is l.47/(1t*3 .5) = 0.134. However, allowable values for a full circumferential crack are used to bound any possible crack growth. From the stress ratios for limit load analysis shown in Table 3 and the 0.75 ratio of flaw length to pipe circumference ( f/ D), the allowable flaw depth to thickness ratios are determined from Tables C-5310-1 through C-5310-5 of ASME Code,Section XI, and also shown in Table 3. Since the maximum stress ratio is less than 0.2, Table C-5310-5 for pure membrane stress allows a flaw depth-to-thickness ratio of 0.75 even for a full circumferential crack. The minimum allowable flaw depth-to-thickness ratio is 0.73 considering all Service Levels for a full circumferential crack, bounding the measured flaw length of 1.47 inches. This is equivalent to a total flaw depth (2a) of 0.73*0.75 = 0.54 inches.
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SJ Structural Integrity Associates, lnc.,s; 4.2 Crack Growth Consideration The pc-CRACK [15] software is used to compute FCG with the built in ASME elliptical subsurface crack model (model 208). The load cases and corresponding number of cycles evaluated are listed in Table 5.
The FCG law for austenitic steels in air is used in the FCG calculation (see Equation 1) since the flaw is a subsurface crack. This crack growth law is built into pc-CRACK and is based on the crack growth law for austenitic steels published in C-8410, Appendix C of Section XI of the ASME Code [1].
da
= Co (!1K )n (1) dN where:
Co = CS log10 C = 10.009 + 8.12 x10 4 T 1.13 x10 6 T 2 +1.02 x10 9 T 3 1 R ,:::; 0 S 1 1 .8 R 0 < R ,:::; 0.79 43.35 + 57.97 R R < 0.79 R = K min / K max
~K = K max K min n = 3.3 T is temperature in oF, which is taken to be the maximum temperature during the transient. The above constants provide crack growth rates in inches per cycle when K is in ksi-in1/2.
To account for weld residual stress, a constant membrane stress equal to the material yield strength of Alloy 82/182 at 650ºF (29.7 ksi) is conservatively applied. Due to variation of the curve fit for stresses that arent linear, 5 ksi was added to the maximum stress for each transient as a membrane stress.
The maximum thermal piping load listed in Table 1 was obtained by taking the largest absolute value for each thermal component from Reference [9]. The SRSS of the bounding thermal moments listed in Table 1 generate a bending stress of 8.8 ksi, which was conservatively added to the maximum stress of each transient as a membrane stress.
The file CrackGrowth.pcf contains the pc-CRACK input, and the file CrackGrowth.rpt contains the output with resulting crack growth. The resulting crack growth for a subsurface flaw with 1.47 inch File No.: 1600492.301 Page 7 of 12 Revision: 0 F0306-01R2
SJ Structural Integrity Associates, lnc.,s; length and 0.19 inch total height is a = 0.0097 over a 10 year evaluation period. The final flaw depth after FCG is 0.19+ 2*0.0097 = 0.2094. At this depth, the flaw remains classified as a subsurface flaw where 0.0503 (S) > 0.0419 (0.4a).
5.0 CONCLUSION
A flaw evaluation was performed to the requirements of ASME Code,Section XI, IWB-3640 [1]
considering the stresses and material properties at the location of the indications to determine the Code allowable flaw size and acceptability after crack growth. The subsurface indication found during PAUT had a maximum measured width of 1.47 and depth of 0.19. The allowable flaw depth was found to be significantly larger at 0.54. Since the final flaw depth after FCG is only 0.2094 and remains classified as subsurface, the flaw is acceptable through the 10 year evaluation period.
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SJ Structural Integrity Associates, lnc.,s;
6.0 REFERENCES
- 1. ASME Boiler and Pressure Vessel and Boiler Code,Section XI, 2007 Edition with 2008 Addenda.
- 2. Email from Brian T Shalayda (Duke) to ONS OSM and Backups, CC Dick Mattson (SI), dated 4/30/2016 7:34am, SI File No. 1600492.204.
- 3. SI Phased Array Ultrasonic Examination Record, Examination Data Sheet No. 3-RC-212-53V, dated May 1, 2016, SI File No. 1600492.202.
- 4. Duke Energy Specification No. OSS-0018.0P-00-008, Rev. 1, Procurement Specification for The Repair of Simple Configuration Nozzles Containing Alloy 600/82/182 materials, SI File No.
1600103.205.
- 5. ASME Boiler and Pressure Vessel and Boiler Code,Section XI, 2013 Edition.
- 6. SI Calculation No. ONS-14Q-301, Rev. 0, HPI Nozzle Geometry and Material Properties.
- 7. Email from David Peltola (Duke) to Dick Mattson (SI), dated 5/3/2016 10:54am, with attachment 23-5015791-00.PDF, SI File No. 1600492.203.
- 8. Duke Power Calculation No. OSC-1522, Attachment 1, Reactor Coolant Loop Piping Stress Report, SI File No. ONS-14Q-202.
- 9. Email Attachment from David Peltola (Duke) to Dick Mattson (SI), dated 4/30/2016 12:11pm, Attachment 2 to OSR-0018-0P-00-0010 Rev 0.docx, SI File No. 1600492.201.
- 10. ASME Boiler and Pressure Vessel and Boiler Code,Section II, Part D, 2001 Edition with 2003 Addenda.
- 11. SI Calculation No. ONS-14Q-305, Rev. 0, Thermal Transient Stress Analyses of the HPI Nozzle.
- 12. ANSYS, Release 8.1 (w/Service Pack 1), ANSYS, Inc.
- 13. ANSYS Mechanical APDL, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc.,
September 2012.
- 14. SI Calculation No. ONS-14Q-307, Rev. 3, ASME Code Stress Range and Fatigue Usage Analysis.
- 15. pc-CRACK, Version 4.1 CS (Project 0900086), Structural Integrity Associates, Inc., December 31, 2013.
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SJ Structural Integrity Associates, lnc.,s; Table 1: Loads at Flaw Location [9]
Msrss, in-2" HPI Forces, lbs Moments, ft-lbs lbs FX FY FZ MX MY MZ Thermal 401.57 446.4 231.55 375.96 923.43 2571.95 33,101 DW -12.9 -287.5 -15.3 -151.4 -74.6 296.0 OBE 97.1 73.2 79.3 72.4 182.0 196.2 SSE 194.1 146.3 158.6 144.9 364.0 392.4 Primary, N/U 109.9 360.6 94.6 223.8 256.6 492.2 7,182 Primary, E/F 206.99 433.76 173.91 296.22 438.59 688.40 10,420 Table 2: Stress Determination at Indications Location Max Resultant Pressure Primary Bending Service Pressure Moment Stress ( (Jm) Stress ( (Jb)
Level (psig) (in-lbs.) (psi) (psi)
A/B 3,192 7,182 3,724 1,910 C/D 3,192 10,420 3,724 2,772 File No.: 1600492.301 Page 10 of 12 Revision: 0 F0306-01R2
SJ Structural Integrity Associates, lnc.,s; Table 3: Stress Ratios and Allowable Flaw Depth-to-Thickness Ratios for a Non-Flux Weld Using Limit Load Analysis Allowable Flaw Service Stress Ratio Allowable 2a/t Depth (2a)
Level (Limit Load) For Full Circ Flaw (inches)
A 0.103 0.74 0.555 B 0.103 0.75 0.5625 C 0.118 0.73 0.54 D 0.118 0.75 0.5625 Table 4: Material Properties for Alloy 600 (N06600) [10]
Youngs Mean Thermal Specific Temperature, Thermal Conductivity, Modulus, Expansion, Heat,
°F x106 psi x10-6 in/in/°F x10-4 Btu/sec-in-°F Btu/lb-°F 70 31.0 6.80 1.99 0.114 200 30.2 7.10 2.11 0.119 300 29.8 7.30 2.22 0.123 400 29.5 7.50 2.34 0.125 500 29.0 7.60 2.45 0.127 600 28.7 7.80 2.57 0.130 700 28.2 7.90 2.69 0.133 Density, p = 0.283 lb/in3, assumed temperature independent.
Poissons Ratio, u = 0.3, assumed temperature independent.
Note: Specific Heat values are derived from the equation shown in General Note (a) of Table TCD [10], Specific Heat = TC / (TD x density).
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SJ Structural Integrity Associates, lnc.,s; Table 5: Fatigue Crack Growth Transients and Cycles 10 Year 60 Year Design Transient Evaluation Cycles [14]
Period Cycles 1A Heatup 360 60 1B Cooldown 360 60 8B Rx Trip High Temp 610 102 8C Rx Trip High Press 170 30 8HPI Activation 70 12 9 Depressurization 40 7 12 HydroTest 5 1 22A HPI Test 40 7 OBE 5 1 File No.: 1600492.301 Page 12 of 12 Revision: 0 F0306-01R2
APPENDIXA PHASED ARRAY ULTRASONIC EXAMINATION RESULTS [3, PDF PG. 4]
File No.: 1600492.301 Page A-1 of A-2 Revision: 0 F0306-01R2
tJ Structural Integrity Associates, Inc.
PHASED ARRAY ULTRASONIC EXAM/NATION RECORD Examination Data Sheet No.: 3-RC-212-53V
/WB-3500 Flaw Evaluation Planar Flaw Comparison to Acceptance Standards - ASME Section XI - 2007 Edition with Addenda Through 2008 Oconee Unit 3 Component ID: HPI 3-RC-212-53V Subsurface Subsurface (S) Depth Depth Flaw Flaw Subsurface Code Flaw No.1 Flaw Flaw Flaw OD to Top OD to Bottom Through Component Flaw Flaw Flaw Flaw Material Flaw Only Flaw Allowable Data View Start End Length of Flaw of Flaw !YM Wall Thickness !!. !!.!! !!r<i ilQ! !!.!! Y=sla ~ Table Results 1 10.30 11 .68 1.38 0.51 0.70 0,19 0,77 0.10 0.07 12.34% Subsurface Austen itic 0.07 1 12.34% 11. 24% Exceeds 2 1089 12.27 1.38 0,37 0.54 0.17 0,77 0.09 0.06 11.04% Subsurface A.Jstenitic 0.06 1 11.04% 11. 19% Acceptable 3 10.29 11 .37 1.08 0.39 0.56 0.17 0.77 0.09 0.08 11.04% Subsurface Austenitic 0.08 1 11.04% 11. 29% Acceptable 4 9.99 11.46 1.47 0.53 0.71 0.18 0.77 0.09 0.06 11.69% Subsurface Austenilic 0.06 11.69% 11.19% Exceeds
~ 0.29 0.87 0.58 0.59 0.69 0.10 0.77 0.05 0.09 6.49% Subsurface Austenitic 0.09 6.49% 11. 35% Acceplable FlawNo.3 1.27 1.76 0.49 0.57 0,70 0.13 0.77 0.07 0.13 8.44% Subsurface Austen,tic 0,13 8.44% 11.48% Acceptable Flaw #1 Data Views:
Data View #1
- Flaw #1 As depic1ed with 45' RL Data from the nozzle side of the weld Data View #2 Flaw #1 As depicted with 60" RL Data from the nozzle side of the weld Data View #3 Flaw #1 As depicted with 45' RL Data from the safe end side of the weld Da1a View #4* Flaw#1 As depicted with 45' shear wave data from the safe end side of the weld File No.: 1600492.301 Page A-2 of A-2 Revision: 0 F0306-01R2