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Category:Letter type:LIC
MONTHYEARLIC-23-0007, Response to Fort Calhoun, Unit 1 & Independent Spent Fuel Storage Installation Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements Request for Additional Information2023-12-0606 December 2023 Response to Fort Calhoun, Unit 1 & Independent Spent Fuel Storage Installation Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements Request for Additional Information LIC-23-0005, Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - 2nd Request for Additional Information (EPID L-2021-LIT-0000) June 2, 20232023-08-24024 August 2023 Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - 2nd Request for Additional Information (EPID L-2021-LIT-0000) June 2, 2023 LIC-23-0004, (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report2023-04-20020 April 2023 (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report LIC-23-0003, Annual Decommissioning Funding / Irradiated Fuel Management Status Report2023-03-15015 March 2023 Annual Decommissioning Funding / Irradiated Fuel Management Status Report LIC-23-0001, Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - Request for Additional Information2023-02-27027 February 2023 Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - Request for Additional Information LIC-23-0002, Independent Spent Fuel Storage Installation, Annual Radioactive Effluent Release Report2023-02-20020 February 2023 Independent Spent Fuel Storage Installation, Annual Radioactive Effluent Release Report LIC-22-0010, Response to Fort Calhoun Station, Unit No. 1 - Review of License Termination Plan Requirements - Request for Additional Information2022-06-15015 June 2022 Response to Fort Calhoun Station, Unit No. 1 - Review of License Termination Plan Requirements - Request for Additional Information LIC-22-0005, (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report2022-04-20020 April 2022 (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report LIC-22-0009, Annual Decommissioning Funding / Irradiated Fuel Management Status Report2022-03-30030 March 2022 Annual Decommissioning Funding / Irradiated Fuel Management Status Report LIC-22-0006, Reactor Head Disposition Project Overview2022-03-17017 March 2022 Reactor Head Disposition Project Overview LIC-22-0004, Independent Spent Fuel Storage Installation, Annual Radioactive Effluent Release Report2022-02-17017 February 2022 Independent Spent Fuel Storage Installation, Annual Radioactive Effluent Release Report LIC-21-0008, Organizational and Management Change2021-10-28028 October 2021 Organizational and Management Change LIC-21-0007, ISFSI Only Emergency Plan Update2021-09-0808 September 2021 ISFSI Only Emergency Plan Update LIC-21-0004, Radiological Effluent Release Report and Radiological Environmental Operating Report2021-04-29029 April 2021 Radiological Effluent Release Report and Radiological Environmental Operating Report LIC-21-0003, Independent Spent Fuel Storage Installation - 2021 Annual Decommissioning Funding / Irradiated Fuel Management Status Report2021-03-30030 March 2021 Independent Spent Fuel Storage Installation - 2021 Annual Decommissioning Funding / Irradiated Fuel Management Status Report LIC-21-0002, Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report2021-02-22022 February 2021 Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report LIC-20-0015, Correction to Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report 2019 (ML20121A092)2020-07-29029 July 2020 Correction to Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report 2019 (ML20121A092) LIC-20-0014, Submittal of Revision 8 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP)2020-07-15015 July 2020 Submittal of Revision 8 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP) LIC-20-0012, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration and Certification of Permanent Removal of All Spent Fuel Assemblies from the Spent Fuel Pool2020-05-18018 May 2020 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration and Certification of Permanent Removal of All Spent Fuel Assemblies from the Spent Fuel Pool LIC-20-0011, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2020-05-0707 May 2020 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-20-0009, (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report2020-04-30030 April 2020 (FCS) Radiological Effluent Release Report and Radiological Environmental Operating Report LIC-20-0008, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2020-04-13013 April 2020 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-20-0006, (Fcs), Unit 1, Request for Exemption from 10 CFR 20, Appendix G, Section Iii.E2020-03-26026 March 2020 (Fcs), Unit 1, Request for Exemption from 10 CFR 20, Appendix G, Section Iii.E LIC-20-0004, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2020-03-10010 March 2020 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-20-0003, Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report2020-02-27027 February 2020 Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report LIC-20-0002, Independent Spent Fuel Storage Installation - Submittal of Revision 7 to Physical Security Plan2020-02-27027 February 2020 Independent Spent Fuel Storage Installation - Submittal of Revision 7 to Physical Security Plan LIC-20-0001, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2020-02-0606 February 2020 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-19-0025, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2019-12-19019 December 2019 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-19-0007, Post-Shutdown Decommissioning Activities Report2019-12-16016 December 2019 Post-Shutdown Decommissioning Activities Report LIC-19-0021, Independent Spent Fuel Storage Installation - Response to Request for Additional Information License Amendment for ISFSI-only EP and EAL Scheme2019-11-20020 November 2019 Independent Spent Fuel Storage Installation - Response to Request for Additional Information License Amendment for ISFSI-only EP and EAL Scheme LIC-19-0022, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2019-11-18018 November 2019 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration LIC-19-0018, Submittal of Revision 6 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP)2019-11-14014 November 2019 Submittal of Revision 6 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP) LIC-19-0020, Independent Spent Fuel Storage Installation - Updated Information Submittal and Revision 1 of the Blast Analysis2019-10-17017 October 2019 Independent Spent Fuel Storage Installation - Updated Information Submittal and Revision 1 of the Blast Analysis LIC-19-0017, Clarification to Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associated with Access Authorization for Fort Calhoun Station Dated 9 November, 20042019-08-0808 August 2019 Clarification to Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associated with Access Authorization for Fort Calhoun Station Dated 9 November, 2004 An LIC-19-0010, License Amendment Request (LAR) 19-03; Revised Fort Calhoun Station License in Support of the Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associa2019-05-20020 May 2019 License Amendment Request (LAR) 19-03; Revised Fort Calhoun Station License in Support of the Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associated LIC-19-0008, Submittal of 2018 Annual Report2019-04-0404 April 2019 Submittal of 2018 Annual Report LIC-19-0004, Independent Spent Fuel Storage Installation - Submittal of Revision 4 to Physical Security Plan (PSP)2019-03-28028 March 2019 Independent Spent Fuel Storage Installation - Submittal of Revision 4 to Physical Security Plan (PSP) LIC-19-0005, 2019 Annual Decommissioning Funding I Irradiated Fuel Management Status Report2019-03-28028 March 2019 2019 Annual Decommissioning Funding I Irradiated Fuel Management Status Report LIC-19-0001, License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme2019-02-28028 February 2019 License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme LIC-19-0003, Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associated with Access Authorization for Fort Calhoun Station Dated 9 November, 2004 and Relaxation R2019-02-28028 February 2019 Revised Response to Orders for Interim Safeguards and Security Compensatory Measures and Implementation of Additional Security Measures Associated with Access Authorization for Fort Calhoun Station Dated 9 November, 2004 and Relaxation Requ LIC-19-0002, Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report2019-01-0303 January 2019 Independent Spent Fuel Storage Installation Annual Radioactive Effluent Release Report LIC-18-0031, Independent Spent Fuel Storage Installation - Transmittal of Emergency Plan Update2018-12-17017 December 2018 Independent Spent Fuel Storage Installation - Transmittal of Emergency Plan Update LIC-18-0030, Request for Partial Site Release Phase 22018-11-12012 November 2018 Request for Partial Site Release Phase 2 LIC-18-0028, Letter of Intent to Submit Request for Additional Partial Site Release2018-10-18018 October 2018 Letter of Intent to Submit Request for Additional Partial Site Release LIC-18-0027, Submittal of Foreign Ownership, Control or Influence Five-Year Renewal Filing and NRC Facility Clearance Update2018-10-0303 October 2018 Submittal of Foreign Ownership, Control or Influence Five-Year Renewal Filing and NRC Facility Clearance Update LIC-18-0003, License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool2018-09-28028 September 2018 License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool LIC-18-0025, Guarantee of Payment of Deferred Premiums for the Period of July 1, 2018 to June 30, 20192018-07-19019 July 2018 Guarantee of Payment of Deferred Premiums for the Period of July 1, 2018 to June 30, 2019 LIC-18-0023, Fort Calhoun Station, Unit 1 Request for Partial Site Release2018-06-29029 June 2018 Fort Calhoun Station, Unit 1 Request for Partial Site Release LIC-18-0021, Transmittal of Revision 3 to the Physical Security Plan (PSP) and Revision 4 to the FCS Protective Strategy (Safeguards Contingency Plan)2018-06-0606 June 2018 Transmittal of Revision 3 to the Physical Security Plan (PSP) and Revision 4 to the FCS Protective Strategy (Safeguards Contingency Plan) LIC-18-0017, Submittal of Revision 2 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP) and Revision 3 of the FCS Protective Strategy2018-04-26026 April 2018 Submittal of Revision 2 to the Fort Calhoun Station (Fcs), Physical Security Plan (PSP) and Revision 3 of the FCS Protective Strategy 2023-08-24
[Table view] Category:Report
MONTHYEARLIC-22-0006, Reactor Head Disposition Project Overview2022-03-17017 March 2022 Reactor Head Disposition Project Overview LIC-22-0001, Well Water Results with River Stage2022-01-13013 January 2022 Well Water Results with River Stage ML22034A5942021-12-22022 December 2021 Hydrogeological Conceptual Site Model, Rev. 5 ML21271A5242021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CM-244 Soil Regression and Correlation Report ML21271A2102021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 4 3 Omaha Public Power District, Fort Calhoun Station Unit 1, Defueled Safety Analysis Report CAC 2 ML21271A1442021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS LTP Rev 0. Final - All ML21271A3052021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Regression and Correlation Report TC-99 ML21271A2112021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 5 18 Radiation Safety and Control Services, Tsd 21-043, Radionuclides of Concern in Support of the Fort Calhoun License Termination Plan ML21271A4082021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Buried Pipe Insitu Uncertainty Report Eu-152 ML21271A3852021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Buried Pipe Excavation Dsr 0.15 M ML21271A5212021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CM-244 Soil Uncertainty Report ML21271A5222021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CM-243 Soil Uncertainty Report ML21271A1802021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 2 19, Omaha Public Power District, FC-20-012, Fort Calhoun Station Decommissioning Project Radiological Characterization Report ML21271A2722021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Drilling Spoils Dsr ML21271A5172021-08-0303 August 2021 License Amendment Request (LAR) 21-01, EU-154 Soil Uncertainty Report ML21271A3262021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Regression and Correlation Report NI-63 ML21271A6002021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 8, 13, Omaha Public Power District, Offsite Dose Calculation Manual CAC2 ML21271A4502021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Soil AF 100 Dcgl 0.15 M.Rad ML21271A4782021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Eu-152 Soil Uncertainty Report ML21271A4802021-08-0303 August 2021 License Amendment Request (LAR) 21-01, C-14 Soil Uncertainty Report ML21271A2582021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 8 Radiation Safety and Control Services, Tsd 21-043, Radionuclides of Concern in Support of the Fort Calhoun -1 ML21271A1962021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 3, 1, Omaha Public Power District, Defueled Safety Analysis Report CAC 2 ML21271A4772021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CO-60 Soil Regression and Correlation Report ML21271A3212021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Uncertainty Report SR-90 ML21271A4852021-08-0303 August 2021 License Amendment Request (LAR) 21-01, NI-63 Soil Regression and Correlation Report ML21271A2902021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Uncertainty Report H-3 ML21271A4832021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Eu-155 Soil Uncertainty Report ML21271A4952021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CS-134 Soil Regression and Correlation Report ML21271A3012021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Uncertainty Report C-14 ML21271A5092021-08-0303 August 2021 License Amendment Request (LAR) 21-01, EU-152 Soil Regression and Correlation Report ML21271A2612021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 1 Haley & Aldrich, Hydrogeological Conceptual Site Model, Rev. 2, Fort Calhoun Station, Blair, Ne, 2021 CAC2 ML21271A2652021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 7 FC-20-007, Fort Calhoun Station Potential Radionuclides of Concern CAC2 ML21271A5052021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CO-58 Soil Uncertainty Report ML21271A4872021-08-0303 August 2021 License Amendment Request (LAR) 21-01, EU-154 Soil Regression and Correlation Report ML21271A4882021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CS-137 Soil Uncertainty Report ML21271A2862021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Regression and Correlation Report H-3 ML21271A2192021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 5 12 FC-20-007, Fort Calhoun Station Potential Radionuclides of Concern CAC2 ML21271A2622021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 2 Radiation Safety and Control Services, Tsd 20-001, Historical Site Assessment for Fort Calhoun Station - F-1 ML21271A2752021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Bfm Insitu Scenario Dsr ML21271A4972021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CO-58 Soil Regression and Correlation Report ML21271A5042021-08-0303 August 2021 License Amendment Request (LAR) 21-01, CM-243 Soil Regression and Correlation Report ML21271A2672021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 3 Omaha Public Power District, Fort Calhoun Station Unit 1, Defueled Safety Analysis Report CAC2 ML21271A2692021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 6 19 E. A. Napier, Hanford Environmental Dosimetry Upgrade Project, Gen II the Hanford Environmental Radiation D-1 ML21271A3862021-08-0303 August 2021 License Amendment Request (LAR) 21-01, FCS Buried Pipe Excavation Dsr 1.0 M ML21271A2002021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 3 14 Backfill Attachment Final CAC 2 ML21271A4572021-07-28028 July 2021 License Amendment Request (LAR) 21-01, FCS Soil AF 143 Dcgl 0.15 M ML21271A4562021-07-28028 July 2021 License Amendment Request (LAR) 21-01, FCS Soil AF 143 Dcgl 1 M ML21271A1982021-07-24024 July 2021 License Amendment Request (LAR) 21-01, Chapter 3, 6, Omaha Public Power District, FC-21-002, Description of Embedded Piping, Penetrations, and Buried Pipe to Remain in Fort Calhoun End State ML21271A2602021-07-24024 July 2021 License Amendment Request (LAR) 21-01, Chapter 6 4, Omaha Public Power District, FC-21-002, Description of Embedded Piping, Penetrations, and Buried Pipe to Remain in Fort Calhoun End State ML21271A2132021-07-23023 July 2021 License Amendment Request (LAR) 21-01, Chapter 5 9 Oddp FC-20-006 - End State Concrete Surface Areas and Volumes Rev 0 CAC2 2022-03-17
[Table view] Category:Miscellaneous
MONTHYEARML21271A2002021-08-0303 August 2021 License Amendment Request (LAR) 21-01, Chapter 3 14 Backfill Attachment Final CAC 2 ML21271A1742017-03-0202 March 2017 License Amendment Request (LAR) 21-01, Chapter 2 20 Storm Water Discharge Authorization Permit Renewal NER910677 Package for FCS cac2 ML16182A3612016-08-0404 August 2016 Staff Review of Spend Fuel Pool Evaluation Associated with Reevaluation Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 LIC-16-0031, Ft. Calhoun, Unit 1 - Annual Report for 2015 Loss-of-Coolant Accident (Loca)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.462016-04-28028 April 2016 Ft. Calhoun, Unit 1 - Annual Report for 2015 Loss-of-Coolant Accident (Loca)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46 LIC-16-0013, Ft. Calhoun, Unit 1 and ISFSI - Transmittal of 10CFR72.48 Evaluation Summary Report for January 1, 2015 Through December 31, 20152016-02-10010 February 2016 Ft. Calhoun, Unit 1 and ISFSI - Transmittal of 10CFR72.48 Evaluation Summary Report for January 1, 2015 Through December 31, 2015 LIC-15-0005, (Fcs), Unit No. 1, 10 CFR 72.48 Evaluation Summary Report for January 1, 2013 Through December 31, 20142015-01-15015 January 2015 (Fcs), Unit No. 1, 10 CFR 72.48 Evaluation Summary Report for January 1, 2013 Through December 31, 2014 ML14226A5512014-08-14014 August 2014 Revision of Standard Practice Procedures Plan for Fort Calhoun Station, Unit 1 ML14157A0792014-06-24024 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Fukushima Dai-Ichi Nuclear Power Plant Accident ML14105A3732014-04-22022 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to Fukushima Dai-Ichi Nuclear Power Plant Accident LIC-14-0047, Omaha Public Power District - Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushi2014-03-31031 March 2014 Omaha Public Power District - Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima ML13172A3432013-06-19019 June 2013 Integrated Performance Improvement Plan, Revision 5, Page 138 Through End LIC-13-0086, Integrated Performance Improvement Plan, Revision 5, Cover Through Page 732013-06-19019 June 2013 Integrated Performance Improvement Plan, Revision 5, Cover Through Page 73 ML13172A3422013-06-19019 June 2013 Integrated Performance Improvement Plan, Revision 5, Page 74 Through Page 137 ML12340A2542012-11-27027 November 2012 OPPD 180-day Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. Part 2 of 3 ML12340A2552012-11-27027 November 2012 OPPD 180-day Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. Part 2 of 3 LIC-12-0098, Integrated Performance Improvement Plan, Rev. 32012-07-0909 July 2012 Integrated Performance Improvement Plan, Rev. 3 ML11276A1202011-09-28028 September 2011 Enclosure 1 to LIC-11-0099, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) LIC-11-0099, Table B-2 Nuclear Safety Capability Assessment Methodology Review2011-09-28028 September 2011 Table B-2 Nuclear Safety Capability Assessment Methodology Review LIC-11-0090, Post-Flooding Recovery Action Plan2011-08-10010 August 2011 Post-Flooding Recovery Action Plan ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants LIC-08-0120, Steam Generator Eddy Current Test Report - 2008 Refueling Outage2008-12-0909 December 2008 Steam Generator Eddy Current Test Report - 2008 Refueling Outage LIC-08-0107, Notice of Completion of Corrective Actions Taken in Response to Generic Letter 2004-02 and Response to Request for Additional Information2008-10-16016 October 2008 Notice of Completion of Corrective Actions Taken in Response to Generic Letter 2004-02 and Response to Request for Additional Information LIC-08-0071, Transmittal of Fort Calhoun, Unit 1, Reactor Coolant System Pressure and Temperature Limits Report, Revision 42008-05-27027 May 2008 Transmittal of Fort Calhoun, Unit 1, Reactor Coolant System Pressure and Temperature Limits Report, Revision 4 LIC-07-0017, Fitness-for-Duty Program Performance Data Report for July 1 Through December 31, 20062007-02-22022 February 2007 Fitness-for-Duty Program Performance Data Report for July 1 Through December 31, 2006 LIC-06-0138, Transmittal of Revision 3 of Fort Calhoun Station, Unit No. 1 (Fcs), Pressure Temperature Limits Report (PTLR)2006-11-29029 November 2006 Transmittal of Revision 3 of Fort Calhoun Station, Unit No. 1 (Fcs), Pressure Temperature Limits Report (PTLR) LIC-06-0083, Fitness-for-Duty Program Performance Report2006-07-26026 July 2006 Fitness-for-Duty Program Performance Report LIC-06-0056, Request for Exemption from Nuhoms Certificate of Compliance No. 1004, Amendment No. 82006-06-0909 June 2006 Request for Exemption from Nuhoms Certificate of Compliance No. 1004, Amendment No. 8 ML0606501772006-03-0606 March 2006 Licensed Operator Positive Drug Tests ML0535403192005-12-0505 December 2005 51-9004805-001, FCS RSG - Afas Verification - RAI Responses. LIC-05-0043, Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors2005-04-15015 April 2005 Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors LIC-05-0016, Omaha Public Power District, Fitness-for-Duty Program Performance Data Report2005-02-0404 February 2005 Omaha Public Power District, Fitness-for-Duty Program Performance Data Report ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement LIC-06-0004, Attachment 1, Engineering Analysis (EA-FC-04-010) Recommendations for Implementing of Compensatory Actions in Response to NRC Bulletin 2003-012004-11-23023 November 2004 Attachment 1, Engineering Analysis (EA-FC-04-010) Recommendations for Implementing of Compensatory Actions in Response to NRC Bulletin 2003-01 LIC-04-0081, Fitness-for-Duty Program Performance Data Report for Six Month Period from January 1 Through June 30, 20042004-07-14014 July 2004 Fitness-for-Duty Program Performance Data Report for Six Month Period from January 1 Through June 30, 2004 ML0532204612004-06-14014 June 2004 Power Reactor Status Report for 6/14/04 ML0626304082004-03-0101 March 2004 PSEG Nuclear, LLC Salem/Hope Creek Safety Culture Assessment LIC-03-0154, Relief Request Pertaining to Reactor Vessel Nozzle Inspection for Third 10-Year Interval Revision2003-11-21021 November 2003 Relief Request Pertaining to Reactor Vessel Nozzle Inspection for Third 10-Year Interval Revision LIC-03-0124, Transmittal of Fort Calhoun Station (FCS) Pressure Temperature Limits Report (Ptlr), Revision 02003-09-10010 September 2003 Transmittal of Fort Calhoun Station (FCS) Pressure Temperature Limits Report (Ptlr), Revision 0 LIC-03-0119, Fitness-for-Duty Program Performance Data Report2003-08-27027 August 2003 Fitness-for-Duty Program Performance Data Report ML0314102122003-05-20020 May 2003 Thesis (Miscellaneous Report - 134 Pages), Fort Calhoun, S107100 LIC-03-0063, Fire Modeling Analysis - Fire Area 322003-04-25025 April 2003 Fire Modeling Analysis - Fire Area 32 LIC-03-0038, Comments on Letter from Us Fish and Wildlife Service Re Impact of Plant Operation on Pallid Sturgeon2003-03-14014 March 2003 Comments on Letter from Us Fish and Wildlife Service Re Impact of Plant Operation on Pallid Sturgeon ML0232904702002-11-22022 November 2002 Issuance of Environmental Scoping Summary Report Associated with Staff'S Review of Application for Renewal of Operating License for Fort Calhoun Station, Unit 1 LIC-02-0104, Omaha Public Power District Fitness-for-Duty Program Performance Data Report2002-08-29029 August 2002 Omaha Public Power District Fitness-for-Duty Program Performance Data Report LIC-02-0054, Annual Report for 2001 Loss of Coolant Accident (Loca)/Emergency Core Cooling System (ECCS) Models Pursuant to 10CFR50.462002-04-30030 April 2002 Annual Report for 2001 Loss of Coolant Accident (Loca)/Emergency Core Cooling System (ECCS) Models Pursuant to 10CFR50.46 LIC-02-0046, Special Report on Inoperability of Main Steam Line Radiation Monitor RM-064 for Post-Accident Monitoring2002-04-16016 April 2002 Special Report on Inoperability of Main Steam Line Radiation Monitor RM-064 for Post-Accident Monitoring LIC-02-0023, Fitness-for-Duty Program Performance Data Report from Omaha Public Power District, July 1 Through December 31, 20012002-02-26026 February 2002 Fitness-for-Duty Program Performance Data Report from Omaha Public Power District, July 1 Through December 31, 2001 2021-08-03
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Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 April 15,2005 LIC-05-0043 U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, D.C. 20555-0001
References:
- 1. Docket No. 50-285
- 2. Letter from Ralph L. Phelps (OPPD) to Document Control Desk (NRC) dated April 7, 2005, Fort Calhoun Station Unit No. 1, Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (LIC-05-0040)
- 3. Letter from R. William Borchardt (NRC) to Ross Ridenoure (OPPD) dated February 20, 2004, Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (NRC-04-0022)
(ML040220181)
SUBJECT:
Fort Calhoun Station Unit No. 1, Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors In Reference 2, the Omaha Public Power District (OPPD) provided information in support of a relaxation request with respect to Reference 3. NRC personnel in an April 4, 2005 phone conversation asked that the information in Reference 2 be provided in advance of forthcoming analysis information to facilitate their review. In Reference 2, OPPD specified plans to submit additional analysis information from Westinghouse and Dominion Engineering, and a status report of nozzle examinations completed to date, as a supplement to support review and approval of Reference 2. This letter forwards the required analysis information, consisting of the hoop stress distribution analysis for the Incore Instrumentation (ICI) and the Control Element Drive Mechanism (CEDM) penetration nozzles, as well as the fracture mechanics crack growth analysis for the CEDM penetration nozzles. A status report of the nozzle penetration inspections will be submitted on or about May 1,2005.
OPPD requests that the NRC complete its review and approval of the subject relaxation request by May 6,2005.
Employment with Equal Opportunity 4171
U. S. Nuclear Regulatory Commission LIC-05-0043 Page 2 The following commitment is made in this submittal:
- 1. If the NRC staff finds that the crack-growth formula in industry report MRP-55 is unacceptable, OPPD shall revise its analysis that justifies relaxation of Reference 3 within 30 days after the NRC informs OPPD of an NRC-approved crack growth formula. If OPPDs revised analysis shows that the crack growth acceptance criteria are exceeded prior to the end of the current operating cycle, this relaxation is rescinded and OPPD shall, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, submit to the NRC written justification for continued operation. If the revised analysis shows that the crack growth acceptance criteria are exceeded during the subsequent operating cycle, OPPD shall, within 30 days, submit the revised analysis for NRC review. If the revised analysis shows that the crack growth acceptance criteria are not exceeded during either the current operating cycle or the subsequent operating cycle, OPPD shall, within 30 days, submit a letter to the NRC confirming that its analysis has been revised. Any future crack-growth analyses performed for this and future cycles for RPV head penetrations must be based on an acceptable crack growth rate formula.
If you have any questions or require additional information, please contact Thomas R. Byrne at (402) 533-7368.
Ralph L. Phelps Division Manager Nuclear Engineering RLP/TRB/trb Attachment 1 - Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors
LIC-05-0043 Attachment I Page 1 Attachment 1 Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors
LIC-05-0043 Page 2 Attachment 1 Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors Executive Summary Section 1.O - Introduction - This section provides an overview of the process used to assess the hoop stresses for the crack growth calculations for the areas in which relaxation is requested.
Section 2.0 - Hoop Stress Distribution Above the Root of the CEDM/ICI Penetration Nozzle J-Groove Weld - A discussion of the hoop stresses found above all of the J-groove welds is provided and graphic results are provided in Figures 1 - 9. Specific information on the proposed relaxation areas (Control Element Drive Mechanism (CEDM) penetration nozzles 22 - 41) is provided and graphic results for these areas are provided in Figures 4 - 7.
Section 3.0 - Crack Growth Calculation for CEDM Penetration Nozzle Area Above the Root of the J-Groove Weld - Crack growth calculations for the proposed relaxation areas in CEDM penetration nozzles 22 - 41 (CEDM nozzle angles 37.3" and 41.7") are presented with the basis of why these crack growths are conservative. Figure 10 graphically summarizes these calculations.
Section 4.0 Hoop Stress Distribution Below the Toe of the CEDM/ICI Penetration Nozzle J-Groove Weld - This section completes the hoop stress assessments by presenting the results in areas below the toe of all of the J-groove welds, as shown in Figures 11 - 19.
Section 5.0 Potential Lack of 100% Eddy Current Coverage of the ICI Penetration Nozzle J-Groove Welds - This section provides the justification for the maintenance of structural integrity of the In-Core Instrumentation (ICI) penetration nozzles with a 10" unexamined area of the ICI J-groove welds.
Section 6.0 References
LIC-05-0043 Page 3 1.O Introduction The objective f this analysis was to obtain accurate stresses in th CEDM and ICI penetration nozzles and their immediate vicinity for FCS. To do so requires a three-dimensional finite element analysis (Reference 1) that considers all the pertinent loadings on the penetrations. Four CEDM locations with nozzle angles of O", 24.6", 37.3", 41.7" and one ICI nozzle with nozzle angle of 54.4" were analyzed. The analyses were used to provide information for the flaw tolerance evaluation and/or determine the adequacy of the inspection coverage. The methodology used in the finite element analysis (Reference 1) is consistent with that performed to support relaxation requests from Reference 7 for other nuclear power plants.
A three-dimensional finite element model comprised of isoparametric brick and wedge elements was used to obtain the stresses and deflections. Taking advantage of the symmetry of the reactor pressure vessel (RPV) head, only half of the CEDM/ICI penetration nozzles were modeled. In the model, the lower portion of the CEDM/ICI penetration nozzle, the adjacent section of the RPV head, and the joining weld were modeled. The vessel to penetration nozzle weld was simulated with two weld passes. The penetration nozzle, weld metal, cladding and the W V head shell were modeled in accordance with the relevant material properties.
The most important loading conditions which exist on the penetration include internal pressure and thermal expansion effects typical of steady state operation.
The RPV head temperature that OPPD used in the analysis is 588°F (Reference 8). In addition, residual stresses due to the welding of the penetrations to the RPV head were considered.
The hoop stress in the penetration nozzle resulting from the steady state operation loadings and welding residual stresses is much higher than the axial stress.
Therefore, only the hoop stress is assessed for crack growth calculations as part of this relaxation request. This is consistent with the field findings, where the cracks discovered are generally oriented axially. Typically, in-service cracks will orient themselves perpendicular to the largest stress component. Also it should be noted that the highest tensile hoop stress is at the uphill side and downhill side locations rather than midway around the penetration, where it is less limiting. This is consistent with finding axial cracks only in the vicinity of the uphill side and downhill side locations. It is these steady state hoop stresses that are used in this analysis to predict crack propagation in the penetration nozzles.
2.0 Hoop Stress Distribution Above the Root of the CEDM/ICI Penetration Nozzle J-Groove Weld Figures 1 - 9 show the hoop stress distributions for the regions that are within 2 inches from the top of the root of the J-groove weld on the uphill side for the FCS
LIC-05-0043 Page 4 CEDM and ICI RPV upper head penetrations. The stress distributions shown are for the inside and outside surface of the FWV upper head penetrations. The stress distributions shown in Figures 1 - 9 are typical of those observed in the upper RPV head penetration nozzles for other nuclear power plants. The stresses are highest in the vicinity of the J-groove weld and decrease rapidly as the distance above the root of the J-groove weld increases.
For the CEDM penetration nozzles numbers 22 - 41 (Figures 4 - 7) inspection coverage is less than 100%. The area where inspection coverage is not obtained is from 1 1/4 inches to 2 inches above the root of the J-groove on the uphill side above the thermal sleeve centering tabs. This area has a maximum hoop stress of 15 ksi. There is nearly universal agreement that high stresses, on the order of the material yield strength, are necessary to initiate Primary Water Stress Corrosion Cracking (PWSCC). There is no known case of stress corrosion cracking of Alloy 600 below the yield stress (Reference 2). Typical yield strengths for wrought Alloy 600 RPV head penetration nozzles are in the range of 37 ksi to 65 ksi. Weld metal yield strengths are generally higher. The yield strength of the CEDM head penetration nozzles for FCS varies from 37 ksi (nozzles 1 - 10 and 15 - 41) to 56 ksi (nozzles 11 - 14) (Reference 3). However, it has been determined that a stress level of 20 ksi is the point below which PWSCC initiation is extremely unlikely (Reference 2). Since the maximum hoop stress is only 15 ksi in the region where inspection coverage is less than loo%, PWSCC initiation in the region not being inspected is extremely unlikely. The area that cannot be inspected for the CEDM Inside Diameter (ID) scans is approximately 1.2% of the total area being inspected.
As shown in Figures 1 - 7, the hoop stresses are highest in the vicinity of the J-groove weld. It is more likely for an indication to exist and be detected in the inspection area of the penetration from 1.25 inches above the J-groove weld to 2.0 inches below the J-groove weld, which includes the high stress region in the vicinity of the J-weld. It is not likely that an indication will exist in the low stress region with a maximum hoop stress of only 15 ksi, which is the area not able to be inspected. The stresses in the area proposed to not be inspected are less than or equal to the stresses in the areas inspected.
LIC-05-0043 Page 5 Figure 1 Hoop Stress in 0' CEDM Nozzle vs. Distance from Top of Weld, Uphill and Downhill 70,000 60,000 50,000 40,000
-a f 30,000 e!
3i 20,000 10,000 0
-10,ooc 0.0 0.5 1.0 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)
LIC-05-0043 Attachment I Page 6 Figure 2 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Top of Weld, Uphill 70,000 60,000 50,000 40,000
-u) a U
I 30,000 I
t 20,000 10,000 0
-10,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)
Figure 3 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Top of Weld, Downhill 70,000 60,000 50,000 40,000
.-a
-a-m 30,000 t
I 20,000 10,000 0
-10,000 0.0 1 .o 2.0 3.0 4.0 5.0 Distance Up from Top of The Weld (in.)
LIC-05-0043 Page 7 Figure 4 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Top of Weld, Uphill 80 000 I . . . _ ~ --"_,-".._-__~..._...._.-
......................... ......................... i I
......................... 1 ........................
I I the Rdot of Uphill Weld
-10 000 ... ................................................ '. .................................................. .....................
-20,0008 " " I ' " ' ~ ' ' I ' ' ~ " " I " '
0.0 0.5 1.o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)
Figure 5 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Top of Weld, Downhill 70,000 I r 0.0 1.o 2.0 3.0 4.0 5.0 Distance Up from Top of The Weld (in.)
LIC-05-0043 Page 8 Figure 6 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Top of Weld, Uphill 0.0 0.5 1.0 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)
Figure 7 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Top of Weld, Downhill 0.0 1 .o 2.0 3.0 4.0 5.0 6.0 Distance Up from Top of The Weld (in.)
LIC-05-0043 Page 9 Figure 8 Hoop Stress in 54.4' ICI Nozzle VS. Distance from Top of Weld, Uphill 50,000 40,000 30,000
-a0 20,000 e!
ai 10,000 0
-1 0,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)
Figure 9 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Top of Weld, Downhill 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 Distance Up from Top of The Weld (in.)
LIC-05-0043 Page 10 3 .O Crack Growth Calculation for CEDM Penetration Nozzle Area Above the Root of the J-Groove Weld A PWSCC crack growth calculation has been performed in the region above the root of the J-groove weld that is proposed to not be inspected. The purpose of the calculation is to determine the maximum flaw size for an axial inside surface flaw that would grow to 75% of the wall thickness in a single fuel cycle (18 months or 1.5 Effective Full Power Years (EFPY)). One fuel cycle was chosen since OPPD will be installing a new RPV head during the Fall 2006 Refueling Outage (RFO).
The methodology used in the crack growth calculation is consistent with the NRC flaw evaluation guidelines for the upper RPV head penetrations (Reference 4).
The PWSCC crack growth rate used in the NRC flaw evaluation guidelines is the same as that recommended in Reference 5. Assuming an aspect ratio of 6 (Reference 6), the crack growth results are shown in Figure 10 and summarized in Table 1 for both the downhill and uphill side of the two outermost rows of CEDM nozzle locations at FCS (CEDM nozzle penetrations 22 - 41).
Table 1 Minimum Flaw Size to Reach 75% of Wall Thickness in One Fuel Cycle for CEDM Nozzle Penetrations (Aspect Ratio = 6)
Minimum Flaw Size (% Through-wall)
CEDM Nozzle Angle (") Downhill Uphill 37.3 68.5 69.1 41.7 68.2 68.6 Based on the results given in Table 1, for an inside diameter axial surface flaw, a minimum initial flaw depth of 0.26 inch (68% part-through wall) is required to reach 75% of the wall thickness in one fuel cycle. For an aspect ratio of 6, the minimum initial flaw length is 1.56 inches long. Due to the low probability of PWSCC initiation in the low stress region that is more than 1.25 inches above the root of the J-groove weld on the uphill side, the existence of a 68% part-through wall inside diameter axial surface flaw with an aspect ratio of 6 in that region is extremely unlikely.
In addition, there is inherent conservatism in the above crack growth results.
From Table 5-3 of Reference 5 , the mean crack growth amplitude (a) for the Huntington Alloy 600 heats used in the development of the recommended PWSCC crack growth rate are summarized in Table 2.
LIC-05-0043 Page 11 Table 2 Mean MRP-55 Crack Growth Amplitude (a)for Huntington Material Test Data NX8101 I Huntindon I 1 . 3 7 ~ 0-l2 1
NX8664 I Huntington I 1.29xio-l2 NX64:20G I Huntinnton I 7.21~10"'
NX9240 Huntington 4 . 9 7 1~0-I' NX8 168G HuntinHon 1.93x10-I3 The recommended crack growth amplitude, a, from the NRC flaw evaluation guidelines (Reference 4) is 2.67x1O-l2,which is a factor of 1.9 higher than the highest growth crack amplitude for any of the Huntington material heats in Table
- 2. Since Huntington is the material supplier for the CEDM penetration nozzles for FCS, the PWSCC crack growth rate used in the above crack growth calculation and the resulting crack growth predictions are conservative.
Due to the low probability of PWSCC initiation or the existence of a 68% part-through inside diameter surface axial flaw in the proposed low stress regions not to be inspected, OPPD believes that compliance with the requirements in Reference 7 for the two outermost rows of CEDM penetration nozzles would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per the information supplied in Reference 9.
LIC-05-0043 Page 12 Figure 10 PWSCC Flaw Growth Prediction at 1.25 Inch or More Above the Root of Uphill Side J-weld 1 .o 0.9 f
1
- 0.8 1
g 5
0.7 5
0.E 0.:
0 2 3 4 5 6 7 Time (Effective Full Power Years) 4.0 Hoop Stress Distribution Below the Toe of the CEDM/ICI Penetration Nozzle J-Groove Weld Figures 11 - 19 show the hoop stress distributions for the regions that are below the toe of the J-groove weld for the RPV upper head penetrations. The stress distributions shown are for the inside and outside diameter surface of the RPV upper head penetrations. The stress distributions shown in Figures 11 - 19 are typical of those observed in the upper head penetration nozzles for other nuclear power plants. The stresses are highest in the vicinity of the J-groove weld and decrease rapidly as the distance below the toe of the J-groove weld increases.
Based on Figures 11 - 19, the hoop stress for all the penetration nozzles is less than 20 ksi at a distance of 1 inch or more below the toe of the downhill side J-groove weld.
LIC-05-0043 Page 13 Figure 11 Hoop Stress in 0' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill and Downhill 60,000 50,000 40.000 30,000 10,000 0
-1 0.000
-20,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)
LIC-05-0043 Page 14 Figure 12 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill 0.0 0.5 1 .o 1.5 2.0 2.5 3.0 3.5 4.0 4.5 Distance from Bottom of Weld (in.)
Figure 13 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 60.000 50,000
\
40,000 30,000
-a; 20,000 e
L 10,000 0
-10,000
-20,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)
LIC-05-0043 Page 15 0.0 0.5 1.o 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 Distance from Bottom of Weld (in.)
Figure 15 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 80,000 70,000 60,000 50,000 40,000 111 g
v) 20,000 10,000 0
-10.000
-20.000
-30,000 0.0 0.5 1.o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)
LIC-05-0043 Page 16 Figure 16 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill 0.0 1.o 2.0 3.0 4.0 5.0 6.0 Distance from Bottom of Weld (in.)
Figure 17 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 0.0 0.5 1.o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)
LIC-05-0043 Page 17 Figure 18 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Bottom of Weld, Uphill 60,000 I . . _ . -. -I_
I .
_- , - ~ . .
50,000 40,000 30,000
-.- 20,000
-a u) 10,000 z! 0
-10,000
-20,000
-30,000
-40,000 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Distance from Bottom of Weld (in.)
Figure 19 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Bottom of Weld, Downhill 0.0 0.5 1.o 1.5 2.0 Distance from Bottom of Weld (in.)
LIC-05-0043 Page 18 5.0 Potential Lack of 100% Eddv Current Coverage of the ICI Penetration Nozzle J-Groove Welds FCS has six ICI nozzles oriented around the outer periphery of the RPV head at the steepest angle (54.4") of any of the 48 nozzles. This steep angle and the original design information on the weld configuration indicated that up to 10" segments of the required inspection region at the downhill side (toes) of the J-groove welds on the ICI nozzles could not be inspected without great hardship, per the information supplied in Reference 9. Not inspecting these 10" segments would still result in 97.2% inspection coverage of the total ICI J-groove weld area.
OPPD has recently used a micro-camera to look more closely at the downhill area of several J-groove welds and has observed additional weld passes made on these welds that were beyond the original design, resulting in a flatter aspect to the J-weld in this steep downhill region. Therefore, there appears to be less potential for probe lift off than indicated on the drawings. Consequently, more inspection coverage of the 10" segments by eddy current testing may be available than initially estimated. OPPD plans to examine as much of the 360" area as possible for all CEDM nozzle penetrations.
OPPD has completed the 2005 RFO bare metal visual inspection on all the RPV head penetrations and found no evidence of any leakage. During the 2005 RFO, OPPD has also completed eddy current testing of the J-groove welds of all the CEDM nozzles and, based on analysis complete at the time this letter was signed (39 out of 41 CEDM nozzles), has found no indications of cracking. No indications of cracking in the CEDM J-groove welds is significant because the stresses in these welds are higher than the stresses in the J-groove welds of the ICI nozzles. The stresses at the toe of the J-groove weld for the ICI nozzles are lower primarily because the offset of weld geometry precludes a longer weld and because the toe of the weld is further away from the high stress area of the J-groove weld. Therefore, cracking would be expected to occur first in the CEDM J-groove welds, and would be less likely in the ICI J-groove welds.
Finally, no incidence of cracking in the ICI J-groove welds resulting in leakage has been found in any other Combustion Engineering (CE) plants to date. This lack of cracking at other CE plants corroborates the stress analyses performed at FCS.
In conclusion, lack of inspection coverage for 2.8% of the total area for the J-groove welds of the ICI nozzles is not significant. OPPD considers that obtaining inspection data for these areas would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per the information supplied in Reference 9. The completed inspection coverage area is sufficient to maintain structural integrity over the single cycle of operation prior to RFV head replacement during the Fall 2006 RFO. Information regarding the completed inspection coverage for the ICI nozzles will be submitted to the NRC
LIC-05-0043 Page 19 as part of the status report of the nozzle penetration inspections on or about May 1,2005.
6.0 References
- 1. Dominion Engineering ~ c Calculation
. Number C-8718 Fort Calhoun CEDM and ICI Nozzle Stress Analysis, April 2005, Rev. 0.
- 2. Materials Reliability Program: Generic Evaluation of Examination Coverage Requirements for Reactor Pressure Vessel Head Penetration Nozzles (MRP-99, EPRI, Palo Alto, CA, 2003. 1009129.
- 3. CE NPSD-903-P, CEOG Task 730, CEOG Program to Address Alloy 600 Cracking of CEDM Penetrations, Subtask 1, Nozzle Evaluation, February 1993.
- 4. USNRC Letter, R. Barrett to A. Marion, Flaw Evaluation Guidelines, April 11,2003.
- 5. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick Wall Alloy 600 Material (MRP-55) Revision 1, EPRI, Palo Alto, CA, November 2002. 1006695.
- 6. ASME Code Section XI, 1998 Edition with 2000 Addenda, Article L-3210.
- 7. Letter from R. William Borchardt (NRC) to Ross Ridenoure (OPPD) dated February 20, 2004, Issuance of First Revised NRC Order (EA-03-009)
Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (NRC-04-0022) (ML040220181).
- 8. EPRI MRP-48, PWR Materials Reliability Program Response to NRC Bulletin 2001-01, August 2001.
- 9. Letter from Ralph L. Phelps (OPPD) to Document Control Desk (NRC) dated April 7, 2005, Fort Calhoun Station Unit No. 1, Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (LIC-05-0040).