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Category:Letter type:CNRO
MONTHYEARCNRO-2021-00023, Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-10-0606 October 2021 Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L CNRO-2021-00002, Entergy Operations, Inc. - Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-01-28028 January 2021 Entergy Operations, Inc. - Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L CNRO-2020-00023, Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2020 Summary2020-12-17017 December 2020 Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2020 Summary CNRO-2020-00016, Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2020-08-12012 August 2020 Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) CNRO-2019-00030, Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2019 Summary2019-12-30030 December 2019 Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2019 Summary CNRO-2019-00010, Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3))2019-03-28028 March 2019 Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2019-00007, Grand Gulf Nuclear Station; and Waterford 3 Steam Electric Station - Decommissioning Funding Status Report2019-03-28028 March 2019 Grand Gulf Nuclear Station; and Waterford 3 Steam Electric Station - Decommissioning Funding Status Report CNRO-2019-00013, Entergy Response to Confirmatory Order EA-17-132/EA-17-1532019-03-21021 March 2019 Entergy Response to Confirmatory Order EA-17-132/EA-17-153 CNRO-2019-00003, Application to Revise Technical Specifications to Adopt TSTF-529 11 Clarify Use and Application Rules, 11 Revision 42019-01-31031 January 2019 Application to Revise Technical Specifications to Adopt TSTF-529 11 Clarify Use and Application Rules, 11 Revision 4 CNRO-2019-00002, Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-01-31031 January 2019 Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 CNRO-2018-00051, Grand Gulf Nuclear Station, Palisades Nuclear Plant, River Bend Station, Pilgrim Nuclear Station and Waterford 3 - Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element K2018-12-28028 December 2018 Grand Gulf Nuclear Station, Palisades Nuclear Plant, River Bend Station, Pilgrim Nuclear Station and Waterford 3 - Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element K CNRO-2018-00039, Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element J2018-09-27027 September 2018 Entergy Response to Confirmatory Order EA-17-132/EA-17-153, Element J CNRO-2018-00011, Submittal of Joint Plant Access Shared Information Table2018-03-19019 March 2018 Submittal of Joint Plant Access Shared Information Table CNRO-2017-00022, Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 12017-11-17017 November 2017 Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 1 CNRO-2017-00010, Entergy Operations, Inc. - Decommissioning Funding Status Report2017-03-31031 March 2017 Entergy Operations, Inc. - Decommissioning Funding Status Report CNRO-2017-00007, Nuclear Property Damage Insurance (10 CFR 50.54(w)(3))2017-03-30030 March 2017 Nuclear Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2016-00023, Request EN-ISI-16-1: Use of Later Edition/Addenda of Amse Section XI Code for ISI Related Activities2016-12-0101 December 2016 Request EN-ISI-16-1: Use of Later Edition/Addenda of Amse Section XI Code for ISI Related Activities CNRO-2016-00017, Entergy Nuclear Operations - Request for Approval of Change to the Quality Assurance Program Manual2016-07-19019 July 2016 Entergy Nuclear Operations - Request for Approval of Change to the Quality Assurance Program Manual CNRO-2016-00016, Entergy - Response to RAI Questions and Submittal of Relief Request EN-ISI-15-1 (Updated)2016-06-0202 June 2016 Entergy - Response to RAI Questions and Submittal of Relief Request EN-ISI-15-1 (Updated) CNRO-2016-00008, Entergy - Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3))2016-03-30030 March 2016 Entergy - Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2016-00005, Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM)2016-02-25025 February 2016 Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM) CNRO-2016-00002, Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Secti2016-01-29029 January 2016 Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Section Xl, CNRO-2015-00002, Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 12015-12-0404 December 2015 Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 1 CNRO-2015-00022, Entergy - Relief Request EN-ISI-15-1: Proposed Alternative to Maintain ISI-Related Activities to the 2001 Edition/2003 Addendum of ASME Section Xl Code2015-11-20020 November 2015 Entergy - Relief Request EN-ISI-15-1: Proposed Alternative to Maintain ISI-Related Activities to the 2001 Edition/2003 Addendum of ASME Section Xl Code CNRO-2015-00023, Request for Approval of Change to the Entergy Quality Assurance Program Manual (QAPM)2015-11-0606 November 2015 Request for Approval of Change to the Entergy Quality Assurance Program Manual (QAPM) CNRO-2015-00018, Entergy - Relief Request Number RR EN-15-2 - Proposed Alternative to Use ASME Code Case N-786-1, Alternative Requirements for Sleeve Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping, Section XI, Division 1.2015-08-20020 August 2015 Entergy - Relief Request Number RR EN-15-2 - Proposed Alternative to Use ASME Code Case N-786-1, Alternative Requirements for Sleeve Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping, Section XI, Division 1. CNRO-2015-00017, Entergy Submits Relief Request RR EN-15-1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Waster Service, Section XI, D2015-06-0505 June 2015 Entergy Submits Relief Request RR EN-15-1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Waster Service, Section XI, Division CNRO-2015-00012, Arkansas, Units 1 & 2, Grand Gulf, River Bend, Waterford, Unit 3, Indian Point, Units 1, 2, & 3, Pilgrim, James A. FitzPatrick, Big Rock, Palisades - Submittal of Revision of Quality Assurance Program Manual2015-04-30030 April 2015 Arkansas, Units 1 & 2, Grand Gulf, River Bend, Waterford, Unit 3, Indian Point, Units 1, 2, & 3, Pilgrim, James A. FitzPatrick, Big Rock, Palisades - Submittal of Revision of Quality Assurance Program Manual CNRO-2015-00010, Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3))2015-03-27027 March 2015 Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2015-00011, Grand Gulf Nuclear Station Unit 1 and Waterford 3 Steam Electric Station Submittal of Decommssioning Funding Status Report for 20142015-03-27027 March 2015 Grand Gulf Nuclear Station Unit 1 and Waterford 3 Steam Electric Station Submittal of Decommssioning Funding Status Report for 2014 CNRO-2014-00009, Response to Request for Additional Information for Review of the Decommissioning Funding Plans for Entergy Operations, Inc., Independent Spent Fuel Storage Installations2014-07-24024 July 2014 Response to Request for Additional Information for Review of the Decommissioning Funding Plans for Entergy Operations, Inc., Independent Spent Fuel Storage Installations CNRO-2014-00003, Entergy Nuclear Operations, Inc., Annual Report for Quality Assurance Program Manual Changes Under 10 CFR 50.54(a)(3) and 10 CFR 72.140(d) Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G2014-04-25025 April 2014 Entergy Nuclear Operations, Inc., Annual Report for Quality Assurance Program Manual Changes Under 10 CFR 50.54(a)(3) and 10 CFR 72.140(d) Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G CNRO-2014-00004, Unit 3, Letter (CNRO-2014-00002/ENOC-14-00010) of Same Subject 03/31/20142014-04-0303 April 2014 Unit 3, Letter (CNRO-2014-00002/ENOC-14-00010) of Same Subject 03/31/2014 CNRO-2014-00001, Completion of Change to the Entergy Quality Assurance Program Manual with Regards of Confirmatory Order EA-11-0962014-02-26026 February 2014 Completion of Change to the Entergy Quality Assurance Program Manual with Regards of Confirmatory Order EA-11-096 CNRO-2013-00007, Annual Report for Quality Assurance Program Manual Changes Under 10 CFR 50.54(a) and 10 CFR 72.140(d) Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G2013-04-26026 April 2013 Annual Report for Quality Assurance Program Manual Changes Under 10 CFR 50.54(a) and 10 CFR 72.140(d) Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G CNRO-2013-00008, Arkansas, Units 1, 2 & ISFSI, Grand Gulf, ESP and ISFSI, River Bend, Unit 1 and ISFSI and Waterford 3 and ISFSI, Application for Order Approving Transfers of Licenses and Conforming License and ESP Amendments2013-04-16016 April 2013 Arkansas, Units 1, 2 & ISFSI, Grand Gulf, ESP and ISFSI, River Bend, Unit 1 and ISFSI and Waterford 3 and ISFSI, Application for Order Approving Transfers of Licenses and Conforming License and ESP Amendments CNRO-2013-00006, Decommissioning Funding Status Report - Entergy Operations, Inc2013-03-29029 March 2013 Decommissioning Funding Status Report - Entergy Operations, Inc CNRO-2013-00004, Entergy Operations, Inc. and Entergy Nuclear Operations, Inc., Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3))2013-03-25025 March 2013 Entergy Operations, Inc. and Entergy Nuclear Operations, Inc., Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2013-00002, Responses to NRC Request for Additional Information Regarding Application for Order Approving Transfers of Licenses and Conforming License and ESP Amendments2013-01-29029 January 2013 Responses to NRC Request for Additional Information Regarding Application for Order Approving Transfers of Licenses and Conforming License and ESP Amendments CNRO-2012-00009, Entergy Responses to Request for Additional Information Regarding Changes to the Quality Assurance Program Manual & Associated Plant Technical Specifications Re Staff Qualifications2012-11-20020 November 2012 Entergy Responses to Request for Additional Information Regarding Changes to the Quality Assurance Program Manual & Associated Plant Technical Specifications Re Staff Qualifications CNRO-2012-00007, Grand Gulf Nuclear Station & Waterford 3 Steam Electric Station Status of Decommissioning Funding - Entergy Operations, Inc2012-10-15015 October 2012 Grand Gulf Nuclear Station & Waterford 3 Steam Electric Station Status of Decommissioning Funding - Entergy Operations, Inc CNRO-2012-00004, Request for Additional Information Regarding Changes to the Entergy Quality Assurance Program Manual (QAPM) and Associated Plant Plant Technical Specifications Regarding Staff Qualifications2012-05-21021 May 2012 Request for Additional Information Regarding Changes to the Entergy Quality Assurance Program Manual (QAPM) and Associated Plant Plant Technical Specifications Regarding Staff Qualifications CNRO-2012-00003, Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(d) Notification of Application of Approved Appendix B to 10CFR72 Subpart G2012-04-30030 April 2012 Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(d) Notification of Application of Approved Appendix B to 10CFR72 Subpart G CNRO-2011-00006, Request for Approval of Change to the Entergy Quality Assurance Program Manual (QAPM) and Associated Plant Technical Specifications Regarding Staff Qualifications2011-12-13013 December 2011 Request for Approval of Change to the Entergy Quality Assurance Program Manual (QAPM) and Associated Plant Technical Specifications Regarding Staff Qualifications CNRO-2011-00007, 2011 Recertification of Foreign Ownership Control or Interest2011-10-0606 October 2011 2011 Recertification of Foreign Ownership Control or Interest CNRO-2011-00005, Response to Request for Additional Information on 2011 Decommissioning Fund Status Report2011-07-0505 July 2011 Response to Request for Additional Information on 2011 Decommissioning Fund Status Report CNRO-2011-00004, Entergy - Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(d) Notification of Application of Approved Appendix B to 10CFR72 Subpart G2011-04-29029 April 2011 Entergy - Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(d) Notification of Application of Approved Appendix B to 10CFR72 Subpart G CNRO-2011-00001, Status of Decommissioning Funding for Plants Operated by Entergy Operations, Inc2011-03-31031 March 2011 Status of Decommissioning Funding for Plants Operated by Entergy Operations, Inc CNRO-2011-00002, Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3))2011-03-31031 March 2011 Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) CNRO-2010-00002, Entergy Nuclear Operations, Inc., Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(e) Notification of Application of Approved Appendix B to 10CFR72 Subpart G2010-04-29029 April 2010 Entergy Nuclear Operations, Inc., Annual Report for Quality Assurance Program Manual Changes Under 10CFR50.54(a)(3) and 10CFR72.140(e) Notification of Application of Approved Appendix B to 10CFR72 Subpart G 2021-10-06
[Table view] Category:Report
MONTHYEAR1CAN072401, Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply2024-07-0202 July 2024 Request for Review and Approval of Changes to the Safety Analysis Report and to a Confirmatory Order Clarifying an Alternate Means of Compliance for Pressurizer Heaters Emergency Power Supply 1CAN062304, Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation2023-06-29029 June 2023 Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23180A1082023-06-20020 June 2023 ANO Unit 1 SAR Amendment 31, TRM, TS Bases, 10 CFR 50.59 Report, and Commitment Change Summary Report 1CAN062302, Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version2023-06-20020 June 2023 Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version ML23088A2172022-12-31031 December 2022 Relief Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 - Technical Report, ANP-4023NP, Revision 0, December 2022 2CAN022202, Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods2022-02-24024 February 2022 Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods 0CAN102102, Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis2021-10-0606 October 2021 Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis CNRO-2021-00023, Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-10-0606 October 2021 Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L ML21272A3032021-09-30030 September 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Westinghouse Evinci; Micro-Reactor Tabletop Exercise Report ML21237A0512021-08-25025 August 2021 Follow-on Risk Informed Performance Based Implementation Guidance Needed for Advanced Non-Light Water Reactors ML21081A1922021-06-30030 June 2021 Enclosure - USNRC-CNSC Joint Report Concerning X-Energy's Reactor Pressure Vessel Construction Code Assessment 2CAN062103, Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval2021-06-29029 June 2021 Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval 0CAN052102, Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes2021-05-10010 May 2021 Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes ML21272A3382021-04-0101 April 2021 Technology Inclusive Content of Application Project (Ticap) for Non-Light Water Reactors Versatile Test Reactor Ticap Tabletop Exercise Report ML21090A0332021-03-31031 March 2021 Historical Context and Perspective on Allowable Stresses and Design Parameters in ASME Section III, Division 5, Subsection Hb, Subpart B (ANL/AMD-21/1) ML21083A1362021-03-23023 March 2021 Completed Activities ML21083A1382021-03-22022 March 2021 Rulemaking ML21083A1402021-03-22022 March 2021 Strategy 2 ML21083A1412021-03-22022 March 2021 Strategy 3 ML21083A1442021-03-22022 March 2021 Strategy 6 ML21083A1422021-03-22022 March 2021 Strategy 4 ML21083A1432021-03-22022 March 2021 Strategy 5 ML21083A1372021-03-22022 March 2021 NEIMA Reporting ML21083A1392021-03-22022 March 2021 Strategy 1 ML21014A2672021-01-14014 January 2021 Preapplication Engagement to Optimize Application Reviews January 12 Version - Copy 1CAN032001, Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values2020-03-19019 March 2020 Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values 0CAN121901, Summary of Lost Specimens Investigation Report2019-12-0303 December 2019 Summary of Lost Specimens Investigation Report ML18215A1782018-06-30030 June 2018 WCAP-18169-NP, Rev 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation. ML17214A0292018-02-12012 February 2018 Staff Assessment of Flooding Focused Evaluation (CAC Nos. MF9809 and MF9810) ML17291A0092017-10-26026 October 2017 Staff Assessment Regarding Program Plan for Aging Management for Reactor Vessel Internals (CAC No. MF8155; EPID L-2016-LRO-0001) ML17236A1792017-08-22022 August 2017 Arkansas, Units 1 and 2, ANO Emergency Plan On-Shift Staffing Analysis Report, Revision 2 0CAN081703, Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report2017-08-16016 August 2017 Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report ML17167A0832017-06-28028 June 2017 Arkansas Nuclear One, Unit 2 - Review of Commitment Submittal for License Renewal Regarding Nickel-Based Alloy Aging Management Program Plan (CAC No. MF8154) 0CAN061701, Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis2017-06-0707 June 2017 Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis 0CAN051704, Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One.2017-03-13013 March 2017 Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One. 2CAN011703, Submittal of Additional Protocol Report2017-01-26026 January 2017 Submittal of Additional Protocol Report ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 0CAN121602, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 32016-12-30030 December 2016 Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 3 ML17003A2902016-12-20020 December 2016 Areva, Inc. - Engineering Information Record - Arkansas Nuclear One HFE-High Frequency Confirmation Report ML16365A0272016-10-31031 October 2016 ANP-3486NP, Revision 0, MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (ANO-1). ML16293A5842016-09-30030 September 2016 WCAP-18166-NP, Revision 0, Analysis of Capsule 284 from the Entergy Operations, Inc. Arkansas Nuclear One, Unit 2 Reactor Vessel Radiation Surveillance Program. 1CAN091601, Submittal of Initial Examination Completion of Post-Examination Analysis2016-09-0101 September 2016 Submittal of Initial Examination Completion of Post-Examination Analysis ML16202A1672016-07-0505 July 2016 Report 1500227.401, PWR Internals Aging Management Program Plan. ML16147A3242016-05-31031 May 2016 ANP-3417NP, Rev. 1, MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One, Unit 1. ML16004A1792015-12-31031 December 2015 Attachment 2, ANP-3418NP, Revision 0, Arkansas Nuclear One Unit 1 Reactor Vessel Internals License Renewal Scope and MRP-189, Revision 1 Comparison (MRP-227-A Action Item 2) Licensing Report. (Non-Proprietary) ML15278A0242015-09-28028 September 2015 Attachment 2, Areva Document ANP-3417NP, Revision 0, MRP-227-A Applicant / Licensee Action Item No. 7 Analysis for Arkansas Nuclear One, Unit 1 (Non-Proprietary), Attachment 3, Affidavit, and Attachment 4, List of Commitments ML15099A1522015-04-16016 April 2015 Review of Spring 2014 Steam Generator Tube Inspection Report, Inspection During Refueling Outage 2R23 ML15071A0552015-03-31031 March 2015 ANP-3300Q2NP, Revision 0, Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1. ML15086A0242015-03-25025 March 2015 ANP-3300Q3NP, Revision 0 to Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 (Non-Proprietary ML15043A1032015-02-10010 February 2015 Areva Document ANP-3383NP, Response to Request for Additional Information for the Reactor Pressure Vessel Internals Aging Management Program Plan for Arkansas Nuclear One Unit 1 2024-07-02
[Table view] Category:Miscellaneous
MONTHYEARCNRO-2021-00023, Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-10-0606 October 2021 Entergy Operations, Inc. - Supplement to CNRO-2021-00002, Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L 0CAN052102, Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes2021-05-10010 May 2021 Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes ML21083A1362021-03-23023 March 2021 Completed Activities ML21083A1432021-03-22022 March 2021 Strategy 5 ML21083A1412021-03-22022 March 2021 Strategy 3 ML21083A1372021-03-22022 March 2021 NEIMA Reporting ML21083A1382021-03-22022 March 2021 Rulemaking ML21083A1392021-03-22022 March 2021 Strategy 1 ML21083A1402021-03-22022 March 2021 Strategy 2 ML21083A1422021-03-22022 March 2021 Strategy 4 ML21083A1442021-03-22022 March 2021 Strategy 6 ML21014A2672021-01-14014 January 2021 Preapplication Engagement to Optimize Application Reviews January 12 Version - Copy 1CAN032001, Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values2020-03-19019 March 2020 Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values 0CAN121901, Summary of Lost Specimens Investigation Report2019-12-0303 December 2019 Summary of Lost Specimens Investigation Report ML17214A0292018-02-12012 February 2018 Staff Assessment of Flooding Focused Evaluation (CAC Nos. MF9809 and MF9810) ML17291A0092017-10-26026 October 2017 Staff Assessment Regarding Program Plan for Aging Management for Reactor Vessel Internals (CAC No. MF8155; EPID L-2016-LRO-0001) ML17167A0832017-06-28028 June 2017 Arkansas Nuclear One, Unit 2 - Review of Commitment Submittal for License Renewal Regarding Nickel-Based Alloy Aging Management Program Plan (CAC No. MF8154) 2CAN011703, Submittal of Additional Protocol Report2017-01-26026 January 2017 Submittal of Additional Protocol Report ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 0CAN121602, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 32016-12-30030 December 2016 Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 3 ML17003A2902016-12-20020 December 2016 Areva, Inc. - Engineering Information Record - Arkansas Nuclear One HFE-High Frequency Confirmation Report 1CAN091601, Submittal of Initial Examination Completion of Post-Examination Analysis2016-09-0101 September 2016 Submittal of Initial Examination Completion of Post-Examination Analysis ML16147A3242016-05-31031 May 2016 ANP-3417NP, Rev. 1, MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One, Unit 1. ML15099A1522015-04-16016 April 2015 Review of Spring 2014 Steam Generator Tube Inspection Report, Inspection During Refueling Outage 2R23 ML15086A0242015-03-25025 March 2015 ANP-3300Q3NP, Revision 0 to Response to Request for Additional Information on Reactor Coolant System Pressure/Temperature and Low Temperature Overpressure Protection System Limits to 54 EFPY for Arkansas Nuclear One, Unit 1 (Non-Proprietary ML15043A1032015-02-10010 February 2015 Areva Document ANP-3383NP, Response to Request for Additional Information for the Reactor Pressure Vessel Internals Aging Management Program Plan for Arkansas Nuclear One Unit 1 0CAN021501, Spent Fuel Storage Radioactive Effluent Release Report for 20142015-02-0303 February 2015 Spent Fuel Storage Radioactive Effluent Release Report for 2014 ML15028A4932015-01-31031 January 2015 ANP-3281Q1NP, Rev. 0, Response to NRC Request for Additional Information on Time-Limited Aging Analysis for the ANO-1 Reactor Vessel Internals, Attacment 2 ML14329A4262014-10-22022 October 2014 U.S. Nuclear Regulatory Commission, Record of Review, Arkansas Nuclear One Unit 2 (ANO-2), LAR Attachment V-Table V-1 Fire PRA Facts and Observations (F&Os). ML14223A8032014-08-12012 August 2014 Review of Steam Generator Tube Inspection Report for Refueling Outage 1R24 0CAN031404, ANO, Units 1 & 2 - Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Nttf)Review of Insights from the Fukushima2014-03-28028 March 2014 ANO, Units 1 & 2 - Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Nttf)Review of Insights from the Fukushima Da ML14051A1882014-03-18018 March 2014 Staff Assessments of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to Fukushima Dai-Ichi Nuclear Power Plant Accident (TAC MF0090-MF0091) 0CAN021403, Units 1 and 2 - 10 CFR 50.59 Summary Report and Commitment Change Summary Report2014-02-11011 February 2014 Units 1 and 2 - 10 CFR 50.59 Summary Report and Commitment Change Summary Report ML14028A1992014-01-16016 January 2014 Units 1 and 2, Unsatisfactory Blind Quality Assurance Drug Testing Samples 2CAN091302, Updated Seismic Walkdown Report2013-09-30030 September 2013 Updated Seismic Walkdown Report 1CAN031305, Cycle 24 COLR, Revision 62013-03-13013 March 2013 Cycle 24 COLR, Revision 6 ML12159A5972012-06-18018 June 2012 Closeout of NRC Bulletin 2011-01, Mitigating Strategies 0CAN021204, Units 1 and 2, Third Five-Year Surveillance of the First Ventilated Storage Cask2012-02-29029 February 2012 Units 1 and 2, Third Five-Year Surveillance of the First Ventilated Storage Cask 2CAN011202, Additional Protocol Report2012-01-31031 January 2012 Additional Protocol Report ML1130703872011-11-0202 November 2011 Fall 2011 RFO - Steam Generator Inspection Call - Part II, Summary of the ANO1 1R23 Steam Generator Inspection 2CAN091103, CFR 50.59 Summary Report2011-09-16016 September 2011 CFR 50.59 Summary Report 1CAN041105, Request for Use of Non-ASME Code Repair to Service Water Piping in Accordance with Generic Letter 90-05 Relief Request ANO1-R&R-0162011-04-20020 April 2011 Request for Use of Non-ASME Code Repair to Service Water Piping in Accordance with Generic Letter 90-05 Relief Request ANO1-R&R-016 0CAN101003, 10 CFR 50.59 Summary Report and Commitment Change Summary Report2010-10-0707 October 2010 10 CFR 50.59 Summary Report and Commitment Change Summary Report 1CAN091001, 10 CFR 50.46 Report - Significant Change in Peak Cladding Temperature2010-09-0808 September 2010 10 CFR 50.46 Report - Significant Change in Peak Cladding Temperature 0CAN031001, Units 1 & 2, Unsatisfactory Laboratory Testing Report2010-03-18018 March 2010 Units 1 & 2, Unsatisfactory Laboratory Testing Report 2CAN011003, Submittal of Us Dept. of Commerce, Bureau of Industry and Security. Additional Protocol Report2010-01-28028 January 2010 Submittal of Us Dept. of Commerce, Bureau of Industry and Security. Additional Protocol Report 0CAN100901, Units 1 & 2, Unsatisfactory Laboratory Testing Report2009-10-13013 October 2009 Units 1 & 2, Unsatisfactory Laboratory Testing Report 0CAN050903, Annual 10 CFR 50.46 Report for Calendar Year 2008 Emergency Core Cooling System Evaluation Changes2009-05-15015 May 2009 Annual 10 CFR 50.46 Report for Calendar Year 2008 Emergency Core Cooling System Evaluation Changes 2CAN100802, CFR 50.59 Summary Report for Period Ending October 6, 20082008-10-13013 October 2008 CFR 50.59 Summary Report for Period Ending October 6, 2008 2CAN070804, Cycle 20 Startup Report2008-07-0303 July 2008 Cycle 20 Startup Report 2021-05-10
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{{#Wiki_filter:Entergy Operations, Inc.
EntergyJackson, 1340 Echelon Parkway MS 39213-8298 Tel 601 368 5758 Michael A. Krupa Director Nuclear Safety & Licensing CNRO-2002-00042 July 24, 2002 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Entergy Operations, Inc.
Control Rod Drive Mechanism Fracture Mechanics Analysis Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51
Dear Sir or Madam:
Enclosed please find Engineering Report #EP-01-002-00, Engineering Report for ASME Code Section XI Analysis of ANO 1 CRDM number 56. This report provides the analytical evaluation performed for the remaining flaw in control rod drive mechanism (CRDM) penetration #56 in the reactor pressure vessel head penetration at Arkansas Nuclear One, Unit 1 (ANO-1). Entergy Operations, Inc. is providing this report for your information in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section Xl Article IWB-3134(b) of the 1992 Edition.
Should you have any questions, please contact Steve Lewis at (601) 368-5444 or Guy Davant at (601) 368-5756.
Very truly yours, MAK/GHD/baa
Enclosure:
Engineering Report #EP-01-002-00 cc: Mr. C. G. Anderson (ANO) (w/o)
Mr. W. R. Campbell (ECH) (w/o)
Mr. G. A. Williams (ECH) (w/o)
Mr. R. L. Bywater, NRC Senior Resident Inspector (ANO) (w/o)
Mr. E. W. Merschoff, NRC Region IV Regional Administrator (w/o)
Mr. W. D. Reckley, NRR Project Manager (ANO-1) (w/o) r\O
CNRO-2002-00042 ENCLOSURE ENGINEERING REPORT #EP-01-002-00 EngineeringReport for ASME Code Section Xl Analysis of ANO 1 CRDM number 56
Engineering Report No.: EP-01-002-00 Page I of 18 ENTERGY OPERATIONS Engineering Report For ASME Code Section XI Analysis of ANO 1 CRDM number 56 APPLICABLE SITES ANO Unit 1: X GGNS: El W-3: El ANO Unit 2: E1 RBS: El ECH: X Safety Related: X Yes No Prepared by: Date: 11/312/001 Reviewed by: 'Date:
er Reviewed by: Date:; .314/0Z t I Z Approved by:
1blzY Snrio/Reviewer Date: /3/ 102 sposibe DE Manage (for multiple site reports only)
Engineering Report EP-01-002-00 Page 2 of 18 RECOMMENDATION FOR APPROVAL Engineering Report No.: *"9-- O---002.-O Reviews Completed: Date: -4"25-0-2 Concurrence: Date: ____
/~e Manager, ANO Concurrence: N/A Date:
Responsible Manager, GGNS Concurrence: N/A Date:
Responsible Manager, W3 Concurrence: N/A Date:
Responsible Manager, RBS
Engineering Report EP-01-002-00 Page 3 of 18 Table of Contents Section Title Paze No.
List of Appendices 3 List of Tables 4 List of Figures 4 1.0 Introduction 5 2.0 Characterization of Nozzle and Flaw 5 3.0 Nozzle Stresses 5 4.0 Fracture Mechanics Evaluation 7 5.0 Conclusions 17 6.0 References 18 List of Appendices Appendix No. Title Number of Pa2es I Limit Load Analysis to Determine Allowable Flaw Size 10 2 Fatigue Crack Growth Analysis for Slice I - Pre-repair 15 Condition 3 Fatigue Crack Growth Analysis for Slice II - Pre-repair 15 Condition 4 Fatigue Crack Growth Analysis for Slice III- Pre-repair 15 Condition 5 Fatigue Crack Growth Analysis for Slice IV - Pre-repair 15 Condition 6 Fatigue Crack Growth Analysis for Slice I - Post-repair 15 Condition 7 Fatigue Crack Growth Analysis for Slice II - Post-repair 15 Condition 8 Fatigue Crack Growth Analysis for Slice III - Post-repair 15 Condition 9 Fatigue Crack Growth Analysis for Slice IV - Post-repair 15 Condition 10 Fatigue Crack Growth Analysis for Slice II - Post-repair 15 Condition for 2,100 cycles 11 Technical Review checklist and Comment Sheets 5
Engineering Report EP-0 1-002-00 Page 4 of 18 List of Tables Table No. Title Paee No.
I Surface Residual Stresses in the Hoop direction 7 2 Results of Fatigue Crack growth Evaluation 16 List of Figures Fi2ure No. Title Pae No.
I Finite Element Analysis stress Contours 6 2 Allowable Flaw Size - Limit Load results Comparison 10 3 Allowable Flaw Size Curves (ASME Section XI & BS-7910) I1 4 Fatigue Crack Growth - Depth {700 cycles} 15 5 Fatigue Crack Growth - Length {700 Cycles} 15 6 Fatigue Crack Growth - Depth {2,100 cycles} 16 7 Comparison of Final Flaw Size with Allowable Flaw Size 17
Engineering Report EP-01-002-00 Page 5 of 18 1.0
Introduction:
The visual inspection of Control Rod Drive Mechanism (CRDM) nozzles, from the outside surface, at Arkansas Nuclear One (ANO) Unit 1 revealed the presence of white deposits at the reactor vessel head penetration (RVHP). Non-destructive examinations (NDE) performed (References 1, 3, and 4) revealed defects in the weld and a part through-wall axial indication in the CRDM nozzle. The axial defect in the nozzle was 0.2 inch deep and extended 1.3 inches above the weld. The repair method utilized was to excavate the weld cavity without disturbing the weld butter of the original weld. The cavity was then repaired utilizing a qualified welding technique. The repair performed on the nozzle weld ensures that the reactor coolant will not wet the remaining flaw. Hence the evaluation of the remaining flaw can be limited to fatigue crack growth owing to normal operating conditions. The evaluation the flaw to ensure continued safe operation in accordance with the requirements of ASME Section XI, IWB-3640 (Reference 6) by fatigue was performed. This report provides the details of the required analysis and the results of the evaluation performed.
2.0 Characterization of Nozzle and Flaw:
The nozzle dimensions, material, and required operational conditions were obtained from Reference 3. The dimension of the flaw remaining in service was:
Flaw Depth: 0.200 inch Flaw Length: 1.3 inches 3.0 Nozzle Stresses:
The operating stresses for the normal and upset conditions were obtained from Reference
- 4. The distribution of residual stresses for the region of interest, obtained from the finite element analysis (FEA) contour plots after the repair was performed, is shown in Figure 1, below. This figure shows the circumferential iso-stress contours, which affect an axial flaw.
The residual stress magnitude was found to be higher on the inner diameter (ID) surface when compared to the outer diameter (OD) surface.
Engineering Report EP-01-002-00 Page 6 of 18 ANSYS 5.6 ASP 4 2001 05.45.27 PLOT NO. 101 4OfAL SOLUTION TINE4Rflfl2 RSYS= 11 D1* 047141 S'C -91099
- 30357 T. 10000
- 3 00005 o0000 4ooo
...000 AN~l CRD(3567Sk42 ?5.E0. After Rep l~~g92 e r~ h Figure 1: Showing the residualstress contours (hoop stress) in the vicinity of/he axialfiaw. The finer mesh represents the weld region and the flaw, is locatedjust above the weld region on the OD surface. The axial section of the tube, where the repair was performwedand where the flaw, remains, is on the left side (down hill),
The residual stresses for the as-built and the weld-repaired conditions were obtained from Reference 5. The region where the flaw was located was bounded by four azimuthal slices, which were obtained from detailed EFA. The maximum stress values are located at the 1 6 'h, 17, ,or 18S row of nodes. The axial location was at the middle of the weld butter plus the rows immediately above the weld. The node numbers in the 5000 series were at the 22.5-degree azimnuth and the node numbers in the 6000 series were at the 45-degree azimuth. The four 5oono slices chosen represent the zone that bounds the flaw location. The hoop stresses on the OD and ID surfaces were obtained from detailed nodal stress files (Reference 5) and are provided in Table I below.
Engineering Report EP-01-002-00 Page 7 of 18 Table 1 Surface Residual Stresses in the Hoop Direction Slice Location Surface Hoop Stress Iksi)
Asbuilt (pre-re air) Condition Post-Repair Condition
@Nodes ID OD ID OD I 43.622 9.372 27.357 -2.977 61501-61506 11 39.397 12.306 33.071 -8.523 51501-51506 11 39.032 8.807 36.073 -13.434 51601-51606 IV 44.203 -0.225 31.340 -16.034 61601-61606 1 The residual stresses determined by the FEA for both the as-built and the post-repair conditions showed that the magnitude of residual stresses was higher on the ID surface than on the OD surface. The table above clearly demonstrates that after the repair the stress distribution, in the vicinity of the axial crack, was reduced in magnitude. This reduction would only affect the R-ratio for the fatigue crack growth.
4.0 Fracture Mechanics Evaluations:
Limit Load Analysis The analysis for flaw tolerance for Inconel-600 material at operating temperatures is based on limit load. The first step is to determine the allowable flaw size (depth and length).
Based on determining the limiting flaw size, crack growth rate analysis can be performed to determine structural integrity. The fatigue crack growth margin is the difference between the allowable flaw size and the expected final flaw size accounting for design fatigue cycles.
Residual stresses are secondary stresses, which only affect the R-ratio in the fatigue analysis and, therefore, are not considered in the limit load analysis. This flaw is captured in the bore of the reactor pressure vessel head and hence limit load-based flaw size is a conservative bounding flaw size.
The critical flaw length, for an axial through wall flaw, was determined to provide a limit for the evaluation. The critical flaw length forms the upper limit on flaw length. This evaluation was based on equation six of Reference 6. The critical flaw length can be defined as:
1',c,= 2
Engineering Report EP-01-002-00 Page 8 of 18 Where:
Lcrit = Criticalflaw length (inch) cy MaterialFlow Stress at OperatingTemperature (ksi)
-Yh Hoop Stress due to InternalOperatingPressure (ksi)
Rm - Mean Radius of Tube (inch)
T = Wall Thickness of Tube (inch).
The critical flaw length was determined to be 13.05 inches.
The allowable flaw size based on limit load criteria were determined with a safety factor (SF) of three (SF-=3.0). This conservative safety factor was used in the analysis, which is higher than the ASME Section XI safety factor of 2.77. Three methods were compared for the determination of the limit load flaw size. The three methods were as follows:
- 1) The British Standard BS-7910, Amendment 1 (Reference 7)
- 2) ASME Section XI, Appendix "C" (Reference 6)
- 3) EPRI Steam Generator Tube Evaluation (Reference 8)
The material flow stress used in the equations was obtained by using the equation derived for Inconel-600 based on testing steam generator tube materials as demonstrated in Reference 8.
The equation for determining flow stress, from Reference 8, is:
r>, = 0.58 * {oY + ua}
Where:
= MaterialFlow Stress at OperatingTemperature (ksi)
-r ay = Material Yield Strength at OperatingTemperature (ksi)
= Materialtensile Strength at OperatingTemperature (ksi)
-ru The hoop stresses used in the analyses were defined using both the thin and thick cylinder formulae. The formulation in the British Standard (Reference 7) uses the thick cylinder approximation whereas the ASME code and the EPRI formulation use the thin cylinder approximation. The hoop stress equation for both formulations are presented below.
2 2 UtroK P*0 R2 + R 2 Thick Cylinder Formula Ro - R 2
-N P *,R Thin Cylinder Formula t
Where:
o'hTK = Hoop Stress using Thick Cylinderformulation(ksi)
'hUTN = Hoop Stress using Thin Cylinderformulation (ksi)
Ro = Outer Radius (inch)
Engineering Report EP-0 1-002-00 Page 9 of 18 Ri Inner Radius (inch)
R Rofor ASME Section XI, and R R, for ERPRI Methods.
The allowable flaw dimensions using the three methods was determined using the equations defined below.
The British Standard method (Reference 7, equation P 11) defines the flaw dimensions by:
d,3" - X *100 Equation in British Standard (Reference 7) 1 -x M,
Where:
dBs = Part Through Wall Flaw Depth (percentof wall thickness)
X = I(r/SFJ/{1.2*rh TK} The factor 1.2 from Reference 7& SF is the safety factor.
MT = Vi1+ 1.6*[l/2]2/R,*t)
Where l is the flaw length and the other variables defined above.
Using the ASME Section X1, Appendix "C" the flaw dimensions can be defined by:
dAs,E = 1-' *100 Equation 5 of Reference 6 (Section C-3420)
Where:
dAstE = Part Through Wall Flaw Depth (percentof wall thickness)
Y = {(u,,v*SF}/qf SF is safetyfactor & other variables defined above MAsME = V-{ + 1.61 *12 /[4*Rm*t] R,, is mean radius & other variables defined above The EPRI method for flaw depths limited to eighty five percent of wall thickness and the flaw dimension can be defines as (rearrange Equation 4-5 [volume 2], Reference 8):
dH'N = [(l + 2
- 1 SF-* (hN*N100 EPRI equation from Reference 8 Where:
dEpem = Part Through-Wall Flaw Depth (percentof wall thickness), and other variables defined above.
The EPRI methods presented in Reference 8 provide two other formulations for the determination of flaw dimensions, which are similar to the ASME method. These methods are
Engineering Report EP-01-002-00 Page 10 of 18 for flaw depths greater than eighty five percent of wall thickness. In these methods the flow stress (af) is replaced by tensile strength (q,). However the hoop stress equation uses the thin wall formulation with inside radius (Rd instead of the outside radius (Rod that is used in the ASME method.
The five formulations, presented in the preceding equations, were used to determine the flaw dimensions in an iterative manner. The internal pressure utilized in the calculations was 2,500 psi. The tube geometry and initial flaw dimensions used were defined in earlier sections.
The material properties used in the analysis were obtained from ASME Code Section 11 Part D tables U and Y (Reference 9). The calculations were performed using Mathcad. The Mathcad worksheets are provided in Appendix I to this report. The results from the calculations, showing a comparison of the five methods, are presented in Figure 2 below.
TW -z
.ir *FLi" SLcC* panopa °nfr Leakage 2
S IT.N5 5r ---
I - - -
7' 21 71 1Q75o 2 F.. j325
- Ro & 2,[Eqlion I ts tfr I (ANCI CRDM56)
S
. .. Rg 3. Equeton 4.5, IEqjtion* 2 Its (sic]
S R¢f 3 oquwjun4-., diFL, [ow ba, 3 hi, Ccl¢i]
¢ telest[
R/3 equataon 6o¢at IEquatko4 masca l ASMESct, Xi. App."C cqusIon b3 fei
- 7. [Equation, els Figure2: Comparison of allowableflaw size determined by the five methods. The method of the British Standard provides the lowest allowableflaw sie.
The results show that the EPRI methods (Reference 8), which were developed for steam generator tubes, show a higher flaw tolerance. The methods used in the ASME Section XI, Appendix "C" (Reference 6) and that in the British standard (Reference 7) provide lower bound flaw sizes. A better comparison between the two lower bound results is presented in Figure 3 below.
C(,z2
Engineering Report EP-01-002-00 Page It of 18 British Standard & ASME Section XI 79 1.3 78 77 76 75 74 C)- 73 PTW 72 71 SPTW 5 70
.69
---- aI 0* 68 67 65 86 66 65 6 6 64 -4 63 62 60 61 60 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 7.5 8 8.5 9 9.5 10 Fi E0.755L 10 Faiw Length {inch}
Method British Standard
------- ASME Section XI Appendix "C"Method Figure 3: Comparison offlaw size by the British Standardmethod and the ASME Section XlAppendix "C" method The British Standardmethodprovidesa lower boundsolution because of the additionalfactorof 1.2 appliedto the allowable hoop stress.
Figure 3 shows that the British Standard method provides the lowest bound solution owing to the additional reduction of the allowable hoop stress for a thick-wall cylinder. The initial flaw length (1.3 inches) is also shown in this figure. The allowed flaw depth at the initial flaw length is 75 % of wall thickness. However, the measured depth was 31.75 % of wall thickness, which is considerably less than the allowable depth. Figure 3 also shows that the maximum allowable flaw depth extends to a flaw length of about 2.6 inches. This suggests a considerable margin for flaw growth. The weld repair performed assures the eradication of the leak path and hence the remaining flaw will not be subjected to wetting by the reactor coolant.
Thus, flaw growth by stress corrosion cracking is mitigated, which implies fatigue due to normal operation as the only viable flaw growth mechanism.
C0o3
Engineering Report EP-01-002-00 Page 12 of 18 Fatigue Crack Growth Evaluation The fatigue crack growth evaluation followed the guidance provided in ASME Code Section XI, Appendices A and C. (Reference 6). The stress intensity factor equations of Appendix A (Reference 6) were numerically modeled in the Mathcad worksheet using the formulations provided in Reference 10. The fatigue crack growth formulation from Appendix C (Reference 6) was modeled in the same Mathcad worksheet. Since the stress intensity factor solution in Appendix A (Reference 6) is for a flat plate geometry, a correction factor from Reference 11 (Mp) was applied. This correction factor provides added conservatism (Reference 12). The evaluations were performed with and without the correction factor in order to ascertain the level of conservatism. In the following paragraphs the basic equations used are described.
The Mathcad worksheets for the evaluations are presented in Appendices 2 through 9.
The equation for stress intensity factor under the combined action of membrane and bending stresses from Appendix A of Reference 6 is:
K,U*-M*= -F*a ,M[F-*-a
=cQM*
K1 ]+Cb h [
Where:
K 1 = Applied Stress Intensity Factor {ksi yin) qm & oab = Membrane and Bending stresses (ksi} as defined below.
Mm & Mb = CorrectionFactorfor membrane and bending stresses a Flaw Depth (inch)
Q Flaw Shape parameterdependent onflaw geometry and appliedstresses Since the applied stress components are dependent on the flaw depth, the model determines these components in a recursive manner using the formulation provided in Reference
- 10. The stress components are based on a linearized stress profile between the two surfaces and flaw depth. The basic equations for the stress components are:
07 or - + U Membrane stress at flaw tip; and a 2 07a U-0.-
f,*
O-r - a - Bending stress at flaw tip Where:
o-, = Stress magnitude at Flaw tip (ksi) ofj, = Stress magnitude at Front Surface (Flaw initiationsurface) {ksi) a & t = Flaw Depth and Wall Thickness respectively (inch)
Engineering Report EP-0 1-002-00 Page 13 of 18 The correction factors (Mm and Mb), that are dependent on the flaw geometry, are numerically determined following the respective equations provided in Reference 10. The equations for these correction factors are detailed in the Mathcad worksheets provided in the Appendices 2 through 9. In a similar manner the flaw shape parameter "Q" was determined using the numerical solution provided in Reference 10. The basic equation for "Q", from Reference 10, is:
2 Q Ek -0.212"{ }2 Where:
Ek = EllipticalIntegralof the second kind and a function of Flaw Geometry U-= Um+ Uh o3, = MaterialYield Strength at temperature.
The correction factor used to apply the flat plate stress intensity factor solution for a cylinder was obtained from Reference 11. The equation for the correction factor is defined as:
a 1+1.255* P 0.0135" Mp- M and M= 2 4*r*t 16*r *t2 "t-a Where:
Ap = CorrectionFactor The equation for fatigue crack growth, obtained from Appendix C of Reference 6, is rearranged to show crack growth increment for a given number of cycles. Thus the total fatigue cycles are divided into a number of blocks and within each block the growth is computed for a set number of cycles. The form of the equation is:
(AK1 )3.3 CGRbhk = CYCblk
CGRblk = Crack Growthper Block (inch)
CYCblk = Number of Fatigue Cycles in a Block C = Scalingparameterto accountfor temperature effect, defined below S Scaling parameterto accountfor R-ratio (Km,,/Kmax)
AKI = Stress Intensity FactorRange {ksi i4n}; as defined below C = (I o 1-009+8 0 12*10 4*T-1 13"10 6.1"2+1.0210 9
,*T1 .; where T is the metal temperature.
S = 1.0 when RO0 S = 1.0 + 1.8*R when 0-:R *<0.79
Engineering Report EP-0 1-002-00 Page 14 of 18 S = -43.35 + 57.97*R when 0.79_*R <1.0 AKI = Kimax - Klmin The parameters in all the equations needed to determine crack growth are dependent on the flaw depth (a). Hence, for each growth increment these parameters would need to be re calculated. The model to determine fatigue crack growth is designed as a recursive computational scheme to capture the interactive relationships. The computational algorithm is shown in the worksheets in Appendices 2 through 9 to this report.
Four axial slices for which the residual hoop stresses both on the ID and OD surfaces were determined to bound the region of the flaw. The FEAs were performed for the two conditions of interest namely: pre-repair and post repair; hence, eight cases were executed for fatigue crack growth to ensure the determination of maximum growth. The stress intensity factors are computed for both the residual and operating stress conditions. The stress intensity factors for residual and operating stress are summed to obtain the maximum stress intensity factor. The minimum stress intensity factor is calculated using the residual stress alone.
The fatigue crack growth is computed in a recursive manner, since the variables and constants used in the model are a function of the flaw geometry. The initial conditions for flaw depth form the starting conditions. The given and fixed conditions of stresses (residual and operational) on the inside and outside surfaces are prescribed. The operating temperature, material properties, and tube geometry are the fixed inputs. The initial fatigue block, consisting of seven operating cycles, uses the initial conditions to determine the stress components (due to residual stress and operating stresses) and the various constants and factors needed for analysis.
The required stress intensity factors are then computed, and using these, the flaw growth for the fatigue block is determined. The numerical algorithm of Reference 10 provides for the determination of stress intensity factors for both the surface and deepest penetration of the flaw.
Two parallel computations to accommodate the geometry of a flat plate and that for a cylinder are performed. Likewise, the fatigue crack growth is performed in parallel. The fatigue crack growth is estimated for both the depth and length dimensions. At the end of the recursive cycle the flaw dimensions (depth and length) are updated for the two geometry conditions and the recursive cycle continues the execution. The evaluation is performed for the required seven hundred operating cycles, requiring one hundred fatigue blocks, and the final flaw dimensions are obtained. Figure 4 below shows the results of flaw depth as a function of the number of fatigue cycles, and Figure 5 shows a similar plot for the flaw length. Both figures are for the highest calculated growth for the post-repair residual stress distribution. In these figures the effect of the correction for cylinder geometry appears to be minimal. Both figures show that the increase in the flaw dimension is sufficiently small and are well below the allowable limiting flaw size.
Engineering Report EP-01-002-00 Page 15 of 18 Crack Extension (Depth) 0 204P.2o05 0.204 0.203 SFatax](j,2)
SFatax ,).0 S
0.201 0.2
,0.2, 0199 L a 100 200 300 400 500 600 700 800 Fataxi 1j 1 ) 707, Number of Cycles No Curvature Correction Curvature -With Conection Figure 4: Fatigue crack growth in the depth directionfor 700 cycles of operation using the operating stresses and the residualstress after the weld repair Crack Length >dc,,sion x
304 305 1 304 1 303 IFatamdtj o3
,j .9)l302 S
I 30]
13 j 3 9 12 o ZO20 3t0 400 w 6Co 7M e0n
.7* FaPtaxl jl .707 Nwnber of C'ycle No NO arvature Con~ectioi With Cirvature Corection Figure 5: Fatigue crack growth in the length directionfor 700 cycles of operationusing the operatingstresses and the residualstress after the weld repair.
Table 2 provides the final flaw sizes for the eight cases evaluated. The data presented show that the expected flaw growth to be very small and that the final flaw dimensions are well within the acceptable size.
co)Ck
Engineering Report EP-01-002-00 Page 16 of 18 Table 2 Results of Fatigue Crack Growth Evaluation Case Initial Final Flaw Dimension (700 cycles)
Number Flaw Dimension Depth (inch) Length (inch)
Residual Stress Depth / With Without With Without Distribution Slice Length Curvature Curvature Curvature Curvature Number (inch) Correction Correction Correction Correction Pre-Repair-I 0.2/1.3 0.20762 0.20732 13078 1.3075 Pre-Repair-ll 0.2/1.3 0.2069 0.20664 13070 1.3067 Pre-Repair-Ill 0.2/1.3 020683 0.20656 1.3069 1.3067 Pre-Repair-IV 0.2/1.3 0.20637 0.20613 1.3065 1.3062 Post-Repair-I 0.2/1.3 0.204194 0.204038 1.30426 1.30410 Post-Repair-Il 0.2/1.3 0.204346 0.204184 1.30442 1.30425 Post-Repair-Ill 0.2/1.3 0.204298 0.204137 1.30437 1.30420 Post Repair-IV 0.2/1.3 0.203818 0.203676 1,30388 1.30374 An additional analysis for expected flaw growth for three times the design operating cycle was performed. The purpose of this analysis was to ascertain the additional expected flaw growth for this extended period of operation. The model used for this analysis was the post repair case that showed the highest growth (Post-Repair II). The result for this evaluation is presented in Figure 6 and the Mathcad worksheet in Appendix 10.
Crack Extension (Depth)
,0-2130.215 O l 0.21 Fataxl 1 2}
1j*g.205 S S
0.195 500 1000 1500 2000 2500 Fataxl(j,i) 2.107x103 Numbet of Cycles No Curvature Correction
-..Correction With Curvature Figure 6: Fatigue crack growth in the depth directionfor 2,100 cycles of operationusing the operatingstresses and the residualstress after the weld repair
-05
Engineering Report EP-01-002-00 Page 17 of 18 5.0
Conclusions:
The critical flaw length for a through-wall axial flaw was determined to be 13.05 inches.
The existing flaw is sufficiently smaller and hence net-section collapse failure of the CRDM tube at the flaw location is not plausible. Figure 7 presents the final results of the evaluation. The final flaw size is compared to the allowable flaw size for both the pre-repair and post repair stress distributions. This figure graphically demonstrates the significant margins that exist for the flaw that was left in service.
Final Flaw Size. Allowable Flaw Size 50 75 70 65 60 55 50 45 40
'5 9
*v 15 2 25 3 ma~w tong~th'{nch}
BS7910 ASME Section XI 0 Post-Repair PriRepair Figure 7: Comparison offinalflaw size with the allowableflaw size, Finalflaw sizes for both the pre-repair and post-repairstress distributionsshow considerable margins to the allowableflaw size COG,
Engineering Report EP-01-002-00 Page 18 of 18 The analysis, presented in this report, demonstrates that sufficient margin is available for fatigue crack growth. The fatigue crack growth analysis showed that the final flaw size for the design life of the plant is well within the acceptance criteria prescribed by the ASME Code Section XI, Paragraph IWB-364 1.1 of Reference 6.
6.0
References:
- 1) "Entergy Nuclear Southwest Design Input and Descriptions for the ANO Unit 1 CRDM Nozzle Flaw Evaluation", Task # 5. Engineering Programs, Central Engineering; Entergy Operations Inc., March 2001
- 2) Intentionally Left Blank
- 3) "Entergy Nuclear Southwest Design Input and Descriptions for the ANO unit I CRDM Nozzle Flaw Evaluation", Task # 9; Engineering Programs, Central Engineering; Entergy Operations Inc., March 2001
- 4) "Entergy Nuclear Southwest Design Input and Descriptions for the ANO unit I CRDM Nozzle Flaw Evaluation", Task # 12. Engineering Programs, Central Engineering; Entergy Operations Inc., March 2001
- 5) "Entergy Nuclear Southwest Design Input and Descriptions for the ANO unit I CRDM Nozzle Flaw Evaluation", Task # 7; Engineering Programs, Central Engineering; Entergy Operations Inc., March 2001
- 6) ASME Boiler and Pressure Vessel Code, Section XI, 1992 Edition
- 7) "Guide on Methods for Assessing the Acceptability of Flaws in Metallic Structures", Amendment No. 1; Annex P; British Standard BS-7910:1999; British Standards Institute; October 2000
- 8) "Steam Generator Tube Integrity, Volumes I and 2", EPRI NP-6865L; Electric Power Research Institute; Palo Alto, CA; June 1991
- 9) ASME Boiler and Pressure Vessel Code, Section II Part D, "Properties", 1992 Edition
- 10) "Computational Method to Perform the Flaw Evaluation Procedure as Specified in the ASME Code, Section XI, Appendix A", EPRI NP-l1181, Part 1; Electric Power Research Institute; Palo Alto, CA; September 1979
- 11) "Advanced Fracture Mechanics", Melvin F. Kanninen and Carl H. Popelar; Oxford University Press; New York, NY; 1985
- 12) "Cold Leg Integrity Evaluation", M. E. Mayfield et al; NUREG/CR- 1319; February 1980}}