Letter Sequence Request |
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EPID:L-2019-LLR-0066, CFR 50.55a Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1 in Accordance with 10 CFR 50.55a(z)(2) (Approved, Closed) |
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MONTHYEARW3F1-2019-0052, CFR 50.55a Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1 in Accordance with 10 CFR 50.55a(z)(2)2019-07-18018 July 2019 CFR 50.55a Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1 in Accordance with 10 CFR 50.55a(z)(2) Project stage: Request W3F1-2019-0054, Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-12019-07-22022 July 2019 Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1 Project stage: Request ML20002A0202020-01-13013 January 2020 Approval of Relief Request WF3-RR-19-2, Relief from the Requirements of Asme Code Section XI Regarding Alternate Repair of Degraded Drain Line of Chemical and Volume Control System Project stage: Approval ML20022A2582020-01-28028 January 2020 Re-Issuance of Approval of Relief Request WF3-RR-19 2, Relief from the Requirements of Asme Code Section XI Regarding Alternate Repair of Degraded Drain Line of Chemical and Volume Control Project stage: Approval 2019-07-22
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Category:Letter type:W
MONTHYEARW3F1-2023-0056, Owner'S Activity Report Form for Inservice Inspection Performed During Operating Cycle 24 / Refuel 242023-12-19019 December 2023 Owner'S Activity Report Form for Inservice Inspection Performed During Operating Cycle 24 / Refuel 24 W3F1-2023-0055, Reply to a Notice of Violation2023-12-14014 December 2023 Reply to a Notice of Violation W3F1-2023-0052, Core Operating Limits Report (COLR) - Cycle 26, Revision O2023-11-0707 November 2023 Core Operating Limits Report (COLR) - Cycle 26, Revision O W3F1-2023-0049, Revise Technical Specification 3/4.3.2 to Remove Exemption from Testing Certain Relays at Power to Support Elimination of Potential Single Point Vulnerability - Withdrawal2023-09-28028 September 2023 Revise Technical Specification 3/4.3.2 to Remove Exemption from Testing Certain Relays at Power to Support Elimination of Potential Single Point Vulnerability - Withdrawal W3F1-2023-0048, Special Report SR 2023-004-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days2023-09-25025 September 2023 Special Report SR 2023-004-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days W3F1-2023-0035, Application for Technical Specification Change to Revise Surveillance Requirements Included in the Surveillance Frequency Control Program2023-07-26026 July 2023 Application for Technical Specification Change to Revise Surveillance Requirements Included in the Surveillance Frequency Control Program W3F1-2023-0036, Special Report SR-2023-003-01 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days2023-05-0404 May 2023 Special Report SR-2023-003-01 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days W3F1-2023-0032, Annual Radioactive Effluent Release Report (ARERR) 20222023-04-27027 April 2023 Annual Radioactive Effluent Release Report (ARERR) 2022 W3F1-2023-0033, Submittal of Annual Radiological Environmental Operating Report - 20222023-04-27027 April 2023 Submittal of Annual Radiological Environmental Operating Report - 2022 W3F1-2023-0025, Annual Report of Individual Monitoring of Radiation Exposure for 2022 Per 10 CFR 20.22062023-04-11011 April 2023 Annual Report of Individual Monitoring of Radiation Exposure for 2022 Per 10 CFR 20.2206 W3F1-2023-0018, Updated Final Supplemental Response to NRC Generic Letter 2004-022023-03-30030 March 2023 Updated Final Supplemental Response to NRC Generic Letter 2004-02 W3F1-2023-0022, Registration of Dry Fuel Storage Cask Use2023-03-22022 March 2023 Registration of Dry Fuel Storage Cask Use W3F1-2023-0021, Submittal of Special Report SR 2023-003-00 Radiation Monitor Inoperable Greater than 7 Days2023-03-17017 March 2023 Submittal of Special Report SR 2023-003-00 Radiation Monitor Inoperable Greater than 7 Days W3F1-2023-0016, Registration of Dry Fuel Storage Cask Use2023-03-0303 March 2023 Registration of Dry Fuel Storage Cask Use W3F1-2023-0014, Reply to a Notice of Violation; EA-22-1192023-02-20020 February 2023 Reply to a Notice of Violation; EA-22-119 W3F1-2023-0013, Notification of Readiness for Supplemental Inspection2023-02-15015 February 2023 Notification of Readiness for Supplemental Inspection W3F1-2023-0007, Registration of Dry Fuel Storage Cask Use2023-02-0606 February 2023 Registration of Dry Fuel Storage Cask Use W3F1-2023-0010, Special Report SR 2023-002-00, Radiation Monitor Inoperable Greater than 7 Days2023-01-25025 January 2023 Special Report SR 2023-002-00, Radiation Monitor Inoperable Greater than 7 Days W3F1-2023-0002, SR 2023-001-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 30 Days2023-01-0505 January 2023 SR 2023-001-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 30 Days W3F1-2022-0067, Commitment Change Notification for Generic Safety Issue 191 and Generic Letter 2004-022022-12-20020 December 2022 Commitment Change Notification for Generic Safety Issue 191 and Generic Letter 2004-02 W3F1-2022-0054, Revise Technical Specification 3/4.3.2 to Remove Exemption from Testing Certain Relays at Power to Support Elimination of Potential Single Point Vulnerability2022-11-0101 November 2022 Revise Technical Specification 3/4.3.2 to Remove Exemption from Testing Certain Relays at Power to Support Elimination of Potential Single Point Vulnerability W3F1-2022-0063, Submittal of Emergency Preparedness Documents. Includes EP-001-001, Revision 372022-10-27027 October 2022 Submittal of Emergency Preparedness Documents. Includes EP-001-001, Revision 37 W3F1-2022-0059, Response to Clarification Questions Concerning Supplement to License Amendment Request to Adopt TSTF-5052022-10-13013 October 2022 Response to Clarification Questions Concerning Supplement to License Amendment Request to Adopt TSTF-505 W3F1-2022-0058, Reply to a Notice of Violation; EA-22-0332022-10-12012 October 2022 Reply to a Notice of Violation; EA-22-033 W3F1-2022-0049, Response to Request for Additional Information Regarding License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-5052022-08-19019 August 2022 Response to Request for Additional Information Regarding License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-505 W3F1-2022-0037, Submittal of Owner'S Activity Report Form for Inservice Inspection Performed During Operating Cycle 24 / Refuel 242022-08-0808 August 2022 Submittal of Owner'S Activity Report Form for Inservice Inspection Performed During Operating Cycle 24 / Refuel 24 W3F1-2022-0044, SR-2022-004-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days2022-07-0606 July 2022 SR-2022-004-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days W3F1-2022-0042, SR-22-003-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days2022-06-27027 June 2022 SR-22-003-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days W3F1-2022-0015, Response to Request for Additional Information to License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - Ritstf.2022-05-16016 May 2022 Response to Request for Additional Information to License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - Ritstf. W3F1-2022-0026, Report of Facility Changes, Tests, and Experiments and Commitment Changes for Two Year Period Ending April 28, 20222022-04-28028 April 2022 Report of Facility Changes, Tests, and Experiments and Commitment Changes for Two Year Period Ending April 28, 2022 W3F1-2022-0028, Annual Radiological Environmental Operating Report - 20212022-04-26026 April 2022 Annual Radiological Environmental Operating Report - 2021 W3F1-2022-0029, Annual Report of Individual Monitoring of Radiation Exposure for 2021 Per 10 CFR 20.22062022-04-26026 April 2022 Annual Report of Individual Monitoring of Radiation Exposure for 2021 Per 10 CFR 20.2206 W3F1-2022-0009, Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors2022-04-25025 April 2022 Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors W3F1-2022-0031, Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule to Reflect Location of Standby Capsules 3/W-104 and 6/W-2842022-04-25025 April 2022 Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule to Reflect Location of Standby Capsules 3/W-104 and 6/W-284 W3F1-2022-0020, Review of Preliminary Accident Sequence Precursor Report2022-04-11011 April 2022 Review of Preliminary Accident Sequence Precursor Report W3F1-2022-0017, Annual Report on Westinghouse Electric Company LLC Combustion Engineering Emergency Core Cooling System Performance Evaluation Models for Calendar Year 20212022-04-0707 April 2022 Annual Report on Westinghouse Electric Company LLC Combustion Engineering Emergency Core Cooling System Performance Evaluation Models for Calendar Year 2021 W3F1-2022-0019, WAT-2022-02 Post Exam Analysis2022-03-0909 March 2022 WAT-2022-02 Post Exam Analysis W3F1-2022-0011, Special Report SR-22-002-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days2022-02-0808 February 2022 Special Report SR-22-002-00 for Waterford Steam Electric Station, Unit 3, Radiation Monitor Inoperable Greater than 7 Days W3F1-2022-0008, SR-22-001-00, Waterford Steam Electric Station, Unit 3, Radiation Monitors Inoperable Greater than 7 Days2022-02-0101 February 2022 SR-22-001-00, Waterford Steam Electric Station, Unit 3, Radiation Monitors Inoperable Greater than 7 Days W3F1-2021-0074, Commitment Change Notification for Generic Safety Issue - 191 and Generic Letter 2004-022021-12-16016 December 2021 Commitment Change Notification for Generic Safety Issue - 191 and Generic Letter 2004-02 W3F1-2021-0064, Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule to Support Relocation of Capsules 104 and 2842021-11-30030 November 2021 Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule to Support Relocation of Capsules 104 and 284 W3F1-2021-0061, Supplement to License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual2021-10-14014 October 2021 Supplement to License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual W3F1-2021-0063, Request for One-Time Exemption from 10 CFR 50, Appendix E Biennial Emergency Preparedness Evaluated Exercise Requirements Due to Severe Storm Recovery2021-10-12012 October 2021 Request for One-Time Exemption from 10 CFR 50, Appendix E Biennial Emergency Preparedness Evaluated Exercise Requirements Due to Severe Storm Recovery W3F1-2021-0050, Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt2021-10-0101 October 2021 Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt W3F1-2021-0060, Response to Request for Additional Information Regarding License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual2021-09-30030 September 2021 Response to Request for Additional Information Regarding License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual W3F1-2021-0039, Application for Technical Specification Change to Revise Pressure/Temperature and Low Temperature Overpressure Protection for 55 Effective Full Power Years2021-08-25025 August 2021 Application for Technical Specification Change to Revise Pressure/Temperature and Low Temperature Overpressure Protection for 55 Effective Full Power Years W3F1-2021-0055, Supplement to License Amendment Request to Relocate Boration Systems Technical Specifications to the Technical Requirements Manual2021-08-20020 August 2021 Supplement to License Amendment Request to Relocate Boration Systems Technical Specifications to the Technical Requirements Manual W3F1-2021-0057, (Waterford 3) - Emergency Plan Revision 0522021-08-18018 August 2021 (Waterford 3) - Emergency Plan Revision 052 W3F1-2021-0054, License Amendment Request to Implement a Digital Upgrade to the Core Protection Calculator (CPC) System and Control Element Assembly Calculator (Ceac) System, Dated July 23, 20202021-07-29029 July 2021 License Amendment Request to Implement a Digital Upgrade to the Core Protection Calculator (CPC) System and Control Element Assembly Calculator (Ceac) System, Dated July 23, 2020 W3F1-2021-0051, Revised Licensing Technical Report for the Common Q Core Protection Calculator System - License Amendment Request to Implement a Digital Upgrade to the Core Protection Calculator System and Control Element Assembly Calculator2021-07-19019 July 2021 Revised Licensing Technical Report for the Common Q Core Protection Calculator System - License Amendment Request to Implement a Digital Upgrade to the Core Protection Calculator System and Control Element Assembly Calculator 2023-09-28
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARML24012A1962024-01-12012 January 2024 Response to 2nd Round Request for Additional Information Concerning Relief Request Number EN-RR-22-001 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and ML23111A2132023-04-21021 April 2023 Responses to RAI Concerning Relief Request Number EN-RR-22-001 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities W3F1-2022-0059, Response to Clarification Questions Concerning Supplement to License Amendment Request to Adopt TSTF-5052022-10-13013 October 2022 Response to Clarification Questions Concerning Supplement to License Amendment Request to Adopt TSTF-505 W3F1-2022-0049, Response to Request for Additional Information Regarding License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-5052022-08-19019 August 2022 Response to Request for Additional Information Regarding License Amendment Requests to Adopt 10 CFR 50.69 and TSTF-505 W3F1-2022-0015, Response to Request for Additional Information to License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - Ritstf.2022-05-16016 May 2022 Response to Request for Additional Information to License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - Ritstf. W3F1-2022-0009, Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors2022-04-25025 April 2022 Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors W3F1-2021-0050, Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt2021-10-0101 October 2021 Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt W3F1-2021-0060, Response to Request for Additional Information Regarding License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual2021-09-30030 September 2021 Response to Request for Additional Information Regarding License Amendment Request to Relocate Chemical Detection Systems Technical Specifications to Technical Requirements Manual W3F1-2021-0041, Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Digital Upgrade to the Core Protection Calculator and Control Element Assembly Calculator System2021-06-0202 June 2021 Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Digital Upgrade to the Core Protection Calculator and Control Element Assembly Calculator System 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 W3F1-2020-0062, Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 62020-12-15015 December 2020 Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 W3F1-2020-0043, Response to NRC Request for Additional Information Regarding License Amendment Request to Relocate Boration Systems Technical Specifications to the Technical Requirements Manual2020-07-0606 July 2020 Response to NRC Request for Additional Information Regarding License Amendment Request to Relocate Boration Systems Technical Specifications to the Technical Requirements Manual W3F1-2020-0033, (Waterford 3) - Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3/4.8.1, A.C. Sources Operating2020-05-29029 May 2020 (Waterford 3) - Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3/4.8.1, A.C. Sources Operating 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 W3F1-2019-0056, Revision to Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-12019-07-31031 July 2019 Revision to Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1 W3F1-2019-0054, Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-12019-07-22022 July 2019 Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Number WF3-RR-19-2, Proposed Alternative to Code Case N-666-1 W3F1-2019-0040, Response to U. S. Nuclear Regulatory Commission Second Round Request for Additional Information Regarding License Amendment Request for Use of the Tranflow Code for Determining Pressure Drops Across the Steam Generator Secondary Side2019-07-11011 July 2019 Response to U. S. Nuclear Regulatory Commission Second Round Request for Additional Information Regarding License Amendment Request for Use of the Tranflow Code for Determining Pressure Drops Across the Steam Generator Secondary Side Intern W3F1-2019-0039, (Waterford 3) - Response to Request for Additional Information Regarding Relief Request W3-ISI-0322019-07-0808 July 2019 (Waterford 3) - Response to Request for Additional Information Regarding Relief Request W3-ISI-032 W3F1-2019-0005, Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding Proposed Change to Technical Specification 3/4.7.4 for Ultimate Heat Sink Design Basis Update2019-02-15015 February 2019 Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding Proposed Change to Technical Specification 3/4.7.4 for Ultimate Heat Sink Design Basis Update W3F1-2019-0011, Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Request WF3-RR-19-1 for Application of Dissimilar Metal Weld Full Structural Weld Overlay2019-02-0404 February 2019 Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding Relief Request WF3-RR-19-1 for Application of Dissimilar Metal Weld Full Structural Weld Overlay W3F1-2019-0002, (Waterford 3) - Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request Regarding Use of the Tranflow Code for Determining the Pressure Drops Across2019-01-19019 January 2019 (Waterford 3) - Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request Regarding Use of the Tranflow Code for Determining the Pressure Drops Across W3F1-2018-0067, Response to NRC Request for Alternative to ASME Code Case N-770-2, Successive Examinations, Relief Request W3-ISI-0312018-11-19019 November 2018 Response to NRC Request for Alternative to ASME Code Case N-770-2, Successive Examinations, Relief Request W3-ISI-031 W3F1-2018-0059, (Waterford 3) - Response to NRC Request for Additional Information Regarding License Amendment Request to Update the Results for the Inadvertent Loading of a Fuel Assembly Into the Improper Position ...2018-10-18018 October 2018 (Waterford 3) - Response to NRC Request for Additional Information Regarding License Amendment Request to Update the Results for the Inadvertent Loading of a Fuel Assembly Into the Improper Position ... CNRO-2018-00021, Response to Request for Supplemental Information Regarding Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in the Spent Fuel Pools for Indian Point Energy Center, Unit 3 and Waterford 3 Steam Electric Station2018-05-30030 May 2018 Response to Request for Supplemental Information Regarding Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in the Spent Fuel Pools for Indian Point Energy Center, Unit 3 and Waterford 3 Steam Electric Station W3F1-2018-0020, Responses to Request for Additional Information Set 18 Regarding the License Renewal Application2018-04-23023 April 2018 Responses to Request for Additional Information Set 18 Regarding the License Renewal Application ML18101A0962018-04-10010 April 2018 Notification of an NRC Triennial Fire Protection Baseline Inspection (NRC Inspection Report 05000382/2018007) and Request for Information W3F1-2017-0039, Responses to Request for Additional Information Set 17 Regarding the License Renewal Application2017-05-12012 May 2017 Responses to Request for Additional Information Set 17 Regarding the License Renewal Application W3F1-2017-0027, Responses to Request for Additional Information Set 16 Regarding the License Renewal Application2017-05-0202 May 2017 Responses to Request for Additional Information Set 16 Regarding the License Renewal Application ML17114A4322017-04-21021 April 2017 Responses to Request for Additional Information for the Environmental Review (SAMA Round 2) Regarding the License Renewal Application W3F1-2017-0026, Responses to Request for Additional Information Set 15 Regarding the License Renewal Application2017-04-11011 April 2017 Responses to Request for Additional Information Set 15 Regarding the License Renewal Application W3F1-2017-0023, Responses to Request for Additional Information Set 14 Regarding the License Renewal Application2017-03-30030 March 2017 Responses to Request for Additional Information Set 14 Regarding the License Renewal Application W3F1-2017-0015, Responses to Request for Additional Information Set 13 Regarding the License Renewal Application2017-03-16016 March 2017 Responses to Request for Additional Information Set 13 Regarding the License Renewal Application W3F1-2017-0018, Supplement to Response to RAI on License Amendment Request to Revise Technical Specification 3/4.3.2 to Relocate Surveillance Frequency Requirements for Engineered Safety Features Actuation System Subgroup Relays to Surveillance.2017-02-27027 February 2017 Supplement to Response to RAI on License Amendment Request to Revise Technical Specification 3/4.3.2 to Relocate Surveillance Frequency Requirements for Engineered Safety Features Actuation System Subgroup Relays to Surveillance. ML17054D2392017-02-23023 February 2017 Responses to Request for Additional Information Set 12 Regarding the License Renewal Application ML17024A2392017-02-0707 February 2017 Alternatives RAI-AL-3 Lanning 2014_Correspondence Lanning-Entergy to Buckley-Entergy ML17037D0172017-02-0707 February 2017 Aquatic Resources (AR) RAI-7 Lpl 1979_WF3 Demonstration Under Section 316(b) ML17038A2642017-02-0707 February 2017 Surface Water Resources (Sw)Rai SW-11 Usace NNOD-SP (Mississippi River) 796 Permit ML17034A3232017-02-0707 February 2017 Aquatic Resources (AR)- RAI AR-7 Ensr 2005_W1&2 PIC ML17038A2772017-02-0707 February 2017 Surface Water Resources (SW) RAI SW-11 Ldwf Letter of No Objection ML17030A1872017-02-0707 February 2017 Aquatic RESOURCES-RAI AR-6 LPDES Permit Modification_2008 ML17023A2682017-02-0707 February 2017 Air Quality RAI AQ-5 WF3 2004a_WF3 Air Permit 2520-00091-00 ML17037D0362017-02-0707 February 2017 Cumultive Impacts (Cu) RAI CU-1 Reasonably Foreseeable Future Projects within WF3 Area ML17037D0552017-02-0707 February 2017 Groundwater Resources (Gw) RAI GW-1 WF3 Groundwater Monitoring Program_Five-Year Review ML17038A4362017-02-0707 February 2017 Responses to Request for Additional Information for the Environmental Review ML17024A1702017-02-0707 February 2017 Alternatives RAI AL-1 Power Replacement Suitable Sites ML17018A1862017-02-0707 February 2017 Air Quality- RAI AQ-2 Temperature Analysis Mississippi River ML17037D2822017-02-0707 February 2017 Surface Water Resources (Sw)Rai SW-1 WF3 Surface Water Withdrawals_Discharges_2011-2015 ML17024A1932017-02-0707 February 2017 Alternatives RAI AL-3 - Enercon 2015b_Calc Supporting Section 07_01 Alternatives ML17024A2472017-02-0707 February 2017 Aquatic Resources RAI AR-4 -Entergy 2005_WF3 316(b) Proposal for Information Collection ML17038A2802017-02-0707 February 2017 Surface Water Resources (SW) RAI SW-11 U.S. Coast Guard Permit CG-2554.pdf 2024-01-12
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CALCULATION PACKAGE File No.:
1900769.30 2 Project No.:
1 900769 Quality Program Type: Nuclear Commercial PROJECT NAME:
Waterford Code Case N
-513 Evaluation of Leaking Class 2 Line CONTRACT NO.:
10585075 CLIENT: Entergy Nuclear PLANT: Waterford Steam Electric Station, Unit 3 CALCULATION TITLE:
Evaluation of Weld Overlay Repair of Socket Weld Region Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) & Checker(s)
Signatures & Date 0 1 - 8 Initial Issue Eric Houston 7/18/19 Prepare r: Andrew Collins 7/18/19 Checker: .Stephen Parker 7/18/19 File No.:
1900769.30 2 Revision:
0 Page 2 of 8 F0306-01R 4
Table of Contents 1.0 OBJECTIVE ..............................................................................................................
3 2.0 METHODOLOGY
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3 2.1 Criteria for Hoop Stress
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3 2.2 Criteria for Axial Stress
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4 3.0 DESIGN INPUTS
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5 4.0 ASSUMPTIONS
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6 5.0 CALCULATIONS
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6 6.0 RESULTS OF ANALYSIS
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7
7.0 CONCLUSION
S
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7
8.0 REFERENCES
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8
List of Tables
Table 1: Applied Moment Loads
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6 Table 2: Minimum Thickness Results
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7
File No.:
1900769.30 2 Revision:
0 Page 3 of 8 F0306-01R 4
1.0 OBJECTIVE A leaking flaw was recently discovered at the Waterford Steam Electric Station, Unit 3 (Waterford) in the Chemical and Volume Control (CVC) system. The Class 2 leaking pipe is upstream of valve CVCMVAAA186 (CVC-186). The objective of this calculation is to determine the minimum required wall thickness for the repaired section of piping with a weld overlay repair.
2.0METHODOLOGY The structural suitability of a weld overlay repair is accomplished by verifying that the thickness of the weld buildup meets the required minimum thickness for the operational condition.
The methodology of this calculation uses the equations for hoop stress and axial stress along with allowable stresses from Section III, 1971 Edition with Addenda through Winter 1972 Code of Construction [
1].
The hoop stress limit is defined by the Code of Construction [
1 , NC-3641.1 Equation 3] and calculates the minimum required wall thickness due to internal pressure, t
- m. The axial stress limits are defined as a series of stress limits based on pressure and piping loads:
Equation 8, Longitudinal Stresses due to Sustained Loads [
1 , NC-3652.1] Equation 9, Longitudinal Stresses due to Occasional Loads [
1 , NC-3652.2] Equation 10, Longitudinal Thermal Expansion Stresses [
1 , NC-3652.3(a)]
Equation 11, Longitudinal Thermal Expansion and Sustained Loads Stress [
1 , NC-3652.3(b)]
The smallest wall thickness that satisfies both the hoop and axial stress limits is defined as the minimum
wall thickness, tmin. Note that only Equation 10 or Equation 11 is required to be met, not both [
1 , NC-3652.3]. Therefore, only Equation 10 is evaluated herein.
For this evaluation, the tmin value will be conservatively calculated using the dimensions of the weld buildup over the socket welded connections and ignoring the structural benefit of the unflawed existing base pipe material.
2.1Criteria for Hoop Stress The minimum thickness required based on hoop stress, Equation 3 [1 , NC-3641.1], assures against gross structural failure due to primary membrane pressure loading. Equation 3 is written as a design thickness calculation based on the maximum allowable stress. The minimum thickness required for design pressure, t m, is defined by Equation 3 as: P = Internal design pressure, psi D o = Outside pipe diameter, in S = Maximum allowable stress at design temperature, psi E = Longitudinal weld joint efficiency factor (1.0 for seamless pipe) y = Pressure coefficient
= 0.4 [1 , NB-3641.1] A = Additional thickness, in
The additional thickness value, A, is taken as zero.
File No.:
1900769.30 2 Revision:
0 Page 4 of 8 F0306-01R 4
2.2Criteria for Axial Stress Equations 8, 9, and 10
[1 , NC-3652] are intended to show that the calculated axial stresses in the piping component, due to pressure and piping loads, meet the Code of Construction stress limits. An iterated uniform wall thickness is used to calculate the piping stresses in these equations, which are compared to an allowable stress for design purposes.
The thickness that results in a stress equal to th e allowable stress is taken as the minimum thickness for that condition.
The Longitudinal Stresses Due to Sustained Loads , S SL, must satisfy the following requirement [
1 , NC-3652.1 , Equation 8
]: P = Internal design pressure, psi D o = Outside pipe diameter, in t n = Nominal wall thickness, in i = Stress intensification factor (Note: 0.75i may not be less than 1.0)
(see [1 , NC-3673.2(b)])
M A = Resultant moment due to sustained loads, in
-lb (see [1, NC-3654]) Z = Section modulus, in 3 (see [1, NC-3654]) S h = Allowable stress at design temperature (equivalent to S for this evaluation), psi
The Longitudinal Stresses Due to Occasional Loads , S OL, must satisfy the following requiremen t 1 , NC-3652.2, Equation 9]: Pmax = Peak pressure (taken as operating pressure), psi D o = Outside pipe diameter, in t n = Nominal wall thickness, in i = Stress intensification factor (Note: 0.75i may not be less than 1.0)
(see [1 , NC-3673.2(b)])
M A = Resultant moment due to sustained loads, in
-lb (see [1, NC-3654]) M R = Resultant moment due to occasional loads, in
-lb (see [1, NC-3652.4]) Z = Section modulus, in 3 (see [1, NC-3654]) S h = Allowable stress at design temperature (equivalent to S for this evaluation), psi
The Thermal Expansion Stress es , S E, must satisfy the following requirement [
1 , NC-3652.3(a), Equation 10]: i = Stress intensification factor M C = Range of resultant moment due to thermal expansion, in
-lb (see [1, NC-3652.2]) Z = Section modulus, in 3 (see [1, NC-3654]) S A = Allowable stress range for expansion stresses, psi
The allowable stress range, S A, is defined as S A=f (1.25S c+0.25S h) [1 , NC-3611.1(b)(3)], where f is defined as the stress range reduction factor for full temperature thermal cycles. S c is the basic material allowable File No.:
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stress at minimum (cold) temperature. For this evaluation, the reduction factor is assumed to be equal to 1.0 since the number of thermal expansion cycles is expected to be below 7,000 (see Section 4.0 , Assumption No.
2). Therefore, S A is equal to 1.5S h as S c and S h are equivalent for this evaluation.
3.0DESIGN INPUTS The following design inputs were provided by Entergy personnel to be used in the analysis:
1.Pipe Material =
SA-376, TP-304 [2] 2.Pipe Nominal Size = 1-inch Schedule 80 [
2] 3.Maximum Operating Temperature = 120
°F [3] 4.Design Temperature = 250
°F [3] 5.Maximum Operating Pressure = 85 psig [
3] 6.System Code of Construction = ASME Section III, 1971 Edition with Addenda through Winter 1972
[3] 7.Code Allowable Stress, interpolated at 120°F, = 18.6 ksi [
1, Table I-7.2] 8.Material Yield Strength = 30 ksi [1, Table I-7.2] 9.Material Ultimate Strength = 75 ksi [
1, Table I-7.2] The pipe outside diameter (1.315 inches) and nominal thickness (0.179 inch) are obtained from readily available industry information based on the pipe size and schedule.
The OD of the socket welded elbow and coupling are assumed to be those of 3000 lb fittings from
[5], which have a maximum fitting OD of 1.8125 inches. Piping loads are taken from the design basis stress report [
4]. The location of interest is at Node 403, but the bounding load s used in the evaluation are taken from Node 402, which is the branch of the socket welded connection to the 4
-inch run piping. The methodology requires moment loads as input, which are obtained from the Reference [
4] nodal outputs for each load case. Three load cases are utilized: thermal expansion (TH), sustained (DW), and operating basis earthquake (OBE). OBE is the only seismic loading evaluated in the stress report and is, therefore, the only seismic loading evaluated herein.
The component moments at Node 402 are taken from each load case in the Reference [
4] output. The square root sum of squares (SRSS) moments are calculated for each load.
Constrained thermal expansion stress in a simple system is roughly linear over small ranges of changes in temperature (T), with slight non
-linearities introduced due to temperature depended material properties. Complex system s, such as piping systems, are not strictly linear due to geometric effects and the potential influence of mixed metals. The design basis stress report only evaluates thermal expansion loading for the design temperature of 250°F. If the reference temperature for thermal expansion stress is taken as 70°F (see Assumption 1), the evaluated T is 180°F. The maximum operating temperature is only 120°F, which represents a T of 50°F. Rather than use the thermal expansion loading for the full 180
°F T, the thermal expansion loads are scaled by a factor of 0.5. This represents an evaluated T of approximately 90°F , which is still considered conservative when the actual T is nearly half that amount.
To conservatively account for the stress intensification due to the weld overlay repair, a stress intensification factor (SIF) of 1.3 is used in the analysis. This SIF is consistent with Figure NC
-3672.9(a)-1 from the Code of Construction for socket welded connections [1]. Application of this SIF to the analysis is conservative as the stress intensification created by the original socket welded joint will be reduced with the addition of the repair weld metal. The geometric configuration of the weld overlay will result in a design that File No.:
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is similar to a butt welded connection, which have an SIF of 1.0 per Figure NC
-3672.9(a)-1 from the Code of Construction [1]. The loads for the analysis, derived from Reference [
4], are shown in Table 1. Table 1: Applied Moment Loads From Stress Report (ft
-lbs) Resultant T Scale Factor Applied Evaluated Loads X Y Z (in-lbs) (in-lbs) (in-lbs) DW 19 1 8 248 --- 248 OBE 21 13 16 353 --- 353 TH 53 170 42 2195 1098 1098 4.0ASSUMPTIONS The following assumptions are used in the evaluation.
1.It is assumed that 70°F is used as the reference temperature (i.e., the zero-stress state) in the design basis thermal expansion stress analysis [
4]. The stress report then evaluates the change in temperature from 70°F to 250°F. Use of 70°F is typical for such an evaluation. Use of a different reference temperature would result in a change in the resulting stress (higher stress for a lower reference temperature, lower stress for a higher reference temperature). However, use of a significantly different reference temperature does not have a meaningful impact on the results of the analysis, and there is no basis for evaluating from a different reference temperature. 2.It is assumed that the full thermal cycles for the piping system total less than 7,000 cycles. Given the operation of the system, this assumption is appropriate for determining the stress range reduction factor (f) for this analysis. 3.The evaluation assumes no structural benefits from the existing base pipe material. This is reasonable and conservative for analyzing the acceptance of the weld overlay buildup.
5.0CALCULATIONS For longitudinal stresses, the thickness , t n, is iterated for Equations 8, 9, and 10 until the calculated stress is equal to the allowable stress, as defined for each equation.
This calculation provides a resulting m inimum thickness necessary that meets the Code of Construction requirements based on the dimensions of the added weld overlay material. T he section modulus of the weld overlay cross section is also iterated as function of the evaluated thickness.
The minimum thickness necessary for hoop stresses is calculated directly from Equation
- 3.
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6.0RESULTS OF ANALYSIS The resulting tmin for each loading condition is shown in Table 2. Table 2: Minimum Thickness Results Loading Condition Minimum Required Thickness (inch) Equation 3
- Hoop Stress 0.004 Equation 8 - Sustained Load
- Deadweight 0.00 8 Equation 9
- Occasional Load - OBE 0.01 3 Equation 10
- Thermal Expansion 0.0 2 1 The Code of Construction minimum required wall thickness for these conditions is taken as the maximum of the resulting thicknesses tabulated in Table 2. Therefore, the minimum required wall thickness for the weld overlay design is 0.021 inch and is limited by thermal expansion.
7.0CONCLUSION
S This evaluation calculates the minimum required thicknesses of the Waterford Steam Electric Station, Unit 3 CVC system Class 2 piping region upstream of valve CVCMVAAA186 (CVC-1 86) in support of a Code Case N
-666-1 weld overlay repair. The Code of Construction minimum required wall thickness for the weld overlay repair is 0.021-inch. This minimum thickness is based on the dimensions of the socket welded fittings and the resulting weld overlay and conservatively ignores the structural benefit of the existing unflawed base pipe material.
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8.0REFERENCES
1.ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition with Addenda through Winter 1972. 2.Waterford Drawing No. 4305-3913, SI File No. 1900769.205.
3.Email from T. House (Entergy) to E. Houston (SI), Subject "RE: SI Contact Information," July 2, 2019, SI File No. 1900769.208.
4.Waterford Stress Report No. SA
-2869-2, Revision 2, "Stress Analy sis of CH-Piping per SMP
-1743 'A s-Buil t'," SI File No. 1900769.201.
5.Ladish General Catalog No. 55, Forged and Seamless Welding Pipe Fittings,1954, SI File No.
1900769.209.