Letter Sequence Request |
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EPID:L-2018-LLR-0050, Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability (Approved, Closed) |
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Category:Inservice/Preservice Inspection and Test Report
MONTHYEARNL-18-063, Submission of IP2 Steam Generator Examination Program Results for the 2018 (2R23) Refueling Outage in Accordance with Technical Specification 5.6.72018-08-13013 August 2018 Submission of IP2 Steam Generator Examination Program Results for the 2018 (2R23) Refueling Outage in Accordance with Technical Specification 5.6.7 NL-18-049, 2018 Form OAR-1 Owners Activity Report for Inservice Inspection and Repairs and Replacements2018-07-10010 July 2018 2018 Form OAR-1 Owners Activity Report for Inservice Inspection and Repairs and Replacements NL-18-021, Indian Point, Unit 2 - Submittal of Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability2018-04-0606 April 2018 Indian Point, Unit 2 - Submittal of Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability NL-18-021, Submittal of Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability2018-04-0606 April 2018 Submittal of Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability ML18103A0302018-04-0606 April 2018 Attachment 1 to NL-18-021, LTR-SDA-18-035-NP, Revision 0, Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability NL-18-019, Relief Request Number IP2-ISI-RR-06 - Proposed Alternative to Use Reactor Vessel Head Penetration Flaw Weld Repair Method2018-04-0404 April 2018 Relief Request Number IP2-ISI-RR-06 - Proposed Alternative to Use Reactor Vessel Head Penetration Flaw Weld Repair Method NL-17-059, Request IP2-ISI-RR-22 for Relief from Examination of Non-Regenerative Heat Exchanger Base Support Welded Attachments for Fourth Ten-Year Inservice Inspection Interval Closeout2017-05-30030 May 2017 Request IP2-ISI-RR-22 for Relief from Examination of Non-Regenerative Heat Exchanger Base Support Welded Attachments for Fourth Ten-Year Inservice Inspection Interval Closeout NL-17-057, Request IP2-ISI-RR-20 for Relief from Examinations of Code Class 1 Component Welds with Less than Essentially 100% Examination Coverage for Fourth Ten-Year Inservice Inspection Interval Closeout2017-05-30030 May 2017 Request IP2-ISI-RR-20 for Relief from Examinations of Code Class 1 Component Welds with Less than Essentially 100% Examination Coverage for Fourth Ten-Year Inservice Inspection Interval Closeout NL-16-096, 2016 Form OAR-1 Owners Activity Report for Inservice Inspection and Repairs and Replacements2016-09-0909 September 2016 2016 Form OAR-1 Owners Activity Report for Inservice Inspection and Repairs and Replacements NL-16-097, Fifth Ten Year Interval Inservice Inspection Program Plan2016-09-0606 September 2016 Fifth Ten Year Interval Inservice Inspection Program Plan NL-16-091, Submittal of Fifth Ten Year Interval Inservice Inspection Program Plan2016-08-17017 August 2016 Submittal of Fifth Ten Year Interval Inservice Inspection Program Plan NL-16-092, Submittal of Fifth Ten Year Interval Lnservice Testing Program Plan2016-08-17017 August 2016 Submittal of Fifth Ten Year Interval Lnservice Testing Program Plan NL-16-032, Entergy Transmittal of Indian Point 2 ASME Section XI, Iwl Concrete Containment Inspection in Accordance with the Parties Approved Settlement of License Renewal Contention NYS-24 Indian Point Unit 22016-03-16016 March 2016 Entergy Transmittal of Indian Point 2 ASME Section XI, Iwl Concrete Containment Inspection in Accordance with the Parties Approved Settlement of License Renewal Contention NYS-24 Indian Point Unit 2 NL-15-073, Request for Relief Request IP2-ISI-RR-18 Maintaining ISI Related Activities on the 2001 Edition/2003A ASME Section XI Code for Fifth 10-Year Inservice Inspection (ISI) Interval2015-06-0101 June 2015 Request for Relief Request IP2-ISI-RR-18 Maintaining ISI Related Activities on the 2001 Edition/2003A ASME Section XI Code for Fifth 10-Year Inservice Inspection (ISI) Interval NL-14-060, 2014 Summary Reports for In-Service Inspection and Repairs or Replacements2014-05-0909 May 2014 2014 Summary Reports for In-Service Inspection and Repairs or Replacements NL-13-032, Technical Specification 5.6.8 - IP3 Steam Generator Tube Inspection Report - Spring 2013 Refueling Outage2013-08-15015 August 2013 Technical Specification 5.6.8 - IP3 Steam Generator Tube Inspection Report - Spring 2013 Refueling Outage NL-13-031, Submittal of 2013 Summary Reports for Inservice Inspection and Repairs or Replacements2013-06-18018 June 2013 Submittal of 2013 Summary Reports for Inservice Inspection and Repairs or Replacements NL-13-040, Relief Request IP2-ISI-RR-16: Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination2013-02-20020 February 2013 Relief Request IP2-ISI-RR-16: Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination NL-13-002, Proposed Technical Specification Bases Changes to Credit Four Fan Cooler Units in Containment Integrity Analysis2013-01-28028 January 2013 Proposed Technical Specification Bases Changes to Credit Four Fan Cooler Units in Containment Integrity Analysis NL-11-075, Submittal of Request for Relief to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination2011-06-29029 June 2011 Submittal of Request for Relief to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination NL-11-050, Submittal of 2011 Summary Reports for In-Service Inspection and Repairs or Replacements2011-06-20020 June 2011 Submittal of 2011 Summary Reports for In-Service Inspection and Repairs or Replacements ML1024400362010-08-24024 August 2010 Submittal of Steam Generator Examination Program Results 2010 Refueling Outage (2R19) NL-10-058, Summary Reports for In-Service Inspection and Repairs or Replacements2010-06-0909 June 2010 Summary Reports for In-Service Inspection and Repairs or Replacements NL-10-059, Relief Request IP2-ISI-RR-11 for Fourth Ten-Year Inservice Inspection Interval2010-06-0303 June 2010 Relief Request IP2-ISI-RR-11 for Fourth Ten-Year Inservice Inspection Interval NL-09-107, Relief Request 2-10 for Fourth Ten-Year Inservice Inspection Interval2009-08-0505 August 2009 Relief Request 2-10 for Fourth Ten-Year Inservice Inspection Interval NL-09-097, Submittal of Fourth Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan2009-07-21021 July 2009 Submittal of Fourth Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan NL-09-090, Relief Requests 08 and 09 for Fourth Ten-Year Inservice Inspection Interval2009-07-0101 July 2009 Relief Requests 08 and 09 for Fourth Ten-Year Inservice Inspection Interval NL-09-069, 2009 Summary Reports for In-Service Inspection and Repairs or Replacements2009-07-0101 July 2009 2009 Summary Reports for In-Service Inspection and Repairs or Replacements NL-09-087, Relief Request IP3-ISI-RR-01, IP3-ISI-RR-02, and IP3-ISI-RR-03 for Fourth Ten-Year Inservice Inspection Interval2009-06-24024 June 2009 Relief Request IP3-ISI-RR-01, IP3-ISI-RR-02, and IP3-ISI-RR-03 for Fourth Ten-Year Inservice Inspection Interval NL-09-037, Response to Request for Information Regarding Request for Relief 3-48 Supporting Refuel Outage 15 Inservice Inspection Program2009-03-23023 March 2009 Response to Request for Information Regarding Request for Relief 3-48 Supporting Refuel Outage 15 Inservice Inspection Program NL-09-003, Supplemental Response to Request for Additional Information on Request for Relief to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination2009-01-20020 January 2009 Supplemental Response to Request for Additional Information on Request for Relief to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination NL-08-110, Summary Reports for In-Service Inspection and Repairs for Replacements2008-07-14014 July 2008 Summary Reports for In-Service Inspection and Repairs for Replacements NL-08-053, 10 CFR 50.55a Request RR-CRV-75 - Relief from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third-Ten Year Inservice Inspection Interval Closeout2008-03-26026 March 2008 10 CFR 50.55a Request RR-CRV-75 - Relief from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third-Ten Year Inservice Inspection Interval Closeout NL-07-069, Inservice Inspection Third Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval2007-06-13013 June 2007 Inservice Inspection Third Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval ML0707100072007-03-19019 March 2007 Summary of the Staff'S Review of the 2006 Steam Generator Tube Inservice Inspection Reports for Refueling Outrage 17 NL-07-028, Inservice Testing Program Summary for 4th Interval, Revision 02007-02-28028 February 2007 Inservice Testing Program Summary for 4th Interval, Revision 0 NL-07-029, 4th Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan2007-02-28028 February 2007 4th Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan NL-05-088, Inservice Inspection (ISI) Second Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval2005-07-0606 July 2005 Inservice Inspection (ISI) Second Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval NL-05-0211, Inservice Testing Program Summary for the Interval July 1, 1994 Through April 6.2006. Revision 32005-02-22022 February 2005 Inservice Testing Program Summary for the Interval July 1, 1994 Through April 6.2006. Revision 3 ML0506304052005-02-22022 February 2005 Inservice Inspection (ISI) Third Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval JPN-04-010, Request to Use 1998 Edition, 2000 Addenda of American Society of Mechanical Engineers (ASME) Section Xl Code Requirements for Examination of Reactor Vessel Closure Studs2004-04-14014 April 2004 Request to Use 1998 Edition, 2000 Addenda of American Society of Mechanical Engineers (ASME) Section Xl Code Requirements for Examination of Reactor Vessel Closure Studs NL-03-188, Request for Approval for Alternative to Use Code Case N-613-1 for Reactor Vessel Nozzle to Vessel Weld Inspection2003-12-30030 December 2003 Request for Approval for Alternative to Use Code Case N-613-1 for Reactor Vessel Nozzle to Vessel Weld Inspection NL-03-078, Relief Request RR 63, Risk-Informed Inservice Inspection (RI-ISI) Program2003-05-12012 May 2003 Relief Request RR 63, Risk-Informed Inservice Inspection (RI-ISI) Program NL-03-055, Withdrawal of Relief Request RR3-29 for Inservice Inspection Program2003-03-27027 March 2003 Withdrawal of Relief Request RR3-29 for Inservice Inspection Program NL-03-026, Refueling Outage Inservice Inspection (ISI) Program Summary Report - Third Outage, Second Period, Third Interval2003-02-25025 February 2003 Refueling Outage Inservice Inspection (ISI) Program Summary Report - Third Outage, Second Period, Third Interval NL-04-006, St Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval2003-01-20020 January 2003 St Period Inspection Results and Repair/Replacement Activities for Third 10-Year Inservice Inspection Interval ML0209900922002-04-0303 April 2002 Revised Relief Request Nos. 3-12, 3-14, 3-16 & 3-17, Third 10-Year Inservice Inspection Interval Program Plan ML0205803912002-02-0505 February 2002 Relief Request RR 3-28, Risk-Informed Inservice Inspection (RI-ISI) Program 2018-08-13
[Table view] Category:Letter type:NL
MONTHYEARNL-21-034, Notification of Expected Date of Transfer of Ownership of Nuclear Units to Holtec Indian Point 2, LLC and Holtec Indian Point 3, LLC; and Notification of Receipt of All Required Regulatory Approvals2021-05-26026 May 2021 Notification of Expected Date of Transfer of Ownership of Nuclear Units to Holtec Indian Point 2, LLC and Holtec Indian Point 3, LLC; and Notification of Receipt of All Required Regulatory Approvals NL-21-039, Response to Request for Additional Information - License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary2021-05-20020 May 2021 Response to Request for Additional Information - License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary NL-21-033, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel2021-05-11011 May 2021 Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel NL-21-032, Termination of Emergency Response Data System Feed to the U.S. Nuclear Regulatory Commission at Indian Point Energy Center2021-05-11011 May 2021 Termination of Emergency Response Data System Feed to the U.S. Nuclear Regulatory Commission at Indian Point Energy Center NL-21-005, Cancellation of Commitments Related to Beyond-Design-Basis External Events Seismic and Flooding Actions2021-05-11011 May 2021 Cancellation of Commitments Related to Beyond-Design-Basis External Events Seismic and Flooding Actions NL-21-030, Submittal of 2020 Annual Radiological Environmental Operating Report2021-05-0606 May 2021 Submittal of 2020 Annual Radiological Environmental Operating Report NL-21-027, Registration of Spent Fuel Cask Use2021-04-20020 April 2021 Registration of Spent Fuel Cask Use NL-21-021, Registration of Spent Fuel Cask Use2021-04-19019 April 2021 Registration of Spent Fuel Cask Use NL-21-017, Pre-Notice of Disbursement from Decommissioning Trusts2021-04-0808 April 2021 Pre-Notice of Disbursement from Decommissioning Trusts NL-21-010, Submittal of 2020 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report2021-02-17017 February 2021 Submittal of 2020 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report NL-21-006, Relief Request IP3-ISI-RR-16, Proposed Alternative to American Society of Mechanical Engineers Code Case N-513-4 Inspection Requirement2021-02-10010 February 2021 Relief Request IP3-ISI-RR-16, Proposed Alternative to American Society of Mechanical Engineers Code Case N-513-4 Inspection Requirement NL-21-014, Response to 2nd Round Request for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2021-01-26026 January 2021 Response to 2nd Round Request for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-082, Notice of Planned Transfer of Decommissioning Funds2020-12-14014 December 2020 Notice of Planned Transfer of Decommissioning Funds NL-20-081, Pre-Notice of Disbursement from Decommissioning Trusts2020-12-0909 December 2020 Pre-Notice of Disbursement from Decommissioning Trusts NL-20-080, Report in Accordance with 10 CFR 71.95(a) for Failure to Comply with Certificate of Compliance No. 71-93212020-11-19019 November 2020 Report in Accordance with 10 CFR 71.95(a) for Failure to Comply with Certificate of Compliance No. 71-9321 NL-20-079, (IP2 and IP3) - Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic2020-11-12012 November 2020 (IP2 and IP3) - Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic NL-20-077, Submittal of Quality Assurance Program Manual Revision 22020-11-0909 November 2020 Submittal of Quality Assurance Program Manual Revision 2 NL-20-078, Response to Requests for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-11-0909 November 2020 Response to Requests for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-076, Revision of Commitment Related to Nuclear Reactor Safeguards Interim Compensatory Measure - Section B.5.b Issue Regarding Spent Fuel Dispersal2020-11-0202 November 2020 Revision of Commitment Related to Nuclear Reactor Safeguards Interim Compensatory Measure - Section B.5.b Issue Regarding Spent Fuel Dispersal NL-20-069, One-time Scheduler Exemption Request from 10 CFR 50, Appendix E Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Public Health Emergency2020-10-0808 October 2020 One-time Scheduler Exemption Request from 10 CFR 50, Appendix E Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Public Health Emergency NL-20-070, Response to Requests for Additional Information, License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-10-0202 October 2020 Response to Requests for Additional Information, License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-067, Redacted Version of Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-09-16016 September 2020 Redacted Version of Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-064, 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments2020-09-0101 September 2020 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments NL-20-060, Status of Remaining Actions for Generic Letter 2004-022020-08-11011 August 2020 Status of Remaining Actions for Generic Letter 2004-02 NL-20-057, Cancellation of Commitment Related to Large Break LOCA Reanalysis2020-07-30030 July 2020 Cancellation of Commitment Related to Large Break LOCA Reanalysis NL-20-0851, 30-Day 10 CFR 21 Notification - Continuously Energized Eaton D26 Relays Could Fail to Deenergize Because of an Organic C3 Insulating Material2020-07-22022 July 2020 30-Day 10 CFR 21 Notification - Continuously Energized Eaton D26 Relays Could Fail to Deenergize Because of an Organic C3 Insulating Material NL-20-051, Submittal of Quality Assurance Program Manual, Revision 1 for the Indian Point Energy Center2020-07-0707 July 2020 Submittal of Quality Assurance Program Manual, Revision 1 for the Indian Point Energy Center NL-20-052, Unsatisfactory 10 CFR 26 Fitness-For-Duty Blind Performance Testing Results2020-07-0707 July 2020 Unsatisfactory 10 CFR 26 Fitness-For-Duty Blind Performance Testing Results NL-20-012, Application to Revise Provisional Operating License and Technical Specifications2020-06-30030 June 2020 Application to Revise Provisional Operating License and Technical Specifications NL-20-050, Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-06-24024 June 2020 Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-041, Registration of Unit 3 Spent Fuel Cask Use2020-05-13013 May 2020 Registration of Unit 3 Spent Fuel Cask Use NL-20-042, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel2020-05-12012 May 2020 Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel NL-20-033, Technical Specifications Proposed Change - Permanently Defueled Technical Specifications2020-04-28028 April 2020 Technical Specifications Proposed Change - Permanently Defueled Technical Specifications NL-20-038, Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-04-23023 April 2020 Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-035, Response to Request for Additional Information - Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic2020-04-16016 April 2020 Response to Request for Additional Information - Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic NL-20-034, Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic2020-04-13013 April 2020 Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic NL-20-021, Proposed License Amendment to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-03-24024 March 2020 Proposed License Amendment to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-020, Submittal of 2019 Annual Fitness for Duty Performance Data Report Update2020-02-26026 February 2020 Submittal of 2019 Annual Fitness for Duty Performance Data Report Update NL-20-015, Submittal of 2019 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report2020-02-10010 February 2020 Submittal of 2019 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report NL-20-008, Transmittal of Presentation Slides for Partially Closed Pre-Submittal Meeting to Discuss a Planned License Amendment Request to Replace the Fuel Handling Building Crane2020-01-0606 January 2020 Transmittal of Presentation Slides for Partially Closed Pre-Submittal Meeting to Discuss a Planned License Amendment Request to Replace the Fuel Handling Building Crane NL-19-094, 2018 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report2019-12-16016 December 2019 2018 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report NL-19-084, Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments2019-11-21021 November 2019 Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments NL-19-093, Proposed Technical Specifications (TS) Changes - Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.12019-11-21021 November 2019 Proposed Technical Specifications (TS) Changes - Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.1 NL-19-092, Request for Rescission of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2019-11-20020 November 2019 Request for Rescission of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) NL-19-043, Request for Partial Exemption from Record Retention Requirements in 10 CFR 50.122019-10-22022 October 2019 Request for Partial Exemption from Record Retention Requirements in 10 CFR 50.12 NL-19-073, Request for Relaxation of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2019-10-22022 October 2019 Request for Relaxation of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) NL-19-078, Supplement to Technical Specifications Proposed Change - Permanently Defueled Technical Specifications2019-10-22022 October 2019 Supplement to Technical Specifications Proposed Change - Permanently Defueled Technical Specifications NL-19-091, Independent Spent Fuel Storage Installation (Isfsi), Registration of Spent Fuel Cask Use2019-10-17017 October 2019 Independent Spent Fuel Storage Installation (Isfsi), Registration of Spent Fuel Cask Use NL-19-090, Registration of Unit 2 Spent Fuel Cask Use2019-10-0909 October 2019 Registration of Unit 2 Spent Fuel Cask Use NL-19-079, 50.59(d)(2) Summary Report of Changes, Tests and Experiments2019-09-26026 September 2019 50.59(d)(2) Summary Report of Changes, Tests and Experiments 2021-05-06
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- ~Entergy Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5000 Mandy Halter Director, Nuclear Licensing NL-18-019 April 04, 2018 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Relief Request Number IP2-1SI-RR Proposed Alternative fo Use Reactor Vessel Head Penetration Embedded Flaw Weld Repair Method Indian Point Unit 2 Docket No. 50-247 License No. DPR-26
Dear Sir or Madam:
On March 31, 2018, during the Indian Point Unit 2 refueling outage, with the reactor defueled and the head removed and located on the head stand, and all fuel from the reactor vessel removed and located in the spent fuel pool, while performing planned examinations on the 97 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per the requirements of 10CFR50.55a for the reactor coolant system pressure boundary. The planned examinations for the remaining 96 penetrations did not identify any unacceptable Indications. Repairs are planned, and will be completed prior to entering Mode 5 from the current refueling outage.
Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Nuclear Operations, Inc. requests NRG approval of a relief request to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"
on the basis that the current code requirements provide an acceptable level of quality and safety.
The proposed relief request is provided in Attachment 1 to this letter.
In accordance with 10 CFR 50.55a(z)(1 ), the proposed alternatives may be approved by the NRG, provided an acceptable level of quality and safety are maintained. Entergy Nuclear Operations, Inc. requests approval of the proposed alternative in order to support Indian Point Unit 2 return to service from the current refueling outage.
This relief request is required to address an emergent condition at the Indian Point Energy Center. Entergy requests NRG approval as soon as possible but no later than April 12, 2018 at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. This schedule is subject to fluctuation. Relief is requested for the duration of the upcoming Unit 2 operating cycle which is currently expected to conclude in the spring of 2020.
NL-18-019 Page 2 of 2 Entergy is formally committing to submit Westinghouse Letter LTR-SDA-18-035, Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability for Indian Point Unit 2, prior to entering Mode 5 at the conclusion of the present refueling outage. Refer to Attachment 2.
Should you have any questions regarding this matter, please contact Mr. Robert Walpole, Manager, Regulatory Assurance, Indian Point Energy Center at (914) 254-6710.
Sincerely,
/RMj Ntdtii MKH/trj : Relief Request IP2-IS1-RR-06 :- Entergy Commitment r
pc: Mr. David Lew, Acting Regional Administrator, NRC Region I Mr. Richard Guzman, NRA Senior Project Manager NRC Resident Inspector's Office
Attachment 1 to Letter NL-18-019 Relief Request IP2-IS1-RR-06
Attachment 1 NL-18-019 Page 1 of 9 RELIEF REQUEST IP2-IS1-RR-06
- 1. ASME Code Component(s) Affected:
Component: Reactor Pressure Vessel (RPV) Head Penetration #3 Code Class: 1 Exam. Cat.: ASME Code Case N-729-4 Item No. B4.1 O and 84.20 Unit: Indian Point Energy Center, Unit 2 (IP2)
Interval: Fifth (51h)
2. Applicable Code Edition and Addenda
ASME Section XI, 2007 Edition through 2008 Addenda ASME Section XI Code Case N-729-4 ASME Section Ill, 1965 Edition through Summer 1965 Addenda including Code Cases 1332, 1335, 1339, and 1359 (Original Construction Code)
\
ASME Section Ill, Subsection NB, 2001 Edition/2003 Addenda
3. Applicable Code Requirement
IWA-4000 of ASME Section XI contains requirements for removing defects such as weld defects from ASME components. The specific ASME Section XI Code requirements for which use of the proposed alternative is being requested are as follows:
Defects shall be removed or mitigated in accordance with the following requirements:
(a) Defect removal by mechanical processing shall be in accordance with IWA,4462.
(b) Defect removal by thermal methods shall be in accordance with IWA-4461.
(c) Defect removal or mitigation by welding or brazing shall be in accordance with IWA-4411.
(d) Defect removal or mitigation by modification shall be in accordance with IWA-4340.
- Note that use of IWA-4340 for "Mitigation of Defects by Modification" is prohibited by the NRC in 10 CFR 50.55a(b)(2)(xxv).
Attachment 1 NL-18-019 Page 2 of 9
Welding, brazing, fabrication, and installation shall be performed in accordance with the Owner's Requirements and, except as modified below, in accordance with the Construction Code of the item.
(a) Later editions and addenda of the Construction Code, or a later different Construction Cade, either in its entirety or portions thereof, and Code Cases may be used provided the substitution is as fisted in IWA-4221(c).
The Construction Code that will be used for performing defect removal and repair of welds on the RPV head at IP2 is the 2001 Edltion/2003 Addenda of ASME Section Ill.
Requirements applicable to repair of weld defects are specified in NB-4450 as follows:
- NB-4451, which provides general requirements for removal of weld metal defects, states:
Defects in weld metal detected by the examinations required by NB-5000, or by the tests of NB-6000, shall be eliminated and repaired when necessary.
- NB-4452, which provides requirements for eliminating weld surface defects without welding, states:
Weld metal surface defects may be removed by grinding or machining, and need not be repaired by welding, provided that the requirements of (a) through (c) below are met.
(a) The remaining thickness of the section is not reduced below that required by NB-3000.
(b) The depression, after defect elimination, is blended uniformly into the surrounding surface.
(c) The area is examined by a magnetic particle or liquid penetrant method in accordance with NB-5110 after blending and meets the acceptance standards of NB-5300 to ensure that the defect has been removed or reduced to an imperfection of acceptable limit. Defects detected by visual or volumetric method and located an an interior surface need only be reexamined by the method which initially detected the defect when the interior surface is inaccessible for surface examination.
- NB-4453.1, which provides requirements for repairing weld defects with welding, states:
Defects may be removed by mechanical means or by thermal gouging processes. The area prepared for repair shall be examined by a liquid penetrant or magnetic particle method in accordance with NB-5110, and meet the acceptance standards of NB-5340 or NB-5350. This examination is not required where defect elimination removes the full thickness of the weld and where the backside of the weld Joint is not accessible for removal of examination materials.
Attachment 1 NL-18-019 Page 3of 9
4. Reason for Request
IP2 is presently in Refueling Outage 2R23. While performing the VT-2 visual examination of the Reactor Pressure Vessel Head (RPVH) required by Code Case N-729-4 (Item No.
B4.10) 1 , a white substance was observed on the annulus region of RPVH Penetration #3.
Ultrasonic (UT) and eddy current (ET) examinations - performed in accordance with Code Case N-729 of the Control Rod Drive Mechanism (CROM) nozzle of RPVH Penetration
- 3 did not reveal any indications. However, the UT leak path assessment did provide evidence of change from previous outages. It should also be noted that the VT-2 visual examination of the RPVH did not identify any other releva.nt indications or base material wastage.
Follow-up surface examinations were also performed on the partial penetration J-groove weld of RPVH Penetration #3 using the liquid penetrant (PT) and ET techniques. The PT and ET examinations confirmed the presence of relevant indications on the face of the J-groove weld. One of the indications was determined to be axial with a measureable through-weld component and is believed to be the initiation source of leakage. Although the depth of this indication could not be conclusively established, its location and characteristics
{i.e. dark appearance rather than red indicating that water prevented the red die from penetrating the indication) indicate that this indication was likely the through-weld leakage source. See Figure 1 for additional details.
Due to the through-weld indication described above, a weld repair of the subject J-groove weld is necessary. As an alternative to the defect removal and weld repair provisions of ASME Section XI, IWA-4000 and ASME Section Ill, NB-4450, Entergy proposes to repair the subject J-groove weld .using the embedded flaw repair process as described in WCAP-15987-P-A (Reference 1).
- 5. Proposed Alternative and Basis for Use:
Pursuant to 10 CFR 50.55a(z)(1 ), Entergy proposes to repair the J-groove weld of RPVH Penetration #3 using an embedded flaw repair process as an alternative to the defect removal and weld repair provisions of ASME Section XI, IWA-4000 and ASME Section Ill, NB-4450. The proposed embedded flaw repair process is described in WCAP-15987-P-A (Reference 1) and was approved by the NRC in Reference 2. This proposed alternative is requested for,one-cycle of operation as Entergy intends to cease, power operations of IP2 upon completion of Cycle 24 (Spring 2020).
Basis for Use:
Entergy's review concludes that the proposed embedded flaw repair process based on WCAP-15987-P-A provides an acceptable level of quality and safety. In the Safety Evaluation Report for WCAP-15987-P-A, the NRC documented the same conclusion subject to their specified conditions and limitations (Reference 2). Consistent with WCAP-15987-P-A methodology, the followin~ repair requirement~ will be met.
1 Code Case N-729-4 as amended in 10CFR50.55a(g)(6)(ii)(D)
Attachment 1 NL-18-019 Page 4 of 9 5.1 J-Groove Weld Repair Methodology A. The interface boundary between the J-groove weld and stainless steel cladding will be located with a hand-held ferrite meter instrument that identifies this interface boundary. This technique has been used successfully for positive identification of the weld cladding Interface to ensure that all of the Alloy 82/182 material of the J-groove weld is overlaid during the repair. Markings are made to locate the interface as well as a boundary of at least one half inch outboard of the stainless steel cladding and Alloy 82/182 interface.
- 8. The J-groove weld will be sealed/isolated from the primary water environment by deposition of a 360 degree overlay consisting of at least three (3) layers of Alloy 52 or 52M weld metal. The seal weld will extend onto and encompass the outside diameter of the Alloy 600 penetration nozzle by at least 'W'. (The seal weld on the Alloy 600 tube will consist of at least 2 layers of Alloy 52 or 52M weld metal.) It will also extend at least %" beyond the stainless steel cladding interface after deposition of the ER309L buffer layer discussed in paragraph 3 below. Excavation or partial excavation of J-groove weld flaws is not required. See Figure 2 for additional details.
- C. Prior to application of the Alloy 52/52M repair weld layers on the RPVH cladding surface, at least three (3) passes (one layer) of ER309L stainless steel buffer will be installed on the clad surface approximately 0.5 inch beyond the interface of the clad and J-groove weld metal 360 degrees around. These weld passes serve to isolate subsequent ERNiCrFe-7A we.Id passes from contaminants in the cladding.
D. Nondestructive examinations, including preservice, of the completed seal weld repair will be performed as specified in the NRC's Safety Evaluation (Reference 3) forWCAP-15987-P, Revision 2-P-Awith modifications summarized below. As previously stated, Entergy intends to cease power operations of IP2 upon completion of Cycle 24 (Spring-Summer 2020); therefore, this proposed alternative does not include any requirements for inservice inspections.
Axial UT and Surface (Notes 2 and 3)
Circumferential UT and Surface (Notes 2 and 3)
Note:
- 1) Preservice inspection will be consistent with 10CFR 50.55a(g)(6)(ii)(D), which requires the implementation of Code Case N-729-4 with conditions; or NRC approved alternatives to these specified conditions.
- 2) UT personnel and procedures qualified in accordance with 10 CFR 50.55a(g)(6)(ii)(D),
which requires the implementation of Code Case N-729-4 with conditions. Accessible portion of the J-groove repaired region will be examined. The UT plus surface examination coverage must equal 100 percent.
Attachment 1 NL-18-019 Page 5 of 9
- 3) Surface examination shall comply with the acceptance standards of ASME Section Ill, NB-5350 as stipulated in 10 CFR 50.55a(g)(6)(ii)(D). The surface examination of the completed seal weld will include Y:z" of adjacent material but need not include the threads or thread relief at the end of the CROM nozzle.
5.2 Technical Basis for Proposed Alternative A. The purpose of the proposed alternative is to embed and isolate the Alloy 600 (lnconel 82/182) J-groove weld and any identified flaws. The repair overlay welds are not credited for providing structural strength to the original pressure boundary materials.
B. As discussed in WCAP-15987-P, the embedded flaw repair technique is considered a permanent repair. As long as a PWSCC flaw remains isolated from the primary water environment, it cannot propagate. Since an Alloy 52/52M weldment is considered highly resistant to PWSCC, a new PWSCC flaw should not initiate and grow through the Alloy 52/52M seal weld to reconnect the primary water environment with the embedded flaw. Structural integrity of the affected J-groove weld will be maintained by the remaining unflawed portion of the weld.
Alloy 52/52M weld metal is highly resistant to PWSCC, as demonstrated by multiple laboratory tests, as well as over twenty years of service experience in replacement steam generators.
C. WCAP-15987-P notes that the residual stresses produced by the embedded flaw technique have been measured and found to be relatively low because of the small thickness of the seal weld. This implies that no new flaws will initiate and grow in the area adjacent to the repair weld. There are no other known mechanisms for significant flaw propagation in the reactor vessel closure head and penetration tube region since cyclic loading is negligible. In Section 3.2 of the Safety Evaluation forWCAP-15987-P, the NRC recognized this conclusion and its basis which is provided in WCAP-13998, Revision 1 (Reference 4).
D. The thermal expansion properties of Alloy 52 or 52M weld metal are not specified in the ASME Code. In this case, the properties of the equivalent base metal (Alloy 690) should be used. For Alloy 690, the thermal expansion coefficient at 600 degrees Fahrenheit (F) is 8.2E-6 in/in/degree F as found in Section II part D. The Alloy 600 base metal has a coefficient of thermal expansion of 7.8 E-6 in/in/degree F, a difference of about 5 percent. The effect of this small difference in thermal expansion is that the weld metal will contract more than the base metal when it cools, thus producing a compressive stress on the Alloy 82/182 J-groove weld.
This beneficial effect has already been accounted for in the residual stress measurements reported in the technical basis for the embedded flaw repair, as noted in the WCAP-15987-P.
E. Westinghouse letter LTR-SDA-18-035, Fracture Mechanics Assessment qf Embedded Flaw Repair Acceptability for Indian Point Unit 2, will provide a technical basis for returning the Reactor Pressure Vessel to service with the embedded flaw repair of Penetration #3 J-groove weld. This technical basis will include a discussion of the significant experience with embedded flaw repairs
(
Attachment 1 NL-18-019 Page 6 of 9 within the PWR fleet using WCAP-15987-P and ASME Section XI flaw analysis.
The letter will also discuss the applicability and conservatism of the historical analyses relative to the Indian Point 2 repair. The conclusion of the Westinghouse evaluation will provide additional technical basis for one additional cycle of operation.
5.3 NRC Submittals Entergy will submit Westinghouse Letter LTR-SDA-18-035, Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability for Indian Point Unit 2, prior to entering Mode 5 during the present refueling outage, 2R23. Mode 5 is currently expected to occur on April 12, 2018 at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
Entergy's review concludes that the proposed embedded flaw repair of the RPVH Penetration #3 J-groove weld is a technically sound alternative to performing an ASME Section XI Code repair in accordance with IWA-4000 and ASME Section Ill, NB-4450. The embedded flaw weld repair isolates the PWSCC susceptible Alloy 81/182 J-groove weld from the primary water environment by deposition of a 360 degree overlay consisting of Alloy 52 or 52M weld metal. This seal weld will extend at least W' beyond the stainless steel cladding interface and encompass the outside diameter of the Alloy 600 penetration nozzle by at least%". The seal weld will be examined by both surface and UT examination methods as described in Section 5.1. Because Alloy 52/52M welds are considered highly resistant to PWSCC, a new PWSCC flaw should not initiate and grow through the Alloy 52/52M seal weld to reconnect the primary water environment with the embedded flaw.
Furthermore, the existing flaw in the J-groove weld can no longer grow by the PWSCC since the weld is isolated from the reactor coolant environment. Additional technical justification is provided in Section 5.2. In conclusion, Entergy believes that this proposed alternative provides an acceptable level of quality and safety as required by 10 CFR
- 6. Duration of Proposed Alternative:.
This proposed alternative is for one-cycle of operation as Entergy intends to cease power operations of IP2 upon completion of Cycle 24 (Spring 2020).
- 7. Precedent:
- 1. Indian Point, Unit 2 (Entergy) - Relief Request RR-07: Reactor Vessel Head Alternative Weld Repair Methods (Docket No. 50-247), dated August 15, 2007
- 2. Beaver Valley, Unit No. 2 (First Energy Nuclear Operating Company)- Relief Request 2-TYP-3-RV-04, Revision 1: Alternative Examination of Reactor Vessel Head Penetration J-Groove Weld Repairs (Docket No. 50-412), dated November 16, 2016
- 3. Byron Station, Units 1 and 2 (Exelon)- Relief.Request 14R-10, Revision 1: Alternative Requirements far the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) (Docket Nos. STN 50-454 and STN 50-455) dated December 29, 2016
Attachment 1 NL-18-019 Page 7 of 9
- 4. Virgil C. Summer Nuclear Station (SCE&G) Relief Request RR-4-05: Alternative Weld Repair for Reactor Vessel Upper Head Penetrations (Docket 50-395), dated February 24,2014
- 8.
References:
- 1. Westinghouse Topical Report, WCAP-15987-P, Revision 2-P-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, dated 11 December 2003.
- 2. Letter H. N. Berkow (U.S. NRC) to H. A. Sepp (Westinghouse Electric Company),
"Acceptance for Referencing-Topical Report WCAP-15987-P, Revision 2, 'Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetration,'
{TAC NO. MB8997)," dated July 3,2003 [ML031840237]
- 3. Letter from H. N. Berkow (U.S. NRC) to H. A. Sepp (Westinghouse Electric Company),
"Acceptance for Referencing -Topical Report WCAP-15987-P, Revision 2, 'Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations,'
(TAC No. MB8997)," dated July 3, 2003, Accession Number ML031840237
- 4. WCAP-13998, Revision 1, "RV Closure Head Penetration Tube ID Weld Overlay Repair, November 1995 11
Attachment 1 NL-18-019 Page 8 of 9 RPVH Penetration #3 0°PT View Looking up from under RVH J. groove We d to Cl ad - - - - -
tnterface ~
J*groove Weld J-groove Weld to Tube Interface 90° 270° Axia Through*weld tndi caUon - Leak 180° Figure 1: Location of Through-weld Indication
Attachment 1 NL-18-019 Page 9 of 9 Alloy 600 I-groove weld
.,. __..._-.., _ ---***- - l/2 " m n be lo>'I
- ----1--- J*sroove we'd Alloy S2/S2M Seal Weld: minimum of 3 layers over I-groove weld and 2 layers over tube OD; 360 degrees around Figure 2: Repair of J .. groove Weld
Attachment 2 to Letter NL-18-019 Commitment
Attachment 2 NL-18-019 Page 1 of 1 This table identifies actions discussed in this letter for which Entergy commits to perform.
Any other actions discussed in this submittal are described for the NRC's information and are not commitments.
TYPE (Check one) SCHEDULED COMMITMENT ONE- COMPLETION CONTINUING DATE (If Required)
TIME COMPLIANCE ACTION Entergy will submit Westinghouse Letter LTR-SDA-18-035, Fracture Mechanics Assessment of Embedded Flaw Repair Prior to entering Acceptability for Indian Point Unit 2, prior Mode 5 during the to entering Mode 5 during the present X present refueling refueling outage, 2R23. Mode 5 is outage, 2R23.
currently expected to occur on April 12, 2018 at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
.