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MONTHYEARNL-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 Project stage: Request NL-09-130, Official Exhibit - NYS000311-00-BD01 - Response to Request for Additional Information Regarding Relief Request 09 for Fourth Ten-Year Lnservice Inspection Interval2009-09-24024 September 2009 Official Exhibit - NYS000311-00-BD01 - Response to Request for Additional Information Regarding Relief Request 09 for Fourth Ten-Year Lnservice Inspection Interval Project stage: Response to RAI NL-09-164, Response to Request for Additional Information Regarding Relief Request 092009-12-22022 December 2009 Response to Request for Additional Information Regarding Relief Request 09 Project stage: Response to RAI ML15334A2342015-08-10010 August 2015 Official Exhibit - ENT000624-00-BD01 - Letter from N. Salgado to Vice Pres., Operations, Entergy Nuclear Operations, Inc., IP Nuclear Generating Unit No. 2 - Relief from the Examination Area for Reactor Vessel Head Penetration . . . Project stage: Other ML15222A8382015-08-10010 August 2015 ENT000624 - Letter from N. Salgado to Vice President, Operations, Entergy Nuclear Operations, Inc., Indian Point Nuclear Generating Unit No. 2 - Relief from the Examination Area for Reactor Vessel Head Penetration Nozzles (TAC No. ME1658) . Project stage: Other 2009-07-01
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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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Enterrgy Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Robert Walpole Licensing Manager Tel 914 734 6710 July 1,2009 NL-09-090 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Relief Request 08 and 09 For Fourth Ten-Year Inservice Inspection Interval Indian Point Unit Number 2 Docket No. 50-247 License No. DPR-29
Dear Sir or Madam:
Entergy Nuclear Operations, Inc. (Entergy) is submitting Relief Request No. 08 (RR-08)(Enclosure 1), and Relief Request No. 09 (RR-09) (Enclosure
- 2) for Indian Point Unit No. 2 (IP2). These relief requests are for the Fourth 10-year Inservice Inspection (ISI) Interval.The enclosed relief requests each evaluate the proposed alternatives and conclude they provide an acceptable level of quality and safety or that the specified Code requirements would result in unnecessary hardship without a compensating increase in the level of quality and safety. The relief requests are requested under the provisions of 1 OCFR 50.55a(a)(3)(i) and 10 CFR 50.55a(a)(3)(ii).
Entergy requests approval of the relief requests by March 9, 2010, to support the IP2 Refueling Outage (RFO) -2R1 9. The relief requests 08 and 09 result from the recent rule change to 10 CFR 50.55a. The first relief request is to ask for an alternative to use a demonstrated leak path assessment as compliance with Code Case N-729-1. The demonstration of a volumetric leak path assessment is an industry effort that is still uncertain as to the schedule.
This relief request allows the NRC ample time to review the relief request and, if the industry is successful, the relief request can be withdrawn.
The second relief request is due to the elimination of the rule for reactor head inspections which requires a relief request to be submitted to replace the prior rule relaxation.
A6Cd7
.- NL-09-090 Docket No. 50-286 Page 2 of 2 There are no new commitments identified in this submittal.
If you have any questions or require additional information, please contact Mr. Robert Walpole, Licensing Manager at 914-734-6710.Very truly yours, RW/sp
Enclosures:
- 1. Relief Request 08 Proposed Alternative For Demonstrated Leak Path Assessment
- 2. Relief Request 09 Proposed Alternative Examination Area cc:. Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. Samuel J. Collins, Regional Administrator, NRC Region I-NRC Resident Inspector's Office Indian Point Mr. Paul Eddy, New York State Department of Public Service'Mr. Robert Callender, Vice President NYSERDA Enclosure 2 To NL-09-090 RELIEF REQUEST 08 PROPOSED ALTERNATIVE FOR DEMONSTRATED LEAK PATH ASSESSMENT ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247 Indian Point Unit 2 Fourth 10-Year ISI Interval Relief Request No: RR-08 Revision 0 Page 1 of 4 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)
-Hardship or Unusual Difficulty Without a Compensating Increase in Level of Safety or Quality -1 ASME Code Component(s)
Affected Component Number: B4.20 (Per Code Case N-729-1 Table 1)Description:
Reactor Pressure Vessel (RPV) Head Penetration Nozzles (97 locations)
Code Class: 1 2. Applicable Code Edition and Addenda The Code of Record for Indian Point Unit 2 Inservice Inspection Fourth Ten-Year Interval is the ASME Section XI Code, 2001 Edition, 2003 Addenda as augmented by Code Case N-729-1 with limitations/modifications for use stated in 10 CFR 50.55a(g)(6)(ii)(D)(3).
Code Case N-729-1 was approved September 8, 2008 and upon implementation supersedes the First Revised NRC Order EA-03-009.
- 3. Applicable Code Requirement Code Case N-729-1, Section 2500 states that components shall be examined as specified in Table 1. The inspections required in Table 1 consist of a visual examination of the bare-metal surface of the outer surface of the head as shown in Figure 1 and volumetric/surface examination as shown in Figure 2. Alternatively, 10CFR50.55a(g)(6)(ii)(D)(3) allows a demonstrated volumetric or surface leak path assessment to be performed in lieu of the examination requirements of Table 1.4. Reason for Request The wording of 1OCFR50.55a(g)(6)(ii)(D)(3) to perform a demonstrated volumetric or surface leak path assessment through all J-groove welds during the upcoming Indian Point Unit 2 nineteenth refueling outage (2R19) scheduled to start in March 2010 poses a potential hardship due to the expedited implementation of the requirement to perform a demonstrated volumetric leak path assessment and the personnel exposure associated with alternative surface examinations.
The industry has initiated efforts to accomplish a volumetric leak path assessment.
However, the extent of remaining tasks may preclude successful completion in time to support the spring 2010 outage.
Relief Request No: RR-08 Page 2 of 4 The optional surface examination of the J-welds poses a hardship due to the greatly increased personnel radiation exposure associated with this examination technique and the additional risk of heat stress to inspection personnel.
The supplementary scans and additional robotic tool reconfigurations required to accomplish surface examinations result in a significant extension to the examination duration and the accompanying increase in the total dose received.
More importantly, the complicated geometry of the J-weld surface, particularly on penetrations other than those very close to the reactor head center, poses an extremely difficult challenge for remote inspection.
Furthermore, the guide funnels attached to the outside diameter (OD) of the nozzles obstruct access to the J-weld surface Dose rates under the head near the J-weld areas are expected to be in the order of 3-5 Rem/hour range based on previous survey data. In addition, the area under the head is posted as a Locked High Radiation Area and a High Contamination Area.The performance of additional manual volumetric and/or surface exams under these hazardous radiological conditions creates a hardship without a compensating increase in the level of quality and safety.5. Proposed Alternative and Basis For Use The First Revised NRC Order, EA-03-009,Section IV.C(5) contained techniques to be used to meet the inspection requirements of Order Section IV.C and included an assessment to determine if leakage has occurred into the annulus between the reactor pressure vessel head (RPV) penetration nozzle and the RPV head low-alloy steel.In lieu of implementing the demonstrated volumetric or surface leak path assessment through all J-groove welds as imposed by 10 CFR 50.55a(g)(6)(ii)(D)(3), Entergy proposes to perform the same volumetric leak path assessment previously used to meet the requirement of the First Revised NRC Order EA-03-009,Section IV.C.(5).The proposed alternative for the 97 control rod drive mechanism (CRDM) nozzles is a volumetric leak path assessment to determine if leakage has occurred into the annulus between the CRDM nozzle and the RPV head low-alloy steel. The examination region will extend from the bottom of the J-groove weld to a minimum of 1 inch above the highest point of the root of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) on each of the CRDM penetrations.
The volumetric (ultrasonic) leak path assessment technology used on the CRDM nozzles to satisfy the First Revised NRC Order EA-03-009 requirements employs a zero degree incidence longitudinal wave introduced from the tube inside diameter (ID). The response from the tube outside diameter (OD) in the interference fit region is monitored for changes in amplitude due to variations in reflected vs. transmitted energy. Because the tube OD is in contact with the reactor head base material as a result of the interference fit, a portion of the ultrasonic energy is transmitted through this interface.
In the case where leakage into the annulus area between the tube Relief Request No: RR-08 Page 3 of 4 and head base material results in corrosion or steam cutting, the contact is lost in a localized area and a groove is formed. This condition is detected by looking for variations in tube OD response signal amplitude in the reduced contact area as compared to the surrounding areas. In addition, leakage resulting in steam cutting would also be detected by the bare metal visual examination performed on the outer surface of the reactor head.Entergy's inspection vendor, WesDyne International, manufactured a mockup for leak path technique development that simulates the corrosion or steam cutting condition.
The mockup consists of a low alloy carbon steel sleeve with machined ID grooves and holes that is installed over a section of Alloy 600 penetration tube with a 2 millimeter interference fit. Test results demonstrated that the machined grooves and holes in the sleeve are readily detectable using the zero degree amplitude discrimination methodology described above when imaged by the analysis software.The presence of water in these grooves and holes had no effect on the ability of the inspection to detect the grooves and holes.In conjunction with the volumetric leak path assessment, Entergy will also conduct a bare metal visual examination of the outside surface of the head as required by the Code Case.The efficacy of the bare metal visual examination is addressed in MRP 117"Materials ReliabilityProgram Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants" section 3.4, "Protection Against Significant Boric Acid Wastage of the Low Alloy Steel Head" which states in part: "Section 7 of the top-level safety assessment report (MRP-1 10) describes the evaluations that verify that protection against boric acid wastage is provided by the bare metal visual examinations for evidence of leakage required by Sections 5 and 6 of this document." This conclusion is supported by the experience with over 50 leaking CRDM nozzles, including the observation that the large wastage cavity at one plant would have been detected relatively early in the wastage progression had bare metal visual examinations been performed at each refueling outage, and likely even if performed less frequently, with appropriate corrective action. In addition, the wastage modeling presented in MRP-1 10 supports the adequacy of bare metal visual examination performed according to the sensitivity and coverage requirements of Section 5.1 and at the frequency defined in Section 6.Entergy has previously completed volumetric leak path examinations in accordance with the NRC Order on all of the 97 CRDM nozzles. The digitally recorded examination results from those examinations provide an excellent baseline for comparison with the pending 2R1 9 inspections.
Moreover, current appraisals indicate that the existing technology used to perform the volumetric leak path assessment in accordance with the First Revised NRC Order EA-03-009 will not need to be significantly altered to meet the new demonstration obligation.
The combination of a volumetric leak path assessment and bare metal visual examination of the reactor closure head outside surface provides a comprehensive approach for detection of leakage past the J-weld for the CRDM nozzles.
Relief Request No: RR-08 Page 4 of 4 6. Duration of Proposed Alternative Relief is requested for the fourth ten-year interval of the Inservice Inspection Program for Indian Point Unit 2, which began March 1, 2007 and concludes April 3, 2016 or until such time that a volumetric leak path assessment is satisfactorily demonstrated.
- 7. Precedents The combination of a volumetric leak path assessment and bare metal visual examination of the reactor closure head outside surface (in addition to the volumetric examination of the nozzle base material) was previously accepted for meeting the requirements of the First Revised NRC Order EA-03-009,Section IV.C. (5)(b)(i)which states "In addition, an assessment shall be made to determine if leakage has occurred into the annulus between the RPV head penetration nozzle and the RPV head low-alloy steel." 8. References None Enclosure 2 To NL-09-090 RELIEF REQUEST 09 PROPOSED ALTERNATIVE EXAMINATION AREA ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247 Indian Point Unit 2 Fourth 10-Year ISI Interval Relief Request No: RR-09 Revision 0 Page 1 of 3 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)-Alternative Provides an Acceptable Level of Quality and Safety-1 ASME Code Components Affected Examination Category:
B-E Item Number: B4.12
Description:
Control Rod Drive Nozzles Code Class: 1 2. Applicability Code Additions and Addenda The Code of Record for Indian Point Unit 2 Inservice Inspection Fourth Ten-Year Interval is the ASME Section Xl Code, 2001 Edition, 2003 Addenda as augmented by Code Case N-729-1 with limitations/modifications for use stated in 10 CFR 50.55a(g)(6)(ii)(D)(3).
Code Case N-729-1 was approved September 8, 2008 and upon implementation supersedes the First Revised NRC Order EA-03-009.3. Applicable Code Requirement Code Case N-729-1, Section 2500 states that components shall be examined as specified in Table 1 of Code Case N-729-1 and if obstructions or limitations prevent examination of the volume or surface required by Figure 2 for one or more nozzles, the analysis of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface of each nozzle. 10 CFR 50.55a(g)(6)(ii)(D)(6) states that Appendix I of ASME Code Case N-729-1 shall not be implemented without prior NRC approval.Code Case N-729-1, Figure 2, Examination Volume for Nozzle Base Metal and Examination Area for Weld and Nozzle Base Metal, identifies the examination volume or surface as "a = 1.5 in. (38 mm) for Incidence Angle, E, -5 30 deg and for all nozzles -> 4.5 in. (115 mm) OD or 1 in. (25 mm) for Incidence Angle, e, >30 deg; or to the end of the tube, whichever is less." 4. Reason for Request The design of the RPV head penetration nozzles (see Figure 1) includes a threaded section, approximately 3/4 inches long, at the bottom of the nozzles. The dimensional configuration at some nozzles is such that the inspectable distance from the lowest point of the toe of the J-groove weld to the bottom of the scanned region is less than the 1-inch and 1 1/2 inch lower boundary limit as defined in Figure 2 of Code Case N-729-1.
Relief Request No: RR-09 Page 2 of 3 Figure 1 reference datum -bottom of Jjroove weld=-chamfer region (approx 0.25 )threaded region approximately 0.75'5. Proposed Alternative and Basis For Relief Use Appendix I of Code Case N-729-1 to define an alternative examination area/volume to that defined in Figure 2 of the Code Case.Perform UT from the inside surface of each RPV head penetration nozzle from 1-inch and 1 1/2 inch above the J-groove weld (i.e., the upper boundary limit defined
.Relief Request No: RR-09 Page 3 of 3 in Figure 2 of Code Case N-729-1) and extending down the nozzle to at least the top of the threaded region. Table 1 provides the minimum inspection coverage required to ensure that a postulated axial through-wall flaw in the un-inspected area of the CRDM penetration nozzle will not propagate into the pressure boundary formed by the J-groove weld prior to a subsequent inspection (i.e. 2 Effective Full Power Years, EFPY). The time estimates are more than the time between successive inspections.
This exam provides reasonable assurance that structurally significant flaws will not exist at or above the weld root and assure that operation between refueling outages can be accomplished without pressure boundary leakage from the examined nozzles.TABLE 1 IP2 RPV Head Penetrations
-Minimum Inspection Coverage Requirements Below the J-Groove Weld to ensure structural integrity and leak tightness between inspections Nozzle Angle of ) Minimum Required UT Time (EFPY)Penetration No. Incidence Coverage Below J-Groove to Reach the (Degrees)
Weld with > 2 EFPY by Lowest Point Crack Growth Evaluation of the Toe of (Inches) the J-Groove Weld 1 through 25 0 to 23.3 0.55 4.6 26 through 69 24.8 to 38.6 0.45 4.4 70 through 81 44.3 0.25 8.4 82 through 89 45.4 0.25 6.8 90 through 97 48.7 0.18 5.0 Note: (1) Length below the lowest point at the toe of the J-groove weld (downhill side) that has an operating stress level of 20 ksi: 0.86 inches at nozzles 1 through 25; 0.40 inches at nozzles 26 through 69; 0.32 inches at nozzles 70 through 81 0.34 inches at nozzles 82 through 89, and 0.32 inches at nozzles 90 through 97.6. Duration of Propose Alternative Relief is requested for the fourth ten-year interval of the Inservice Inspection Program for Indian Point Unit 2, which began March 1, 2007 and concludes April 3,2016.7. Precedents
- 1. Safety Evaluation for Unit 3, "Relaxation of First Revised Order on Reactor Vessel Nozzles, Indian Point No. 2 (TAC MC9230) dated February 27, 2006.8. References None