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Category:Code Relief or Alternative
MONTHYEARML21299A0032021-10-28028 October 2021 and Waterford Steam Electric Station, Unit 3 - Approval of Request for Alternative EN-20-RR-003 from Certain Requirements of the ASME Code ML21054A3302021-02-24024 February 2021 Approval of Alternative IP3-ISI-RR-16 to American Society of Mechanical Engineers Code Case N-513-4 Inspection Requirement CNRO-2020-00016, Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2020-08-12012 August 2020 Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) ML19254A6032019-09-19019 September 2019 Units 2 and 3; Palisades Nuclear Plant; River Bend; and Waterford Steam Electric Station, Unit 3 - Relief Request No. EN-19-RR-1, Use of ASME Code Case N-831-1 ML19039A1492019-02-25025 February 2019 Issuance of Relief Request IP3-ISI-RR-14 Alternative Examination Required by ASME Code Case N-724-4 CNRO-2019-00002, Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-01-31031 January 2019 Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 ML18251A0042018-09-18018 September 2018 Safety Evaluation for Relief Request IP3-ISI-RR-11, RR-12, RR-15 Approval of Alternative Associated with Extension of Fourth Interval Reactor Vessel and Piping Weld Inspections (EPID: L-2017-LLR-0124,0127) ML18193B0302018-07-18018 July 2018 Safety Evaluation for Relief Request IP3-ISI-RR-13 Fourth Ten-year Inservice Inspection Interval Extension ML18128A0672018-06-0808 June 2018 Arkansas, Units 1 and 2; Grand Gulf, Unit 1; Indian Point Unit Nos. 2 and 3; Palisades; Pilgrim; River Bend Station, Unit 1; and Waterford, Unit-3 Relief Request No. EN-17-RR-1, Alternative to Use ASME Code Case N-513-4 ML18099A3732018-04-0909 April 2018 04/09/2018 E-mail from R. Guzman to R. Walpole, Verbal Authorization for Relief Request IP2-ISI-RR-06 ML18059A1562018-03-0606 March 2018 Safety Evaluation for Relief Request IP2-ISI-RR-05 Alternative Examination Volume Required by ASME Code Case N-729-4 ML18005A0662018-01-23023 January 2018 Safety Evaluation of Relief Requests ISI-RR-20, ISI-RR-21, and ISI-RR-22 Regarding the Fourth 10-Year Interval of the Inservice Inspection Program CNRO-2017-00022, Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 12017-11-17017 November 2017 Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 1 ML17174B1442017-07-12012 July 2017 Relief Request for EN-ISI-16-1 Regarding Use of Later Edition and Addenda of the ASME Code ML17069A2832017-03-16016 March 2017 Relief Request No. IP3-ISI-RR-09, for Alternative to the Depth Sizing Qualification Requirement ML16358A4442017-01-11011 January 2017 Relief from the Requirements of the ASME Code Regarding Alternate IP3-RR-10 to the Full Circumferential Inspection Requirement of Code Case N-513-3 ML16167A0812016-07-15015 July 2016 Request for Alternative IP2-ISI-RR-03 to Weld Reference System Examination Required by ASME Code Subarticle IWA-2600 ML16096A2692016-06-0606 June 2016 Arkansas; Grand Gulf; James A. Fitzpatrick; Indian Point; Palisades; Pilgrim; River Bend; and Waterford - Relief Request RR-EN-15-2, Proposed Alternative to Use ASME Boiler and Pressure Vessel Code Case N-786-1 ML16053A0252016-03-0303 March 2016 IP2-ISI-44-18, Relief from the Requirements of the ASME Code CNRO-2015-00017, Entergy Submits Relief Request RR EN-15-1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Waster Service, Section XI, D2015-06-0505 June 2015 Entergy Submits Relief Request RR EN-15-1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Waster Service, Section XI, Division ML14198A3312014-07-23023 July 2014 Safety Evaluation for Relief Request IP3-ISI-RR-06 for Reactor Vessel Weld Examinations (Tac No. MF3345) NL-13-041, Relief Request IP2-ISI-RR-17: Code Case N-770-1 Weld Inspection Frequency Extension2013-02-20020 February 2013 Relief Request IP2-ISI-RR-17: Code Case N-770-1 Weld Inspection Frequency Extension ML12334A3172012-12-0303 December 2012 Relief Request IP2-ISI-RR-15 - Proposed Alternative to the Use of a Weld Reference System NL-12-065, 2012 Summary Report for In-Service Inspection and Repairers and Replacements2012-06-13013 June 2012 2012 Summary Report for In-Service Inspection and Repairers and Replacements NL-12-069, Unit Number 2, Relief Request IP2-1SI-RR-15 - Proposed Alternative to the Use of a Weld Reference System2012-05-23023 May 2012 Unit Number 2, Relief Request IP2-1SI-RR-15 - Proposed Alternative to the Use of a Weld Reference System ML11105A1222011-04-25025 April 2011 Relief from the Requirements of the ASME Code to Perform Essentially 100 Percent Volumetric Examination of the Weld and Adjacent Base Material for the Third 10-Year Inservice Inspection ML11109A0162011-04-25025 April 2011 Relief Request No. IP2-ISI-RR-12, Reactor Vessel Shell-To-Flange Weld Inspection for the Fourth 10-Year Inservice Inspection Interval (Tac No. ME5180) NL-10-136, Submittal of 10 CFR 50.55a Relief Request IP2-ISI-RR-12 for 4th Ten-Year Inservice Inspection Interval2010-12-14014 December 2010 Submittal of 10 CFR 50.55a Relief Request IP2-ISI-RR-12 for 4th Ten-Year Inservice Inspection Interval ML1017400482010-07-15015 July 2010 Relief Request RR-11 for the Fourth 10-Year Inservice Inspection Interval NL-10-061, CFR 50.55a Relief Requests RR-3-49 and RR-3-50 from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third Ten-Year Inservice Inspection Interval2010-07-0505 July 2010 CFR 50.55a Relief Requests RR-3-49 and RR-3-50 from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third Ten-Year Inservice Inspection Interval ML1015303122010-06-0707 June 2010 Relief Request RR-02 for the Fourth 10-Year Inservice Inspection Interval NL-09-022, Supplement to Request for Relief 3-48 and 3-47 (I) to Support Refuel Outage 15 Inservice Inspection Program2009-02-0606 February 2009 Supplement to Request for Relief 3-48 and 3-47 (I) to Support Refuel Outage 15 Inservice Inspection Program NL-09-0111, Submittal of Relief Requests No. 3-45, 3-46, 3-47(I) and 3-48 to Support the Unit 3 Refuel Outage 15 Inservice Inspection Program2009-01-22022 January 2009 Submittal of Relief Requests No. 3-45, 3-46, 3-47(I) and 3-48 to Support the Unit 3 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-08-096, Request for Relief to Extend Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses2008-07-0808 July 2008 Request for Relief to Extend Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses ML0721304872007-09-0505 September 2007 Relief Request No. RR-01 NOC-AE-06002031, Request for Relief from ASME Boiler and Pressure Vessel Code, Section XI Requirements for Use of Penetrameters in Radiographic Examinations2006-06-14014 June 2006 Request for Relief from ASME Boiler and Pressure Vessel Code, Section XI Requirements for Use of Penetrameters in Radiographic Examinations ML0602600762006-02-0808 February 2006 Relief Request (RR) No. 74 NL-05-0720, Request for Relief to Extend the Third 10-Year Inservice Inspection Interval for the Reactor Vessel Weld Examination2005-06-0808 June 2005 Request for Relief to Extend the Third 10-Year Inservice Inspection Interval for the Reactor Vessel Weld Examination ML0509401362005-04-0404 April 2005 Relief, Relaxation of First Revised Order on Reactor Vessel Nozzles ML0507700102005-03-18018 March 2005 Relaxation of First Revised Order on Reactor Vessel Nozzles ML0427406642004-10-14014 October 2004 Relief Request Nos. R-33, R-71, R 3-40(A) and R-41, James A. FitzPatrick Nuclear Power Plant, Indian Point Nuclear Generating Unit Nos. 2 and No. 3 and Pilgrim Nuclear Power Station ML0427406282004-10-14014 October 2004 Relief Request Nos. 65, 66, 3-34 and 3-35 Regarding Alternative Nondestructive Examination Qualification Requirements ML0425203922004-10-0505 October 2004 Relief, Requirements of American Society of Mechanical Engineers Boiler & Pressure Vessel Code, Section III, 1965 Edition, & Section XI, 1989 Edition, for Repair & Inspection of Reactor Pressure Vessel Head Penetrations ML0418901542004-07-0707 July 2004 Relief, Relief Request Nos. RR-67 and RR 3-36, TAC Nos. MC1698 and MC1699 ML0410700882004-07-0606 July 2004 Relief Request to Use American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-600 ML0408205162004-03-22022 March 2004 Relief Request Nos. RR-68, RR3-37, and PRR-34 (TAC MC1559, MC1560, & MC1561) ML0408506682004-03-19019 March 2004 Relief Request Nos. 70 and 3-39 Regarding Alternative Depth Sizing Criteria.(Tac MC1696 & MC1697) ML0408600062004-03-19019 March 2004 Relief Request No. RR 63 Regarding risk-informed Inservice Inspection Program ML0335000092003-12-16016 December 2003 Inservice Testing Program Relief Request Nos. 47 and 48, MB9111 and MB9112 2021-02-24
[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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Text
Entergy Nuclear Northeast Entergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 John A. Ventosa Site Vice President Administration February 20, 2013 NL-13-041 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Relief Request IP2-ISI-RR-17: Code Case N-770-1 Weld Inspection Frequency Extension Indian Point Unit Number 2 Docket No. 50-247 License No. DPR-26
REFERENCES:
1 Entergy Letter NL-1 1-094 Regarding Request for Relief 14 - Code case N-770-1 Weld Inspection Frequency Extension, dated August 3, 2011 2 Entergy Letter NL-1 1-123, Regarding Request for Additional Information on Relief Request IP2-ISI-RR-14 for Code Case N-770-1 Weld Inspection Frequency Extension (TAC No. ME6689), dated November 8, 2011.
3 NRC Letter Regarding Relief Request No. IP2-ISI-RR-14, Code Case N-770-1, Reactor Coolant System Cold Leg Nozzle Weld Inspection Frequency Extension (TAC NO. ME6801), dated February 2, 2014
Dear Sir or Madam:
Entergy Nuclear Operations, Inc. (Entergy) is submitting Relief Request No. 17 (IP2-ISI-RR-
- 17) (Attachment) for Indian Point Unit No. 2 (IP2). This relief request is for the Fourth 10-year Inservice Inspection (ISI) Interval.
The purpose of this relief request is to extend the inspection of the reactor vessel cold leg nozzle to safe-end welds (21-14A,22-14A, 23-14A,24-14A), which are Alloy 600 welds covered by Code Case N-770-1, Table 1, Inspection Item B. The requested extension is until Refueling Outage 22 (2R22) which is scheduled for Spring 2016. This request is made in accordance with 10 CFR 50.55a(a)(3)(i), an alternative provides an acceptable level of quality and safety.
NL-13-041 Docket 50-247 Page 2 of 2 Entergy submitted a similar Relief Request in Reference 1 and responded to a request for additional Information in Reference 2. This Relief Request was approved in Reference 3.
Additional circumstances have arisen affecting the need for this request.
Entergy requests approval of the relief request by August 2013. Performance of this inspection in 2R21 would require planning to start at that time.
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 914-254-6710.
Sincerely, JV/sp
Attachment:
Relief Request No IP2-ISI-RR-17 Code Case N-770-1 Weld Inspection Frequency Extension cc: Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORL Mr. William M. Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office Indian Point Ms. Bridget Frymire, New York State Department of Public Service Mr. Francis J. Murray, Jr., President and CEO, NYSERDA
ATTACHMENT TO NL-13-041 RELIEF REQUEST NO IP2-ISI-RR-17 CODE CASE N-770-1 WELD INSPECTION FREQUENCY EXTENSION ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247
NL-13-041 Attachment Docket No. 50-247 Page 1 of 5 Indian Point Unit 2 Fourth 10-year ISI Interval Relief Request No: IP2-ISI-RR-17 Code Case N-770-1 Weld Inspection Frequency Extension Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
-Alternative Provides Acceptable Level of Quality and Safety-
- 1. ASME Code Component(s) Affected The affected components are the Indian Point Unit 2 (IP2) reactor vessel cold leg nozzle to safe-end welds (21-14A,22-14A, 23-14A,24-14A), which are Alloy 600 welds covered by Code Case N-770-1, Table 1, Inspection Item B.
These welds had an Alloy 600 ID onlay installed during original fabrication and do not join any cast stainless steel materials.
Examination Inspection Category Item Description CC N-770-1 B Weld 21-14A - Loop 21 cold leg nozzle to safe-end weld CC N-770-1 B Weld 22-14A - Loop 22 cold leg nozzle to safe-end weld CC N-770-1 B Weld 23-14A - Loop 23 cold leg nozzle to safe-end weld CC N-770-1 B Weld 24-14A - Loop 24 cold leg nozzle to safe-end weld
2. Applicable Code Edition and Addenda
Code Case N-770-1 as referenced in 10CFR50.55a(g)(6)(ii)(F).
3. Applicable Code Requirement
Table 1 of Code Case N-770-1, requires volumetric examination of essentially 100% of Inspection Item B pressure retaining welds once every second inspection period not to exceed 7 years.
4. Reason for Request
Relief is being requested at this time to extend the cold leg weld inspection until Refueling Outage 22 (2R22) scheduled for Spring 2016 to allow the refueling cavity liner to be repaired in order to maximize the water level in the cavity during inspection activities in order to minimize dose.
Examination of Item A-2 (Hotleg) and Item B (Coldleg) welds are performed from the inside surface of the pipe (ID) at IP2 due to extremely limited access provisions from the outside surface of the pipe. The IP2 Item A-2 and Item B welds are located inside a "sandbox" which
NL-13-041 Enclosure 1 Docket No. 50-247 Page 2 of 5 was installed during original plant construction after all welding was completed. Additionally, these welds are covered with asbestos insulation. The cost and personnel radiation exposure (approximately 11 Rem) to perform these examinations from the OD make the OD exam undesirable. The inspection of the Item A-2 (Hotleg) welds from the ID does not require removal of the reactor vessel core barrel, while the inspection of the Item B (Coldleg) welds from the ID does require removal of the reactor vessel core barrel.
Baseline inspections of Code Case N-770-1 Inspection Item B welds,21-14A, 22-14A,23-14A and 24-14A were performed in May 2006. The ultrasonic examinations performed in 2006 met Section Xl, Appendix VIII requirements, including examination volume of essentially 100%. The Safety Evaluation Report provided in Reference 3 requires examination of these welds by March 2014. Therefore, inspection of these welds would require removal of the core barrel during the March 2014 refueling outage.
Since inspection of these welds requires that the core barrel be removed from the reactor vessel, these inspections had previously been planned to be performed concurrently with the vessel shell weld inspections and the vessel internals inspections required by MRP-227 during the refuel outage of 2014. A separate IP2 Relief Request IP2-ISI-RR-16 has been submitted to the NRC staff to allow deferral of the vessel shell weld inspections from 2014 to 2016.
Removal of the Core Barrel and the lower internals requires that the water level in the refueling cavity to be increased to minimize the radiation fields since the height of the core barrel is greater than the depth of the water level during normal refueling operations. This increased water level and the displacement due to the weight from the core barrel and lower internals results in a significant increase in leakage through the existing cavity liner defects. This makes it more difficult to stabilize the water level at a higher value. IPEC is currently planning on repairing these liner indications during the 2014 refueling outage. Therefore, deferral of the Cold Leg Nozzle inspections from the 2014 to the 2016 refueling outage would eliminate the increased cavity liner leakage associated with the removal of the core barrel.
Repair of the liner would allow better control of the water level in the cavity and this water level must be maximized to minimize dose. The Core barrel (lower internals) is stored in the lower cavity stand and the Upper Internals are stored in the Upper Internals stand in the upper cavity.
The repair of the cavity liner is expected to allow the maximized refueling cavity water level to be maintained because leakage will have been reduced or eliminated. Maximizing the water level reduces dose by approximately a factor of 10 if the water level is six inches higher. The dose rate for a water level of 94 feet 2 inches is about 18.6 R/hour at the cavity level and at 94 feet 8 inches is 1.47 R/hour.
IP2 is currently planning to perform the MRP-227 (i.e. Vessel Internals) inspections in 2R22 since the actual inspection scope has not yet been finalized (i.e. Entergy is still performing internals evaluations in response to NRC RAIs and these evaluations have the potential to impact the MRP-227 inspection scope). In addition, a significant pre-outage effort will be required to finalize inspection tooling and acceptance criteria which can not be completed prior to 2R21 which is currently scheduled to begin on February 24, 2014.
NL-13-041 Enclosure 1 Docket No. 50-247 Page 3 of 5
- 5. Proposed Alternative and Basis for Use 10CFR50.55a(a)(3) states:
"Proposed alternatives to the requirements of (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director of Nuclear Reactor Regulation. The applicant shall demonstrate that:
(i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
Entergy believes that the proposed alternatives of this request provide an acceptable level of quality and safety.
IP2 proposes a one time extension to the Code Case N-770-1, Table 1, Inspection Item B, volumetric examinations from a period of not to exceed 7 years to a period of not to exceed 10 years. The inspections which are currently required to be performed will be performed not later than the March 2016 refueling outage.
Operating experience on Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82/182 welds show that weld repairs performed during original plant construction are a significant contributor in the initiation and propagation of cracking. A review of the construction records and a weld repair search performed for the IP2 Reactor Vessel nozzle Alloy 82/182 welds did not identify any weld repairs performed on these welds during original plant construction.
Additionally, IP2 has implemented Zinc injection since December 2007 which contributes to lower probability of PWSCC crack initiation during future plant operation.
The susceptibility to PWSCC of Alloy 82/182 welds is largely a function of time at temperature.
Since IP2 operated at a low cold leg temperature (< 535F) for a significant portion of its operating life, it is ranked only moderately susceptible to PWSCC based on the susceptibility formula provided in NRC Order EA 03-009 for the upper vessel head penetrations.
Examination of Item A-2 (Hotleg) and Item B (Coldleg) welds are performed from the ID at IP2 due to extremely limited access provisions from the outside surface of the pipe. The IP2 Item A-2 and Item B welds are located inside a "sandbox" which was installed during original plant construction after all welding was completed. Additionally, these welds are covered with asbestos insulation. The estimated cost ($750,000) and personnel radiation exposure (approximately 11 Rem) to perform these examinations from the OD make the OD exam undesireable. The inspection of the Item A-2 (Hotleg) welds from the IDdoes not require
NL-13-041 Enclosure 1 Docket No. 50-247 Page 4 of 5 removal of the reactor vessel core barrel, while the inspection of the Item B (Coldleg) welds from the ID does require removal of the reactor vessel core barrel.
In March 2012, ultrasonic (volumetric) and eddy current (surface) exams were performed on the Code Case N-770-1 Inspection Item A-2 (Hotleg) welds and no indications were identified. In 2014, ultrasonic (volumetric) and eddy current (surface) exams are scheduled to be performed on the Code Case N-770-1 Inspection Item A-2 (Hotleg) welds. Since PWSCC is temperature dependant, it would be expected that Inspection Item A-2 (Hotleg) welds would show evidence of crack initiation before Inspection Item B (Coldleg) welds. Therefore, the lack of any indications in the Inspection Item A-2 (Hotleg) welds provides added assurance that the one time extension of the inspection of the Inspection Item B (Coldleg) welds by two years provides an acceptable level of quality and safety.
The baseline inspection requirements of the Code Case N-770-1 Inspection Item B (Coldleg) welds, as required by Code Case N-770-1-2200, were satisfied by crediting the MRP-139 examination that was performed in May 2006. At that time, in addition to the ultrasonic (volumetric) examination, an additional surface examination utilizing an eddy current technique was performed. Both the ultrasonic (volumetric) and eddy current (surface) examinations were performed from the ID surface and confirmed the absence of any unacceptable indications after approximately 33 years of operation. The ultrasonic examinations performed in 2006 met Section XI, Appendix VIII requirements, including examination volume of essentially 100%. The use of eddy current examination in addition to the Code Case N-770-1 required ultrasonic examination provides a higher probability of detection of smaller flaws than an ultrasonic examination alone. Since the Code Case N-770-1 inspection frequency is based on flaw sizes associated with ultrasonic examination, the proposed alternative provides an equivalent protection against unacceptable PWSCC as the Code Case N-770-1 exam schedule. to Reference 2 provides a flaw tolerance analysis performed by Westinghouse for the IP2 RPV inlet nozzle to safe end DM welds. The purpose of this analysis was to assess the impact of extending the inspection frequency beyond the 7 year inspection frequency required by Code Case N-770-1. to Reference 2 calculated the length and the depth of the largest axial and circumferential flaws which, if left in service for more than 7 years would not grow beyond the limits provided in sub section IWB-3600 of the ASME Section XI Code.
The evaluation also established the maximum flaw size which could have reasonably been missed during the 2006 inspection considering the detection capabilities of the NDE techniques used during the inspection. It was estimated that the volumetric inspection technique (i.e.
Ultrasound) used during the inspection was capable of reliably detecting a 10% through wall, surface breaking flaw. This size flaw was then assumed to have been missed during the 2006 inspection even though a surface examination (i.e. Eddy Current) did not identify any surface breaking flaws. Based on this, assuming that the 2006 inspection could have missed a 10%
through wall flaw is a conservative assumption.
NL-13-041 Enclosure 1 Docket No. 50-247 Page 5 of 5 A comparison of the depth of the maximum flaw which could have been reasonably missed during the 2006 inspection (i.e. 0.25") with the depth of the maximum flaw (i.e. 1.05") which would not grow beyond the ASME Section Xl limits after 10 years of service results in a margin of conservatism of approximately 4 beyond the margins of safety required by the ASME Section Xl Code. Based on these results and the details provided in the Enclosure 1 to Reference 2 evaluation, it is concluded that performing the next cold leg nozzle DM weld inspections during the 2016 refueling outage is acceptable since it will not result in flaws which exceed the ASME Section Xl Code limits.
Based on the fact that no weld repairs were documented on these welds during plant construction, zinc addition which decreases the probability of PWSCC crack initiation has been implemented at IP2 since December 2007, the hotleg examinations including both ultrasonic and eddy current inspections were performed in 2012 with no indications, hotleg examinations including both ultrasonic and eddy current inspections are scheduled to be performed in 2014, and the coldleg examinations including both ultrasonic and eddy current inspections were performed from the ID in 2006 with no indications, the one time alternative inspection frequency of every 10 years instead of every 7 years provides an acceptable level of quality and safety.
- 6. Duration of Proposed Alternative This request is applicable to Entergy's inservice inspection program for the fourth interval for Indian Point Unit 2.
- 7. References
- 1. Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of listed Mitigation Activities Section Xl, Division 1.
- 2. Entergy Letter NL-1 1-123, Regarding Request for Additional Information on Relief Request IP2-1SI-RR-14 for Code Case N-770-1 Weld Inspection Frequency Extension (TAC No.
ME6689), dated November 8, 2011.
- 3. NRC Letter Regarding Relief Request No. IP2-1SI-RR-14, Code Case N-770-1, Reactor Coolant System Cold Leg Nozzle Weld Inspection Frequency Extension (TAC NO. ME6801)