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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
MONTHYEARML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status IR 05000003/20240022024-08-0606 August 2024 NRC Inspection Report 05000003/2024002, 05000247/2024002, 05000286/2024002 PNP 2024-030, Update Report for Holtec Decommissioning International Fleet Decommissioning Quality Assurance Program Rev. 3 and Palisades Transitioning Quality Assurance Plan, Rev 02024-08-0202 August 2024 Update Report for Holtec Decommissioning International Fleet Decommissioning Quality Assurance Program Rev. 3 and Palisades Transitioning Quality Assurance Plan, Rev 0 ML24171A0122024-06-18018 June 2024 Reply to a Notice of Violation EA-24-037 ML24156A1162024-06-0404 June 2024 Correction - Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities IR 05000003/20240052024-05-21021 May 2024 and 3 - NRC Inspection Report Nos. 05000003/2024005, 05000247/2024005, 05000286/2024005, 07200051/2024001, and Notice of Violation ML24128A0632024-05-0707 May 2024 Submittal of 2023 Annual Radiological Environmental Operating Report L-24-009, HDI Annual Occupational Radiation Exposure Data Reports - 20232024-04-29029 April 2024 HDI Annual Occupational Radiation Exposure Data Reports - 2023 ML24116A2412024-04-25025 April 2024 Annual Environmental Protection Plan Report ML24114A2282024-04-23023 April 2024 Annual Radioactive Effluent Release Report L-24-007, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations – Holtec Decommissioning International, LLC (HDI)2024-03-29029 March 2024 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations – Holtec Decommissioning International, LLC (HDI) ML24080A1722024-03-20020 March 2024 Reply to a Notice of Violation EA-2024-010 IR 05000003/20240012024-03-20020 March 2024 NRC Inspection Report Nos. 05000003/2024001, 05000247/2024001, and 05000286/2024001 (Cover Letter Only) ML24045A0882024-02-22022 February 2024 Correction to the Technical Specifications to Reflect Appropriate Pages Removed and Retained by Previous License Amendments ML24053A0642024-02-22022 February 2024 2023 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report IR 05000003/20230042024-02-22022 February 2024 NRC Inspection Report Nos. 05000003/2023004, 05000247/2023004, 05000286/2023004, and 07200051/2023004 and Notice of Violation ML24011A1982024-01-12012 January 2024 ISFSI, Notice of Organization Change for Site Vice President ML23342A1082024-01-0909 January 2024 – Independent Spent Fuel Storage Installation Security Inspection Plan ML23353A1742023-12-19019 December 2023 ISFSI, Emergency Plan, Revision 23-04 L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 ML23326A1322023-12-0505 December 2023 Issuance of Amendment No. 67, 300 & 276 to Implement the Independent Spent Fuel Storage Installation Only Emergency Plan ML23339A0442023-12-0505 December 2023 Issuance of Amendment No. 68, 301 and 277 Regarding Changes to Implement the Independent Spent Fuel Storage Installation Physical Security Plan ML23338A2262023-12-0404 December 2023 Signed Amendment No. 27 to Indemnity Agreement No. B-19 ML23356A0212023-12-0101 December 2023 American Nuclear Insurers, Secondary Financial Protection (SFP) Program ML23242A2772023-11-30030 November 2023 NRC Letter Issuance - IP LAR for Units 2 and 3 Renewed Facility Licenses and PDTS to Reflect Permanent Removal of Spent Fuel from SFPs ML23338A0482023-11-30030 November 2023 ISFSI, Report of Changes to Physical Security, Training and Qualification, Safeguards Contingency Plan, and ISFSI Security Program, Revision 28 ML22339A1572023-11-27027 November 2023 Letter - Indian Point - Ea/Fonsi Request for Exemptions from Certain Emergency Planning Requirements for 10 CFR 50.47 and 10 CFR Part 50, Appendix E IR 05000003/20230032023-11-21021 November 2023 NRC Inspection Report Nos. 05000003/2023003, 05000247/2023003, 05000286/2023003, and 07200051/2023003 ML23100A1252023-11-17017 November 2023 Letter and Enclosure 1 - Issuance Indian Point Energy Center Units 1, 2, and 3 Exemption for Offsite Primary and Secondary Liability Insurance Indemnity Agreement ML23100A1172023-11-17017 November 2023 NRC Response - Indian Point Energy Center Generating Units 1, 2, and 3 Letter with Enclosures Regarding Changes to Remove the Cyber Security Plan License Condition ML23050A0032023-11-17017 November 2023 Letter - Issuance Indian Point Unit 2 License Amendment Request to Modify Tech Specs for Staffing Requirements Following Spent Fuel Transfer to Dry Storage ML23100A1432023-11-16016 November 2023 Letter - Issuance Indian Point Energy Center Generating Units 1, 2, and 3 Exemption Concerning Onsite Property Damage Insurance (Docket Nos. 50-003, 50-247, 50-286) L-23-012, Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point2023-11-13013 November 2023 Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point ML23064A0002023-11-13013 November 2023 NRC Issuance for Approval-Indian Point EC Units 1, 2 and 3 Emergency Plan and Emergency Action Level Scheme Amendments ML23306A0992023-11-0202 November 2023 and Indian Point Energy Center, Notification of Changes in Schedule in Accordance with 10 CFR 50.82(a)(7) ML23063A1432023-11-0101 November 2023 Letter - Issuance Holtec Request for Indian Point Energy Center Generating Units 1, 2, and 3 Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and Part 50 ML23292A0262023-10-19019 October 2023 LTR-23-0211-RI Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report-RI ML23289A1582023-10-16016 October 2023 Decommissioning International - Registration of Spent Fuel Casks and Notification of Permanent Removal of All Indian Point Unit 3 Spent Fuel Assemblies from the Spent Fuel Pit ML23270A0082023-09-27027 September 2023 Registration of Spent Fuel Casks ML23237A5712023-09-22022 September 2023 09-22-2023 Letter to Dwaine Perry, Chief, Ramapo Munsee Nation, from Chair Hanson, Responds to Letter Regarding Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River ML23242A2182023-09-12012 September 2023 IPEC NRC Response to the Town of New Windsor, Ny Board Certified Motion Letter Regarding Treated Water Release from IP Site (Dockets 50-003, 50-247, 50-286) ML23250A0812023-09-0707 September 2023 Registration of Spent Fuel Casks ML23255A0142023-08-31031 August 2023 LTR-23-0211 Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report IR 05000003/20230022023-08-22022 August 2023 NRC Inspection Report 05000003/2023002, 05000247/2023002, 05000286/2023002, and 07200051/2023002 ML23227A1852023-08-15015 August 2023 Request for a Revised Approval Date Regarding the Indian Point Energy Center Permanently Defueled Emergency Plan and Emergency Action Level Scheme ML23222A1442023-08-10010 August 2023 Registration of Spent Fuel Casks ML23208A1642023-07-26026 July 2023 Village of Croton-on-Hudson New York Letter Dated 7-26-23 Re Holtec Wastewater ML23200A0422023-07-19019 July 2023 Registration of Spent Fuel Casks ML23235A0602023-07-17017 July 2023 LTR-23-0194 Dwaine Perry, Chief, Ramapo Munsee Nation, Ltr Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River 2024-09-18
[Table view] Category:Safety Evaluation
MONTHYEARML23339A0442023-12-0505 December 2023 Issuance of Amendment No. 68, 301 and 277 Regarding Changes to Implement the Independent Spent Fuel Storage Installation Physical Security Plan ML23326A1322023-12-0505 December 2023 Issuance of Amendment No. 67, 300 & 276 to Implement the Independent Spent Fuel Storage Installation Only Emergency Plan ML23243A8452023-11-30030 November 2023 Enclosure 3: Issuance - IP LAR for SE Renewed Facility Licenses and PDTS to Reflect Permanent Removal of Spent Fuel from SFPs ML23050A0022023-11-17017 November 2023 Enclosure 2 - Safety Evaluation for Indian Point Unit 2 License Amendment Request to Modify Technical Specifications for Staffing Requirements Following Spent Fuel Transfer to Dry Storage ML23100A1172023-11-17017 November 2023 NRC Response - Indian Point Energy Center Generating Units 1, 2, and 3 Letter with Enclosures Regarding Changes to Remove the Cyber Security Plan License Condition ML23064A0002023-11-13013 November 2023 NRC Issuance for Approval-Indian Point EC Units 1, 2 and 3 Emergency Plan and Emergency Action Level Scheme Amendments ML23067A0822023-11-0101 November 2023 Enclosure 2 - Issuance Indian Point Energy Center Generating Units 1, 2, and 3 Safety Exemption Evaluation for Certain Emergency Planning Requirements of 10 CFR 50.47 and Part 50 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML21091A3052022-02-28028 February 2022 Issuance of Amendment No. 272 Revision to Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device (EPID L-2020-LLA-0051) (Non-Proprietary) ML21074A0002021-04-22022 April 2021 Issuance of Amendment No. 270 Permanently Defueled Technical Specifications ML21083A0002021-04-14014 April 2021 Issuance of Amendment No. 63 Permanently Defueled Technical Specifications 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 ML20297A3332020-11-23023 November 2020 Enclosure 3, Safety Evaluation for Transfer of Renewed Facility Operating Licenses to Holtec International, Owner, and Holtec Decommissioning International, LLC, Operator ML20226A2722020-08-18018 August 2020 Request to Use a Provision of a Later Edition of the ASME BPV Code, Section XI 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 ML20100H9922020-06-0202 June 2020 Issuance of Amendment No. 269 Proposed Technical Specification Changes to City Water Surveillance Requirement and Condensate Storage Tank Required Action A.1 ML20122A2622020-05-0404 May 2020 Correction to Amendment No. 294 Dated April 28, 2020, Permanently Defueled Technical Specifications ML20081J4022020-04-28028 April 2020 Issuance of Amendment No. 294 Permanently Defueled Technical Specifications ML20078L1402020-04-15015 April 2020 Issuance of Amendment Nos. 62, 293, and 268 Changes to Emergency Plan for Post-Shutdown and Permanently Defueled Condition ML20099A1822020-04-13013 April 2020 Issuance of Relief Request IP3-IST-RR-001 - Alternative to Certain Requirements of the ASME Code for Extension of the Fourth 10-Year Inservice Test Interval ML20071Q7172020-04-10010 April 2020 Issuance of Amendment Nos. 292 and No. 267 Changes to Technical Specification Sections 1.1, 4.0, and 5.0 for a Permanently Defueled Condition ML19333B8682019-12-18018 December 2019 Approval of Certified Fuel Handler Training and Retraining Program 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 ML19175A0422019-09-11011 September 2019 Arkansas Units 1 and 2; Grand Gulf, Unit 1; Indian Point 2 and 3; Palisades; River Bend, Unit 1; Waterford Unit 3 - Issuance of Amendments to Adopt TSTF-529, Clarify Use and Application Rules ML19209C9662019-09-0404 September 2019 Issuance of Amendment No. 290 Storage of Fresh and Spent Nuclear Fuel in the Spent Fuel Pool ML19065A1012019-03-21021 March 2019 Issuance of Amendment No. 61 and No. 289 Deletion of License Conditions Related to Decommissioning Trust Provision ML19039A1492019-02-25025 February 2019 Issuance of Relief Request IP3-ISI-RR-14 Alternative Examination Required by ASME Code Case N-724-4 ML18337A4222018-12-20020 December 2018 Issuance of Amendment No. 265 One-Time Extension of 10 CFR Part 50, Appendix J, Type a, Integrated Leakage Rate Test Interval 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 ML18142A4312018-05-31031 May 2018 Safety Evaluation for Relief Request IP2-ISI-RR-06 Approval of Alternative to Use Embedded Weld Repair 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 ML17348A6952018-01-11011 January 2018 Issuance of Amendment Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank (CAC No. MF9578; EPID L-2017-LLA-0202) ML17320A3542017-12-22022 December 2017 Issuance of Amendments Amendment of Inter-Unit Transfer of Spent Fuel (CAC Nos. MF8991 and MF8992; EPID L-2016-LLA-0039) ML17315A0002017-12-0808 December 2017 Issuance of Amendments Cyber Security Plan Implementation Schedule (CAC Nos. MF9656, MF9657, and MF9658; EPID: L-2017-LLA-0217) ML17174B1442017-07-12012 July 2017 Relief Request for EN-ISI-16-1 Regarding Use of Later Edition and Addenda of the ASME Code ML17065A1712017-03-27027 March 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML17069A2832017-03-16016 March 2017 Relief Request No. IP3-ISI-RR-09, for Alternative to the Depth Sizing Qualification Requirement ML16336A4922017-01-27027 January 2017 Transmittal Letter: Order Approving Transfer of Master Decommissioning Trust Funds for Indian Point, No. 3 & FitzPatrick Nuclear Plant from the Power Authority of the State of New York to Entergy Nuclear Operations, Inc. 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 ML16215A2432016-11-15015 November 2016 Issuance of Amendment Nos. 285 and 261 Conditional Exemption from End-of-Life Moderator Temperature Coefficient ML16179A1782016-09-14014 September 2016 Safety Evaluation for Relief Request IP2-ISI-RR-01, Examination of Upper Pressurizer Welds ML16251A6202016-09-13013 September 2016 Entergy Fleet Request for Approval of Change to the Entergy Quality Assurance Program Manual (CAC Nos. MF7086 - MF7097) ML16167A0812016-07-15015 July 2016 Request for Alternative IP2-ISI-RR-03 to Weld Reference System Examination Required by ASME Code Subarticle IWA-2600 ML16147A5192016-07-14014 July 2016 Safety Evaluation for Relief Request IP2-ISI-RR-02 Alternative Examination Volume Required by Code Case N-729-1 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 ML16093A0282016-05-31031 May 2016 Entergy Services, Inc., Proposed Alternative to Utilize ASME Code Case N-789-1, Relief Request RR-EN-15-1, Revision 1 ML16064A2152016-04-12012 April 2016 Issuance of Amendments Cyber Security Plan Implementation Schedule 2023-05-01
[Table view] |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Vice President, Operations Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 March 16, 2017
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3-RELIEF REQUEST NO. IP3-ISl-RR-09 FOR ALTERNATIVE TO THE DEPTH SIZING QUALIFICATION REQUIREMENT (CAC NO. MF8896)
Dear Sir or Madam:
By letter dated December 2, 2016, as supplemented by letter dated January 26, 2017, Entergy Nuclear Operations, Inc. (the licensee), submitted Relief Request No. IP3-ISl-RR-09 to the U.S. Nuclear Regulatory Commission (NRC), requesting the use of an inspection procedure at Indian Point Nuclear Generating Unit No. 3 with a depth sizing error that is greater than the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and ASME Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," for the fourth 10-year inservice-inspection (ISi) interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee requested relief from the depth sizing uncertainty qualification requirement for ultrasonic examinations conducted from the inside diameter of pipes (i.e., root-mean square (RMS) error not greater than 0.125 inches) contained in ASME Code Cases N-695 and N-696. The licensee requested relief from the requirements for ISi items on the basis that the ASME Code requirement is impractical.
The NRC staff has concluded, as set forth in the enclosed safety evaluation, that using a vendor with a 0.189-inch RMS error for depth sizing for ASME Code Case N-695 and a 0.245-inch RMS error for depth sizing for ASME Code Case N-696, provides reasonable assurance of the structural integrity and leak-tightness in the subject welds. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff grants the licensee's Relief Request No. IP3-ISl-RR-09 at Indian Point Nuclear Generating Unit No. 3 for the fourth 10-year ISi interval, which began on July 21, 2009, and is currently scheduled to end on July 20, 2019. All other ASME Code,Section XI requirements for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear lnservice Inspector.
Vice President, Operations If you have any questions, please contact the Project Manager, Douglas Pickett, at 301-415-1364 or by e-mail at Douglas.Pickett@nrc.gov.
Docket No. 50-286
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, :-J ""' C/VVL-i.-_ I G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELIEF REQUEST NO. IP3-ISl-RR-09 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL ENTERGY NUCLEAR OPERATIONS, INC. INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated December 2, 2016, as supplemented by letter dated January 26, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 16350A104 and ML 17038A408, respectively), Entergy Nuclear Operations, Inc. (the licensee) submitted Relief Request No. IP3-ISl-RR-09 to the U.S. Nuclear Regulatory Commission (NRC or the Commission) requesting the use of an inspection procedure at Indian Point Nuclear Generating Unit No. 3 with a depth sizing error that is greater than the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and ASME Code Case N-696, "Qualification Requirements for Mandatory Appendix VIII Piping Examinations Conducted From the Inside Surface,Section XI, Division 1," for the fourth 10-year inservice-inspection (ISi) interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee requested relief from the depth sizing uncertainty qualification requirement for ultrasonic examinations conducted from the inside diameter (ID) of pipes (i.e., root-mean square (RMS) error not greater than 0.125 inches) contained in ASME Code Cases N-695 and N-696. The licensee requested relief from the requirements for ISi items on the basis that the ASME Code requirement is impractical.
2.0 REGULATORY EVALUATION
In its letter dated December 2, 2016, the licensee requested relief from the 0.125-inch RMS error depth sizing acceptance criteria contained in ASME Code Cases N-695 and N-696 pursuant to 10 CFR 50.55a(g)(5)(iii).
The following regulations were considered in the NRC staff's review:
Enclosure
- 10 CFR 50.55a(g)(4)(ii) states, in part, that: lnservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147).
- 1 O CFR 50.55a(g)(5)(iii) states, in part, that licensees may determine that conformance with certain Code requirements is impractical and that the licensee shall notify the Commission and submit information in support of the determination.
- 1 O CFR 50.55a(g)(6)(i) states, in part, that the Commission will evaluate determinations under paragraph (g)(5) of this section that Code requirements are impractical and that the Commission may grant such relief and may impose such alternative requirements as it determines are authorized by law and will not endanger life or property.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to grant, the relief requested by the licensee.
3.0 TECHNICAL
EVALUATION 3.1 The Licensee's Relief Request Component Descriptions The welds covered by Relief Request No. IP3-ISl-RR-09 are four reactor vessel hot leg and four cold leg nozzle to safe-end dissimilar metal (OM) welds and the eight corresponding safe-end to pipe elbow austenitic steel welds. Applicable Code Requirement The code of record for the fourth 10-year ISi interval is ASME Code,Section XI, 2001 Edition with 2003 Addenda. The hot and cold leg OM weld examination requirements are covered in ASME 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 XI, Division 1." ASME Section XI, Code Case N-770-1, Item A-2, "Unmitigated Hot Leg Butt Welds," specifies that a visual inspection is performed every refueling outage, and a volumetric examination is performed every 5 years. ASME Section XI, Code Case N-770-1, Item B, "Unmitigated Cold Leg Butt Welds," specifies that a visual examination is performed once per interval, and a volumetric examination must be performed every second inspection period, not to exceed 7 years. The safe-end to pipe elbow welds are covered under the risk-informed program. Risk-informed R-A requires volumetric examination of safe-end to pipe/elbow austenitic welds every 10 years. The welds will be examined from the inner diameter, requiring the use of ASME Code Cases N-695 and N-696, as ASME Code Section XI, 2001 Edition, with 2003 Addenda, has no rules for Appendix VIII qualifications for inner-diameter examinations.
Proposed Inspection The licensee proposes to use Code Cases N-695-1 and N-696-1, which allow a depth sizing RMS error of 0.25 inches instead of the 0.125 inches specified for depth sizing in Code Cases N-695 and N-696, and ASME Code,Section XI, Appendix VIII, Supplements 2 and 10. The examination vendor contracted to perform the safe end examinations has demonstrated the ability to depth size indications in OM welds with an RMS error of 0.189 inches for the reactor pressure vessel (RPV) nozzle to safe-end OM welds (Appendix VIII, Supplement
- 10) and 0.245 inches RMS error for the safe-end to pipe austenitic welds. If the examination vendor demonstrates an improved depth sizing RMS error prior to the examination, improved RMS error will be used in any flaw sizing instead of the 0.189-inch and 0.245-inch RMS error. If a reportable flaw is detected and determined to be ID surface connected during examination of the welds in accordance with this relief request, the licensee will provide a flaw evaluation, including the measured flaw size as determined by ultrasonic examination for NRC review. Eddy current testing will be used to determine if flaws are surface connected.
Additional data, including details of the surrounding ID surface contour in the region of the flaw and percentage of the examination area where ultrasonic testing (UT) probe lift-off is evident, if any, will be included.
In the event that any flaws requiring depth sizing are detected during examination of welds in accordance with this relief request, the following criteria shall be implemented:
- Flaws detected and measured as less than 50 percent through-wall in depth shall be adjusted by adding a correction factor to the flaw depth such that the adjusted flaw depth is equal to the measured flaw depth +(applicable vendor RMS error-0. 125 in.), prior to comparison to the applicable acceptance criteria;
- For flaws detected and measured as 50 percent through-wall depth or greater, and to remain in service without mitigation or repair, the licensee will submit flaw evaluations for review and approval prior to reactor startup. The flaw evaluation will include: 1. Information concerning the mechanism that caused the flaw. 2. Information concerning the inside surface roughness and/or profile of the region surrounding the flaw in the examined piping weld. 3. Information concerning areas where UT probe lift-off is observed, if any. Basis for the Request During the upcoming Spring 2017 refueling outage, the licensee will perform ultrasonic examination of the four hot leg safe-end to nozzle OM welds, and during the 2019 outage, the four cold leg safe-end to nozzle OM welds will be ultrasonically examined.
These examinations will be performed from the ID of the weld utilizing remote inspection equipment.
ASME Code Case N-695 will be used to establish the qualification requirements when performing OM weld inspections only, and ASME Code Case N-696 will be used to establish the qualification requirements for the inspectors when performing coordinated inspections of both DM and austenitic stainless steel welds. To date, the contracted vendor, who is qualified for detection and length sizing on these welds, has not met the RMS error requirement for depth sizing. The examination vendor has demonstrated ability to meet the depth sizing qualification requirement with an RMS error of 0.189 inches for the RPV nozzle to safe-end DM welds and 0.245 inches RMS error for the safe-end to pipe welds instead of the 0.125 inches required by the Code Case. Duration of the Proposed Relief Relief is requested for the fourth 10-year interval, which began on July 21, 2009, and is currently scheduled to end on July 20, 2019. 3.2 NRC Staff Evaluation The licensee will use NRC-approved Code Case N-695 to satisfy the requirements of ASME Code,Section XI, Appendix VIII, Supplement 10, and ASME Code Case N-696 to satisfy the requirements of ASME Code,Section XI, Appendix VIII, Supplement
- 2. Code Cases N-695 and N-696 require that procedures used to inspect welds from the inside surface of the pipe be qualified by performance demonstration.
The acceptance criterion in Code Cases N-695 and N-696 specify that the RMS error of the examination procedures shall not be greater than 0.125 inches. The licensee's inspection vendor was able to depth size with an RMS error of 0.189 inches. The licensee is requesting relief from the 0.125-inch depth sizing requirement in ASME Code Cases N-695 and N-696, in accordance with 10 CFR 50.55a(g)(5)(iii).
The NRC staff has confirmed that since 2002, the industry has not been able to satisfy the RMS error acceptance criterion of less than 0.125 inches when qualifying the volumetric examination inspection procedures performed from the inside surface of a pipe. Developing new technology capable of meeting the 0.125-inch RMS error, and qualifying the new technology to meet the requirements of ASME Code Cases N-695 and N-696, would be a burden on the licensee.
The staff concludes that this repeated inability to qualify inside surface UT inspection techniques in accordance with ASME Code Cases N-695 and N-696 constitutes an impracticality, as described in 10 CFR 50.55a(g)(5)(iii).
To address the issue of increased potential for undersizing of flaws by inside surface UT inspection procedures that do not meet ASME Code Cases N-695 and N-696 acceptance criterion, in 2012, the NRC staff, in conjunction with personnel from the Performance Demonstration Initiative (POI), examined the proprietary UT examination data set compiled from all attempts to date to qualify inside surface UT inspection procedures to the acceptance criterion contained in ASME Code Cases N-695 and N-696. Based on this examination, the NRC staff concluded that: (a) For flaw depths less than or equal to 50 percent pipe wall thickness, a flaw could be appropriately depth sized if a correction factor is added to the measured flaw depth such that the adjusted flaw depth is equal to the measured flaw depth plus the difference between the vendor procedure qualification RMS error and 0.125 inches. (b) For flaw depths greater than 50 percent wall thickness, the variability of sizing errors is sufficiently large so that no single mathematic flaw size adjustment formula is sufficient to provide reasonable assurance of appropriate flaw depth sizing. As a result, the NRC staff finds it necessary to evaluate the flaws that have depth greater than 50 percent through-wall on a case-by-case basis. To provide reasonable assurance of the structural integrity of examined welds, the NRC staff determined that the following compensatory measures shall be applied to any inspection not meeting the 0.125-inch RMS error for depth sizing to address the measurement uncertainty in flaw depth sizing when examining welds from the inside surface: (1) Examine the welds under consideration using a UT technique that is qualified for flaw detection and length sizing. (2) For flaws with a measured depth of less than 50 percent of the wall thickness, the depth shall be adjusted by adding the measured flaw depth to the difference between the procedure qualification RMS error and 0.125 inches. (3) For flaws with measured depth of greater than 50 percent of the wall thickness, either the degraded weld needs to be repaired in accordance with the ASME Code, or a flaw evaluation needs be submitted to the NRC staff for review and approval prior to reactor startup. (4) In addition to information normally contained in flaw evaluations performed in accordance with the ASME Code,Section XI, IWB-3600, the submitted flaw evaluation shall include (a) information concerning the degradation mechanism that caused the crack, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or weld, and (c) information concerning areas in which the UT probe may "lift off' from the surface of the pipe and/or weld. (5) Perform eddy current examinations to confirm whether a flaw is connected to the inside surface of the pipe and/or weld. The nozzle-to-safe-end and safe-end-to-pipe configurations differ from the POI mockups in two respects.
The weld surfaces were machined smooth, and the stainless steel safe-ends and welds were cladded on both inner and outer surfaces.
The NRC staff has no technical issues with the ID examinations of the wrought pipe to cast stainless steel elbow welds. As with the corrosion resistant clad, UT and followup ET inspections from the ID should be effective at detecting and length sizing flaws. Depth sizing is very challenging in cast stainless steels, but the proposed inspection procedure addresses this concern effectively.
The NRC staff concludes that the licensee's alternative is consistent with the compensatory measures discussed above, because (1) the licensee will add the correction factor to the crack tips, (2) the licensee will use eddy current testing to verify whether an embedded flaw is connected to the inside surface, and (3) the licensee will submit any flaw analysis for flaws greater than 50 percent through-wall to the NRC staff for review and approval prior to startup. Therefore, the NRC staff determines that relief from the depth sizing RMS error acceptance criterion of ASME Code Case N-695 and using a vendor with a 0.189-inch RMS error for depth sizing and ASME Code Case N-696 with a 0.245-inch RMS error for depth sizing provides reasonable assurance of the structural integrity and leak tightness of the subject welds.
4.0 CONCLUSION
As set forth above, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law, will not endanger life or property or the common defense and security, and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(5)(iii).
Therefore, the staff grants the licensee's Relief Request No. IP3-ISl-RR-09 at Indian Point Nuclear Generating Unit No. 3 for the fourth 10-year ISi interval, which began on July 21, 2009, and is scheduled to end on July 20, 2019. All other ASME Code,Section XI requirements for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor:
S. Cumblidge Date: March 16, 2017 Vice President, Operations
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3-RELIEF REQUEST IP3-ISl-RR-09 FOR ALTERNATIVE TO THE DEPTH SIZING QUALIFICATION REQUIREMENT (CAC NO. MF8896) DATED MARCH 16, 2017 DISTRIBUTION:
Public LPL 1 Reading File RidsRgn1 MailCenter RidsNrrDeEpnb RidsNrrDorlLpl 1 RidsNrrLALRonewicz TSetzer, R-1 SCumblidge, NRR ADAMS A ccess1on N b ML 17069A283 um er: OFFICE DORL/LPLl/PM DORL/LPLl/LA NAME DPickett LRonewicz DATE 03/13/2017 03/13/2017 RidsNrrPMlndianPoint RidsACRS_MailCTR JBowen, OEDO *b ., 1y e-ma1 DE/EPNB/BC*
DORL/LP LI/BC DAiley JDanna 03/09/2017 03/16/2017 OFFICIAL RECORD COPY