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MONTHYEARML18219A9212018-08-0707 August 2018 Acceptance Review Determination: Relief Request IP3-ISI-RR-14, Proposed Alternative to ASME Code Case N-729-4 Project stage: Acceptance Review ML19039A1492019-02-25025 February 2019 Issuance of Relief Request IP3-ISI-RR-14 Alternative Examination Required by ASME Code Case N-724-4 Project stage: Approval 2018-08-07
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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
MONTHYEARML24011A1982024-01-12012 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 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 ML23326A1322023-12-0505 December 2023 Issuance of Amendment No. 67, 300 & 276 to Implement the Independent Spent Fuel Storage Installation Only Emergency 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 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 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 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) 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 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 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 ML23194A0442023-07-11011 July 2023 Clarification for Indian Point Energy Center License Amendment Request, Independent Spent Fuel Storage Installation Physical Security Plan ML23192A1002023-07-11011 July 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise the Emergency Plan and Emergency Action Level Scheme ML23171B0432023-06-23023 June 2023 Letter - Indian Point Energy Center - Request for Additional Information for Independent Spent Fuel Storage Installation Facility-Only Emergency Plan License Amendment ML23118A0972023-06-0606 June 2023 06-06-23 Letter to the Honorable Michael V. Lawler, Et Al., from Chair Hanson Regarding Holtec'S Announcement to Expedite Plans to Release Over 500,000 Gallons of Radioactive Wastewater from Indian Point Energy Center Into the Hudson River ML23144A3512023-05-25025 May 2023 Clementina Bartolotta of Pearl River, New York Email Against Treated Water Release from Indian Point Site ML23144A3522023-05-25025 May 2023 Loredana Bidmead of New York E-Mail Against Treated Water Release from Indian Point Site ML23144A3412023-05-25025 May 2023 Dianne Schirripa of Rockland County, New York Email Against Treated Water Release from Indian Point Site ML23144A3472023-05-25025 May 2023 David Mart of Blauvelt, New York Email Against Treated Water Release from Indian Point Site ML23144A3402023-05-25025 May 2023 Melvin Israel of New York Email Against Treated Water Release from Indian Point Site ML23144A3542023-05-25025 May 2023 Terri Thal of New City, New York Email Against Treated Water Release from Indian Point Site ML23144A3532023-05-25025 May 2023 John Shaw of New York Email Against Treated Water Release from Indian Point Site 2024-01-09
[Table view] Category:Safety Evaluation
MONTHYEARML23326A1322023-12-0505 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 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 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 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 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
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 25, 2019 Vice President, Operations Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATIN.G UNIT NO. 2 -ISSUANCE OF RELIEF REQUEST IP3-ISI-RR-14 RE: ALTERNATIVE EXAMINATION REQUIRED BY ASME CODE CASE N-729-4 (EPID L-2018-LLR-0102)
Dear Sir or Madam:
By letter dated July 23, 2018 (Agencywide Documents Access and Management System Accession No. ML 18211A299), Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted Relief Request IP3-ISI-RR-14 to the U.S. Nuclear Regulatory Commission (NRC) for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). Entergy requested relief from the augmented inspection in accordance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-729-4, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI," as mandated by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D), with conditions.
In Relief Request IP3-ISI-RR-14, the licensee proposed to use the alternative requirements of Appendix I, "Analysis Procedure for Alternative Examination Area or Volume Definition," in Code Case N-729-4, as the use of Appendix I requires NRC prior approval.
Specifically, pursuant to 10 CFR 50.55a(z)(2), the licensee proposed alternative volumetric examination requirements in accordance with Appendix I of Code Case N-729-4 on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRG staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC staff authorizes the proposed alternative for the remainder of the fourth 10-year inservice inspection interval at Indian Point 3, which began on June 21, 2009, and is scheduled to end on July 20, 2020. All other ASME Code requirements for which relief was not specifically requested and approved remain applicable. If you have any questions concerning this matter, please contact the Indian Point 3 Project Manager, Mr. Richard Guzman, at (301) 415-1030 or Richard.Guzman@nrc.gov.
Docket No. 50-286
Enclosure:
Safety Evaluation cc: Listserv Sincerely, Plant Licensing Branch 1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST IP3-ISI-RR-14 REGARDING ASME CODE CASE N-729-4 ALTERNATIVE EXAMINATION ENTERGY NUCLEAR OPERATIONS, INC. INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated July 23, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 18211A299), Entergy Nuclear Operations, Inc. (Entergy or the licensee) requested relief from the augmented inspection in accordance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-729-4, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI," as mandated by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D), with conditions.
In Relief Request IP3-ISI-RR-14, the licensee proposed to use the alternative requirements of Appendix I, "Analysis Procedure for Alternative Examination Area or Volume Definition," in Code Case N-729-4, as the use of Appendix I requires U.S. Nuclear Regulatory Commission (NRC) prior approval.
This relief request is for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). Specifically, pursuant to 10 CFR 50.55a(z)(2), the licensee proposed alternative volumetric examination requirements in accordance with Appendix I of Code Case N-729-4 on the basis that compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
Components (including supports) that are classified as ASME Code Class 1, 2, and 3 must meet the requirements in 10 CFR 50.55a(g)(4), "lnservice inspection standards requirement for operating plants," throughout the service life of a boiling or pressurized water-reactor (PWR). Pursuant to 10 CFR 50.55a(g)(4)(ii), "Applicable ISi Code: Successive 120-month intervals," inservice 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 ASME Code incorporated by reference in paragraph (a) of 10 CFR 50.55a 12 months before the start of the 120-month inspection interval ( or the optional ASME Code Enclosure Cases listed in NRC Regulatory Guide (RG) 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," when using ASME Code,Section XI, as incorporated by reference in paragraph (a)(3)(ii) of 50.55a), subject to the conditions listed in paragraph (b) of 10 CFR 50.55a. Pursuant to 10 CFR 50.55a(g)(6)(ii)(D), "Augmented ISi requirements:
Reactor vessel head inspections," (1) "Implementation," holders of operating licenses or combined licenses for PWRs as of or after August 17, 2017, shall implement the requirements of ASME Code Case N-729-4. Pursuant to 10 CFR 50.55a(g)(6)(ii)(D)(2), "Appendix I use," Appendix I of ASME Code Case N-729-4 shall not be implemented without prior NRC approval.
Pursuant to 10 CFR 50.55a(z), "Alternatives to codes and standards requirements," alternatives to the requirements of paragraphs (b) through {h) of 10 CFR 50.55a, or portions thereof, may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate.
A proposed alternative must be submitted and authorized prior to implementation.
The applicant or licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements of 10 CFR 50.55a would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. 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 NRC to authorize, the alternative requested by the licensee.
3.0 3.1 TECHNICAL EVALUATION Background By letter dated October 1, 2010 (ADAMS Accession No. ML 102590213), the NRC approved similar Relief Request IP3-ISI-RR-04 for the fourth 10-year inservice inspection interval (ISi) of Indian Point 3, authorizing the licensee to use the alternative requirements of Appendix I in Code Case N-729-1. The regulations 10 CFR 50.55a(g)(6)(ii)(D) mandated the augmented inspection in accordance with Code Case N-729-1, with conditions.
The latest regulations in 10 CFR 50.55a(g)(6)(ii)(D) require the augmented inspection in accordance with Code Case N-729-4, instead of Code Case N-729-1, with conditions.
Thus, the licensee submitted IP3-ISI-RR-14 requesting to use the alternative requirements of Appendix I in Code Case N-729-4 for the remainder of the fourth 10-year ISi interval of Indian Point 3. 3.2 Components Affected ASME Code Class 1 reactor pressure vessel (RPV) closure head penetration nozzles and their partial penetration (J-groove) dissimilar metal attachment welds are affected.
In accordance with ASME Code Case N-729-4 {Table 1), these dissimilar metal welds are classified as Item No. B4.20. Below, the licensee identified the RPV closure head penetration nozzles that are affected by this relief request.
- Penetration Nos. 1 through 37
- Penetration Nos. 38, 41, 42, 43, 44, 46, 47, 48, 52, 55, 56, 61, 62, 65, 67, 68, 70, 71, 72, 73, 74, 75, 76, 77, and 78
- Penetration Nos. 40, 49, 50, 53, 54, 57, 58, 60, 64, and 66 The licensee stated that the RPV closure head penetration nozzles and associated J-groove welds at Indian Point 3 are made of Alloy 600/82/182 and known to be susceptible to primary water stress corrosion cracking. 3.3 Applicable Code Edition and Addenda The code of record for the fourth 10-year ISi interval is the 2001 Edition through 2003 Addenda of the ASME Code. 3.4 Duration of Relief Request The licensee submitted this request for remainder of the fourth 10-year ISi interval, which began on June 21, 2009, and is scheduled to end on July 20, 2020. 3.5 ASME Code Requirement ASME Code Case N-729-4 (Table 1, Item No. B4.20) requires that all RPV closure head penetration nozzles and their associated DM welds be subjected to volumetric and surface examinations every 8 calendar years, or before reinspection years (RIY = 2.25), whichever is less.
- Code Case N-729-4, Figure 2, "Examination Volume for Nozzle Base Metal and Examination Area for Weld and Nozzle Base Metal," identifies the required examination volume or area of tube to be inspected, including a distance "a" above the highest point of the root of the J-groove weld to a distance "a" below the lowest point of the toe of the J-groove weld. Distance "a" is equal to 1.5 inches for incidence angle less than or equal to 30 degrees to the horizontal plane, or 1.0 inch for incidence angle greater than 30 degrees to the horizontal plane.
- Code Case N-729-4 (Table 1, Note 6) requires the volumetric or surface examinations of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Code Case N-729-4 (Figure 2).
- Code Case N-729-4 (Table 1, Note 5) requires that if the examination area or volume requirements of Figure 2 cannot be met, the alternative requirements of Mandatory Appendix I in Code Case N-729-4 shall be used, and the evaluation shall be submitted to the NRC. 3.6 Proposed Alternative In lieu of examining the required nozzle volume in accordance with Code Case N-729-4, as discussed above, the licensee proposed to use the analysis procedure of Appendix I in Code Case N-729-4 to define the alternative examination volume, since the impediments prevented the examinations of the required volume below the toe of downhill side of the J-groove weld, as defined by Code Case N-729-4 (Figure 2). The analysis of Appendix I is also used to demonstrate the adequacy of the alternative examination volume of each nozzle. Based on the analysis of Appendix I, the licensee proposed minimum examination volume for each nozzle. The proposed alternative examination volume for the ultrasonic test (UT) is from a distance "a" above the root of the J-groove weld to a distance "Minimum Required UT Coverage" below the toe of downhill side of the J-groove weld, as defined in IP3-IS1-RR-14 (Table 1) below. IP3-1SI-RR-14 (Table 1): Minimum Inspection Coverage Requirement Nozzle Nozzle (1 lMinimum Required UT Time (EFPY) to Reach Penetration Angle of Coverage Below J-groove Weld the Lowest Point of the No. Incidence with> 2 Effective Full Power Toe of J-groove Weld (Degrees)
Years (EFPY) by Crack Growth (Years) Evaluation (Inches) 01 through 29 0.0 to 24.8 0.4 3.0 30 throuqh 37 26.2 0.4 2.7 38 through 69 30.2 to 38.6 0.4 2.7 70 through 73 44.3 0.3 3.0 7 4 throuqh 78 48.7 0.3 4.2 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 kilo-pound per square inch (ksi): 0.86 inch at nozzles 1 through 29; 0.50 inch at nozzles 30 through 69; 0.35 inch at nozzles 70 through 73; and 0.35 inch at nozzles 74 through 78. 3. 7 Basis for Use of Alternative The licensee stated that the design of RPV closure head penetration nozzles includes an approximately
- 0. 75 inch long threaded section at the bottom of the nozzles tube, as shown in Figure 1 (Attachment to NL-18-052) of IP3-IS1-RR-14.
There is no qualified volumetric inspection technique available to interrogate the threaded region at the nozzle end and provide meaningful results. Furthermore, 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 1.5 inch lower boundary limits defined by Code Case N-729-4 (Figure 2). The licensee stated that the inspection by the surface examination techniques is an option to meet the surface area requirement as defined by Code Case N-729-4 (Figure 2). However, the radiation dose rates under the RPV head near the J-groove weld areas are expected to be within 3 to 5 rem/hour.
Additionally, the area under the RPV head is posted as a locked high radiation and contamination area. Therefore, performance of the required surface examinations is considered a hardship.
The licensee stated that the proposed alternative will be used for the RPV closure head penetration nozzles as described below:
- Penetration Nos. 1 through 37 do not meet the 1.5 inch examination volume criterion specified in Code Case N-729-4 (Figure 2) for an incident angle of less than or equal to 30 degrees. *
- Penetration Nos. 38, 41, 42, 43, 44, 46, 47, 48, 52, 55, 56, 61, 62, 65, 67, 68, 70, 71, 72, 73, 74, 75, 76, 77, and 78 do not meet the 1 inch examination volume criterion specified in Code Case N-729-4 (Figure 2) for an incident angle of greater than 30 degrees.
- During previous Refueling Outages 3R14 to 3R19, the examination volume for Penetration Nos. 40, 49, 50, 53, 54, 57, 58, 60, 64, and 66 was measured to be within 0.080 inch of the examination volume criterion specified in Code Case N-729-4 (Figure 2). The licensee also requested relief for these penetrations due to the potential that the examination volume criterion cannot be met. The licensee stated that its justification for the adequacy of the proposed alternative examination volume is based on: (1) a finite element stress analysis of the nozzle and its attachment J-groove weld; (2) a postulation of an initial flaw size and crack growth rate based on the stress analysis; and (3) a demonstration that a postulated crack in the unexamined volume is not expected to grow to the toe of the J-groove weld prior to the next inspection, which is scheduled for the next refueling outage in approximately 2 years. The NRC previously reviewed and approved the licensee's supporting stress and flaw analyses used in developing the above Table 1.1 The licensee stated that it will perform the volumetric examination by the UT procedure and personnel qualified in accordance with Section 2500 requirements of Code Case N-729-4 for the flaw detection from the inside surface of the RPV head penetration nozzle. The scanning area extends from 1 inch and 1.5 inches above the J-groove weld (i.e., the upper boundary limits defined in Code Case N-729-4 (Figure 2)) and continues down the nozzle to at least the top of the threaded region. IP3-ISI-RR-14 (Table 1) provides the minimum inspection coverage required to ensure that a postulated axial through-wall flaw in the uninspected area of the RPV penetration nozzle will not propagate into the pressure boundary formed by the J-groove weld prior to a subsequent inspection or in 2 effective full power years (EFPY). The licensee stated that, for all RPV penetration nozzles, the requirements of Code Case N-729-4 (Figure 2) for the inspection of the examination volume above the J-groove weld will be met. The proposed alternative is applicable for the examination volume of the nozzles below the J-groove weld. 3.8 NRC Staff Evaluation The NRC staff has evaluated IP3-ISI-RR-14 pursuant to 10 CFR 50.55a(z)(2).
The NRC staff focuses on: (1) whether compliance with the specified requirements.
of 1 O CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety; and (2) that the licensee's proposed alternative (accepting Code Case N-729-4 (Appendix I) analysis procedure to define a reduced examination volume in this case) provides reasonable assurance of structural integrity and leaktightness of the RPV penetration nozzles. 1 See Entergy Relaxation Requests for Inspection of RPV Heads dated May 19, 2004 (ADAMS Accession No. ML041460199);
Relief Request IP3-ISI-RR-04 dated December 23, 2009 (ADAMS Accession Nos. ML 100050207);
and NRC letter dated October 1, 2010 (ADAMS Accession No. ML 102590213). Hardship Within the context of IP3-IS1-RR-14, the NRC staff has determined that the licensee identified the limitations that prevented the volumetric or surface inspections of the required examination volume or surface of the penetration nozzle tube below the toe of the J-groove weld, as defined by Code Case N-729-4 (Figure 2). The design of penetration nozzles tube has the threaded region of approximately
- 0. 75 inch long at the bottom end of the penetration nozzles. The NRC staff confirms that the volumetric examination of the threaded region is not possible with the current technologies.
The distance between the toe of the J-groove welds and the top of the threaded region of the nozzle is, in some cases, less than the inspection distance required by Code Case N-729-4 (Figure 2). The NRC staff confirms that the licensee could not meet the required examination volume criteria.
Although the inspection of the required surface area of the nozzle's threaded region by the surface examination techniques is an acceptable option, the length of time needed to perform the surface inspection would expose the involved personnel to unnecessary, high radiation dose rates. Therefore, the NRC staff determines that the nozzle tube design limitations, from an as low as reasonably achievable and safety hazards standpoint, constitute a hardship.
The NRC staff finds that the licensee has provided sufficient basis to demonstrate that performing the required examination constitutes a hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Proposed Alternative
-Assurance of Structural Integrity and Leaktightness In evaluating the licensee's proposed alternative, the NRC staff assessed the licensee's plant-specific stress analysis and crack growth calculation.
These analyses serve as a technical basis to demonstrate that the proposed inspection volume/area of nozzle tube below the toe of the J-groove weld is adequate, and any potential cracks that might exist in the uninspected volume/area of the nozzle tube would not grow to the pressure boundary formed by the J-groove weld prior to the next scheduled inspection (i.e., 2 EFPY). The NRC staff verified that:
- The licensee's plant-specific stress analysis was performed according to the ASME Code and industry standards to plot the distributions of the stresses along the length of the penetration nozzles below the J-groove weld. The operating stresses in the penetration nozzles were determined to be the highest at or close to the pressure boundary formed by the J-groove weld and decrease rapidly as the distance below the toe of the J-groove weld increases.
- The lowest point below the toe of the downhill side of the J-groove weld that experiences an operating stress level of 20 ksi was identified in each nozzle tube. This distance is 0.86 inch for nozzles 1 through 29; 0.50 inch for nozzles 30 through 69; 0.35 inch for nozzles 70 through 73; and 0.35 inch for nozzles 74 through 78, as documented in IP3-IS1-RR-14 (Table 1). The NRC staff notes that the crack initiation in the primary water stress corrosion cracking susceptible materials with 20 ksi or lower operational stresses theoretically has a reduced probability compared to areas at or very close to yield stress.
- A plant-specific crack growth calculation was performed according to the ASME Code and industry standards to demonstrate that any crack postulated in the uninspected volume/area below the toe of the J-groove weld would not grow to the pressure boundary formed by the J-groove weld before 2 EFPY.
- The postulated crack upper tip was assumed to lie where the proposed inspection volume/area ends, which is either 0.3 inch or 0.4 inch below the toe of downhill side of the J-groove weld at the applicable nozzle, as documented in IP3-ISI-RR-14 (Table 1).
- The time required for the postulated crack to grow from the uninspected volume/area to reach the pressure boundary at the toe of the J-groove weld was calculated to be within 2. 7 EFPY and 4.2-EFPY at the applicable nozzle, as documented in IP3-ISI-RR-14
{Table 1 ). However, the NRC staff notes that the licensee will inspect these nozzles every refueling outage, which is less than 2 EFPY. The NRC staff finds that the above-calculated time for a postulated axial through-wall flaw in the uninspected area of the RPV penetration nozzle reaching the pressure boundary is greater than 2 EFPY. Therefore, the licensee will be able to detect any unanticipated flaw growth prior to the pressure boundary being challenged.
This provides a reasonable assurance of structural integrity and leaktightness of the pressure boundary formed by the subject J-groove welds because the next inspection will be performed in 2 EFPY. In summary, the NRC staff concludes that the licensee's proposed alternative provides reasonable assurance of structural integrity, public health, and safety of the RPV closure head penetration nozzles. With these considerations, compliance with examination coverage requirements of Code Case N-729-4, as mandated by the regulations in 10 CFR 50.55a(g)(6)(ii)(D), with conditions, would result in hardship, without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leaktightness of the subject RPV closure head penetration nozzles and associated J-groove welds, and complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC staff authorizes the use of IP3-ISI-RR-14 at Indian Point 3 for the remainder of the fourth 10-year ISi interval, which is scheduled to end on July 20, 2020. All other ASME Code,Section XI requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor:
A. Rezai Date: February 25, 2019
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 -ISSUANCE OF RELIEF REQUEST IP3-ISI-RR-14 RE: ALTERNATIVE EXAMINATION REQUIRED BY ASME CODE CASE N-729-4 (EPID L-2018-LLR-0102)
DATED FEBRUARY 25, 2019 DISTRIBUTION:
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