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MONTHYEARML14342A0942014-12-17017 December 2014 Request for Additional Information Regarding Relief Request PRR-24, Nozzle-to-Vessel Welds and Nozzle Inner Radii Examinations Project stage: RAI ML15016A1152015-01-0808 January 2015 Response to Request for Additional Information Regarding Relief Request PRR-24, Nozzle-to-Vessel Welds and Nozzle Inner Radii Examinations Project stage: Response to RAI ML15084A0252015-03-23023 March 2015 Non-Proprietary - Safety Evaluation - Fourth 10-Year Interval Inservice Inspection - Request for Relief PRR-24 Project stage: Approval ML15103A0692015-04-21021 April 2015 Relief Request PRR-24 Regarding Nozzle-to-Vessel Welds and Nozzle Inner Radii Examination Project stage: Other 2015-01-08
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Category:Code Relief or Alternative
MONTHYEARML18128A0672018-06-0808 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 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 ML17093A8942017-05-12012 May 2017 Pilgrim Nuclear Power Station - Relief Requests PNPS-ISI-001 and PNPS-ISI-002, Relief from ASME Code Volumetric Examination Requirements for the Fourth 10-Year Inservice Inspection Interval (CAC Nos. MF8092 and MF8093) ML16257A5732016-09-15015 September 2016 Issuance of Relief Request No. PRR-52-Relief from ASME Code, Section XI Requirements for Pressure Testing of Class 1 Pressure Retaining Components as a Result of Repair/Replacement Activity and Use of Code Case N-795 ML16194A3262016-08-0404 August 2016 Issuance of Relief Request No. PRR-53 - Relief from ASME Code, Section XI Requirements for Ultrasonic Inspection Qualifications of Weld Overlays 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 ML16042A2912016-03-15015 March 2016 Issuance of Proposed Alternative Relief Request PRR-51, Relief from the Requirements of the ASME Code ML16057A1772016-02-16016 February 2016 Supplement to Request for Approval of Pilgrim Relief Request (PRR)-52, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Pressure Testing of Mechanical Joints as a Result of Performance of a Repair/Replacement Activity and Use of ML15338A3092016-01-0505 January 2016 Relief Request PRR-50, Relief from the Requirement of the ASME Code, Implementation of Code Case N-702 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 ML15103A0692015-04-21021 April 2015 Relief Request PRR-24 Regarding Nozzle-to-Vessel Welds and Nozzle Inner Radii Examination ML14198A1572014-08-0101 August 2014 Relief Requests PR-03 and PR-05 Regarding the Inservice Testing Program (Tac MF0370) ML12174A1472012-07-10010 July 2012 Safety Evaluation for Relief Request PRR-21, Rev 4, to Install a Weld Overlay on RPV-n14-1 Standby Liquid Control Safe-End Nozzle Weld at Pilgrim JAFP-11-0112, Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP2011-10-0303 October 2011 Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP ML0923705492009-09-11011 September 2009 Relief Request (PRR)19, Install a Weld Overlay on Jet Pump Instrumentation Nozzle Weld RPV-N9A-1- Pilgrim Nuclear Power Station ML0911304562009-04-30030 April 2009 Relief Request ISI-2008-1, Use of Later ASME Section XI Code Edition and Addenda for Repair and Replacement, Pressure Testing, and Destructive Testing Activities-Pilgrim Nuclear Power Station CNRO-2009-00003, Letter from Entergy Response to Request for Additional Information Regarding Request ISI-2008-1 Use of Later ASME Section Xl Code Edition and Addenda for Repair and Replacement, Pressure Testing, and Non Destructive Examination Activ2009-03-17017 March 2009 Letter from Entergy Response to Request for Additional Information Regarding Request ISI-2008-1 Use of Later ASME Section Xl Code Edition and Addenda for Repair and Replacement, Pressure Testing, and Non Destructive Examination Activities ML0813004082008-05-27027 May 2008 Relief Request No. RV-07, Alternate to the ASME OM Code 5-Year Test Interval for Main Steam Safety Relief Valves - Pilgrim Nuclear Power Station ML0805801992008-02-14014 February 2008 Request for Approval of Relief Request No. PRR-16, Implementation of BWRVIP Guidelines in Lieu of ASME Section XI Code Requirements on Reactor Vessel Internals and Components Inspection ML0724201612007-09-27027 September 2007 Relief for the Reactor Core Shroud Stabilizer Assemblies ML0618701632006-06-28028 June 2006 Request for Approval of ASME Code, Section XI, Third Ten-Year Relief Request, PRR-42, Examinations of Component Welds with Less than Essentially 100% Examination Coverage ML0606601322006-04-0505 April 2006 Relief Request No. PIL-05-R-002, ML0602400552006-03-22022 March 2006 Relief Request No. PRR-9, ML0601201272006-02-17017 February 2006 Relief Request No. PPR-05 ML0519902762005-08-29029 August 2005 Entergy Relief Request PR-03 High-Pressure Coolant Injection Pump ML0519201572005-06-29029 June 2005 Fourth Ten-Year Inservice Inspection Program Plan and the Associated Relief Requests for NRC Approval ML0429203582005-02-25025 February 2005 Request - Alternative Repair Plan for Generic Letter 88-01, Reactor Pressure Vessel Nozzle-To Cap Weld in the Control Rod Drive Return Line ML0423100082005-01-0606 January 2005 Relief Request No. PRR-29, Relief from System Hydrostatic Test Requirements for Small Bore ASME Code Class 1 Reactor Coolant Pressure Boundary Vent, Drain and Branch Lines and Connections 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 ML0429505512004-10-12012 October 2004 Response to NRC Request for Additional Information and Revised Relief Request, PRR-39, Rev. 1 ML0417401842004-07-0606 July 2004 Relief Request, Nos. RR-34 and PRR for the Third 10-Year Inservice Inspection (ISI) Interval, MC1999 and MC2006 ML0410700882004-07-0606 July 2004 Relief Request to Use American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-600 ML0407807052004-04-30030 April 2004 Relief Request, Fourth 10-Year Inservice Testing (IST) Program and Request for Approval of IST Relief Requests ML0408205162004-03-22022 March 2004 Relief Request Nos. RR-68, RR3-37, and PRR-34 (TAC MC1559, MC1560, & MC1561) ML0402600132004-02-26026 February 2004 Pilgrim Relief Request Review, Relief Request No. 38, Relief from ASME Code, Section XI, Appendix Viii, Supplement 11, Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds. JPN-03-020, Indian Point Nuclear Generating Station, Units 2 & 3, Pilgrim Nuclear Power Station, Vermont Yankee Nuclear Power Station, Relief, Relief Request to Use ASME Code Case N-6002003-08-11011 August 2003 Indian Point Nuclear Generating Station, Units 2 & 3, Pilgrim Nuclear Power Station, Vermont Yankee Nuclear Power Station, Relief, Relief Request to Use ASME Code Case N-600 ML0312701492003-05-0808 May 2003 Relief Request, Relief from ASME Code, Section XI, Appendix Viii, Supplement 10, Performance Demonstration for Ultrasonic Examination Systems ML0306402042003-04-11011 April 2003 Relief Request, Examinations of Reactor Pressure Vessel Circumferential Shell Welds ML0224002392002-09-17017 September 2002 Relief, Code Relief Request from Section XI, the Pump and Valve Inservice Testing Program Regarding Inclusion of Additional Excess Flow Check Valves ML0216800622002-06-0505 June 2002 Code Relief, Pilgrim Relief Request (PRR)-27 Relief from 1989 ASME Code Section XI Requirements for Certification of VT-2 Visual Examination Personnel 2018-06-08
[Table view] Category:Letter
MONTHYEARL-24-002, Late LLRW Shipment Investigation Report Pursuant to 10 CFR 20, Appendix G2024-02-0202 February 2024 Late LLRW Shipment Investigation Report Pursuant to 10 CFR 20, Appendix G ML23342A1182024-01-0909 January 2024 Independent Spent Fuel Storage Installation Security Inspection Plan L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 ML23334A1822023-11-30030 November 2023 Biennial Report for the Defueled Safety Analysis Report Update, Technical Specification Bases Changes, 10 CFR 50.59 Evaluation Summary, and Regulatory Commitment Change Summary November 2021 Through October 2023 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) L-23-011, 10 CFR 72.48 Biennial Change Summary Report2023-10-27027 October 2023 10 CFR 72.48 Biennial Change Summary Report IR 05000293/20234012023-08-31031 August 2023 NRC Inspection Report No. 05000293/2023401 & 2023001 (Cover Letter Only) IR 05000293/20230022023-08-0404 August 2023 NRC Inspection Report No. 05000293/2023002 L-23-008, Correction to Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations Holtec Decommissioning International, LLC (HDI)2023-05-23023 May 2023 Correction to Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations Holtec Decommissioning International, LLC (HDI) ML23136A7792023-05-15015 May 2023 Annual Radiological Environmental Operating Report, January 1 Through December 31, 2022 ML23135A2152023-05-15015 May 2023 Annual Radioactive Effluent Release Report, January 1 Through December 31, 2022 L-23-004, HDI Annual Occupational Radiation Exposure Data Reports - 20222023-04-24024 April 2023 HDI Annual Occupational Radiation Exposure Data Reports - 2022 L-23-003, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-31031 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations ML23088A0382023-03-29029 March 2023 Stations 1, 2, & 3, Palisades Nuclear Plant, and Big Rock Point - Nuclear Onsite Property Damage Insurance ML23069A2782023-03-13013 March 2023 Request for Scoping Comments Concerning the Environmental Review of Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application ML22361A1022023-02-24024 February 2023 Reactor Decommissioning Branch Project Management Changes for Some Decommissioning Facilities and Establishment of Backup Project Manager for All Decommissioning Facilities IR 05000293/20220042023-02-15015 February 2023 NRC Inspection Report No. 05000293/2022004 ML22356A0712023-01-31031 January 2023 Issuance of Exemption for Pilgrim Nuclear Power Station ISFSI Regarding Annual Radioactive Effluent Release Report - Cover Letter ML22347A2782022-12-21021 December 2022 Independent Spent Fuel Storage Installation Security Inspection Plan Dated December 21, 2022 L-22-042, Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.152022-12-14014 December 2022 Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.15 L-22-041, Supplemental Information to Enhance Exemption Request Detail for Pilgrim ISFSI Annual Radioactive Effluent Release Report Due Date Extension2022-12-0909 December 2022 Supplemental Information to Enhance Exemption Request Detail for Pilgrim ISFSI Annual Radioactive Effluent Release Report Due Date Extension IR 05000293/20220032022-11-18018 November 2022 NRC Inspection Report No. 05000293/2022003 L-22-036, Decommissioning Trust Fund Agreement2022-11-0808 November 2022 Decommissioning Trust Fund Agreement ML22276A1762022-10-24024 October 2022 Decommissioning International Proposed Revisions to the Quality Assurance Program Approval Forms for Radioactive Material Packages ML22266A1922022-09-23023 September 2022 and Pilgrim Nuclear Power Station - Request to Withdraw Prior Submissions from NRC Consideration ML22272A0352022-09-22022 September 2022 S. Lynch-Benttinen Letter Regarding U.S. Citizen Intent to Sue U.S. Fish and Wildlife and NOAA Fisheries Representing the Endangered Species (Na Right Whale) Which Will Be Adversely Affected by Holtec International Potential Actions ML22269A4202022-09-22022 September 2022 Citizen Lawsuit ML22241A1122022-08-29029 August 2022 Request for Exemption from 10 CFR 72.212(a)(2), (b)(2), (b)(3), (b)(4), (B)(5)(i), (b)(11), and 72.214 for Pilgrim ISFSI Annual Radioactive Effluent Release Report IR 05000293/20220022022-08-12012 August 2022 NRC Inspection Report No. 05000293/2022002 ML22215A1772022-08-0303 August 2022 Decommissioning International (HDI) Proposed Revisions to the Quality Assurance Program Approval Forms for Radioactive Material Packages ML22206A1512022-08-0101 August 2022 NRC Office of Investigations Case Nos. 1-2022-002 & 1-2022-006 ML22221A2592022-08-0101 August 2022 LTR-22-0217-1-NMSS - Town of Duxbury Letter Opposing the Irradiated Water Release from Pilgrim (Docket No. 05000293) ML22193A1662022-07-28028 July 2022 LTR-22-0154-1 - Heather Govern, VP, Clean Air and Water Program, Et Al., Letter Regarding Radioactive Wastewater Disposal from the Pilgrim Nuclear Power Station (Docket No. 05000293) ML22175A1732022-07-28028 July 2022 LTR-22-0153-1 - Response Letter to D. Turco, Cape Downwinders, from A. Roberts, NRC, Regarding Holtec-Pilgrim Plans to Dump One Million Gallons of Radioactive Waste Into Cape Cod Bay ML22154A4882022-06-0101 June 2022 Letter from Conservation Law Foundation Regarding Irradiated Water Release from Pilgrim ML22154A1622022-05-26026 May 2022 Letter and Email from Save Our Bay/Diane Turco Regarding Irradiated Water Release from Pilgrim ML22136A2602022-05-16016 May 2022 Submittal of Annual Radiological Environmental Operating Report for January 1 Through December 31, 2021 ML22136A2572022-05-16016 May 2022 Submittal of Annual Radioactive Effluent Release Report for January 1 Through December 31, 2021 ML22102A0932022-05-12012 May 2022 LTR-22-0067 Response to Matthew P. Levesque, President, Barnstable Town Council Regarding Irradiated Water Release from Pilgrim IR 05000293/20220012022-05-11011 May 2022 NRC Inspection Report No. 05000293/2022001 ML22104A0542022-04-30030 April 2022 LTR-22-0093 Response to Sheila Lynch-Benttinen, Regarding Irradiated Water Release from Pilgrim L-22-026, Occupational Radiation Exposure Data Report - 20212022-04-29029 April 2022 Occupational Radiation Exposure Data Report - 2021 ML22152A2592022-04-25025 April 2022 Zaccagnini Letter Dated 04/25/22 ML22152A2642022-04-19019 April 2022 Flynn Letter Dated 04/19/22 ML22091A1062022-04-0101 April 2022 Nuclear Onsite Property Damage Insurance (10 CFR 50.54(w)(3)) L-22-022, and Indian Point Nuclear Generating Stations 1, 2, & 3 - Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations - Holtec.2022-03-25025 March 2022 and Indian Point Nuclear Generating Stations 1, 2, & 3 - Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations - Holtec. ML22130A6762022-03-14014 March 2022 Decola Ltr Dtd 03/14/22 Re Potential Discharge of Radioactive Water from Pilgrim Nuclear Power Station ML22041B0762022-03-0707 March 2022 03-07-22 - Letter to the Honorable William R. Keating, Responds to Letter Regarding Proposed Release of Irradiated Water at Pilgrim Nuclear Power Station Into Cape Cod Bay ML22054A2962022-02-23023 February 2022 Annual ISFSI Radioactive Effluent Release Report for 2021 2024-02-02
[Table view] Category:Safety Evaluation
MONTHYEARML21217A1752021-08-0505 August 2021 Amendment No 255 Pilgrim Independent Spent Fuel Storage Installation (ISFSI) Only Physical Security Plan - Public Version ML20328A2972020-12-0101 December 2020 Amendment No. 253 - Non-Safeguards Version ML19276C4202020-01-0202 January 2020 Issuance of Amendment No. 252, Request to Remove Cyber Security Plan Requirements for the Permanently Defueled Condition ML19274C6742020-01-0202 January 2020 Issuance of Amendment No. 251, Revise Emergency Plan and Emergency Action Level Scheme to Address Permanently Defueled Condition ML19142A0432019-12-18018 December 2019 Letter and Safety Evaluation, Exemption to Allow Reduced Emergency Planning Requirements; Revise Radiological Emergency Response Plan Consistent with Permanently Defueled Reactor ML19235A0502019-08-27027 August 2019 Issuance of Amendment No. 249 Order Approving Direct Transfer of Renewed Facility Operating License and ISFSI General License and Conforming Amendment ML19170A2502019-08-22022 August 2019 Enclosure 3, Safety Evaluation for Direct and Indirect Transfer of Renewed Facility Operating License to Holtec Pilgrim, LLC, Owner and Holtec Decommissioning International, LLC, Operator (L-2018-LLO-0003) ML19122A1992019-06-11011 June 2019 Review of Spent Fuel Management Plan ML18284A3752018-11-30030 November 2018 Issuance of Amendment No. 248, Revise Site Emergency Plan for On-Shift and Emergency Response Organization Staffing to Address Permanently Defueled Condition 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 ML17066A1302017-07-10010 July 2017 Pilgrim Nuclear Power Station - Issuance of Amendment No. 246, Revise Administrative Controls Section of Technical Specifications to Change Staffing and Training Requirements for Permanently Defueled Condition (CAC No. MF9304) ML17093A8942017-05-12012 May 2017 Pilgrim Nuclear Power Station - Relief Requests PNPS-ISI-001 and PNPS-ISI-002, Relief from ASME Code Volumetric Examination Requirements for the Fourth 10-Year Inservice Inspection Interval (CAC Nos. MF8092 and MF8093) ML17058A3252017-04-12012 April 2017 Approval of Certified Fuel Handler Training and Retraining Program ML16250A2232016-10-28028 October 2016 Issuance of Amendments Proposed Changes to Emergency Plan to Revise Training for the on - Shift Chemistry Technician ML16251A6202016-09-13013 September 2016 Entergy Fleet Request for Approval of Change to the Entergy Quality Assurance Program Manual (CAC Nos. MF7086 - MF7097) ML16082A4602016-06-0606 June 2016 Issuance of Amendment Cyber Security Plan Implementation Schedule 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 ML16124B0872016-05-0303 May 2016 E-mail Review of Safety Evaluation for Proposed Alternative Relief Request No. PRR-51 ML16042A2912016-03-15015 March 2016 Issuance of Proposed Alternative Relief Request PRR-51, Relief from the Requirements of the ASME Code ML16008B0772016-03-0303 March 2016 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML15338A3092016-01-0505 January 2016 Relief Request PRR-50, Relief from the Requirement of the ASME Code, Implementation of Code Case N-702 ML15166A4012015-06-19019 June 2015 Request for Alternative PRR-26 for the Fifth 10-year Inservice Inspection Interval ML15114A0212015-05-0606 May 2015 Issuance of Amendment Regarding the Minimum Critical Power Ratio License Amendment Request ML15103A0692015-04-21021 April 2015 Relief Request PRR-24 Regarding Nozzle-to-Vessel Welds and Nozzle Inner Radii Examination ML15084A0252015-03-23023 March 2015 Non-Proprietary - Safety Evaluation - Fourth 10-Year Interval Inservice Inspection - Request for Relief PRR-24 ML14272A0702015-03-12012 March 2015 Issuance of Amendment 242 Re Revision to Technical Specification 2.1, Safety Limits to Resolve Pressure Regulator Fail-Open Transient ML14336A6612014-12-11011 December 2014 Issuance of Amendment Regarding Cyber Security Plan Implementation Schedule Milestone 8 (Tac No. MF3482) ML14295A6852014-10-31031 October 2014 Issuance of Amendment Regarding Heavy Loads to Facilitate Dry Storage Handling Operations ML14210A2662014-08-0808 August 2014 Arkansas, Units 1 & 2, Big Rock Point, James A. Fitzpatrick, Grand Gulf, Unit 1, Indian Point, Units 1, 2 & 3, Palisades, Pilgrim, River Bend, Unit 1, Vermont Yankee, Waterford, Safety Evaluation Quality Assurance Program Manual, Rev. 24 & ML14198A1572014-08-0101 August 2014 Relief Requests PR-03 and PR-05 Regarding the Inservice Testing Program (Tac MF0370) ML14083A6312014-03-27027 March 2014 Relief Request PRR-22 Regarding a Risk-Informed Inservice Inspection Program for Class 1 and 2 Piping Welds ML13149A2152013-05-29029 May 2013 Correction to Staff Assessment Letter to Recommendation 9.3 of the Near-Term Task Force Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML13127A1792013-05-21021 May 2013 Staff Assessment in Response to Recommendation 9.3 of the Near-Term Task Force Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML13025A3062013-03-14014 March 2013 ANO 1 & 2, Big Rock, FitzPatrick, GGNS, Indian Point 1, 2 & 3, Palisades, Pilgrim, RBS, Vermont Yankee, and Waterford - Correction to Amendments Issued on 12/28/12, Revise QA Program Manual and Staff Qualification Technical Specifications ML12261A1302012-11-13013 November 2012 Issuance of Amendments to Renewed Facility Operating License Changes in Cyber Security Plan Implementation Milestone ML1220100482012-08-0707 August 2012 Issuance of Amendment No. 237 Revision to Condensate Storage Tank Low Level Trip Setpoint ML12174A1472012-07-10010 July 2012 Safety Evaluation for Relief Request PRR-21, Rev 4, to Install a Weld Overlay on RPV-n14-1 Standby Liquid Control Safe-End Nozzle Weld at Pilgrim ML11152A0432011-07-22022 July 2011 License Amendment, Cyber Security Plan ML1106500092011-03-28028 March 2011 Issuance of Amendment No. 235 Revised Technical Specifications for Setpoint and Setpoint Tolerance Increases for Safety Relief Valves and Spring Safety Valves ML1100502982011-01-26026 January 2011 Issuance of Amendment Regarding Revised Pressure and Temperature (P-T) Limit Curves and Relocation of P-T Curves to the PTLR ML1022901632010-08-25025 August 2010 Relief Request PRR-20, Alternative Examination Requirements for Nozzle-To-Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108 - Pilgrim Nuclear Power Station ML0936208072010-01-22022 January 2010 Cover Letter, (Non-Proprietary) Order Extending the Effectiveness of the Approval of the Indirect Transfer of Facility Operating Licenses for Big Rock Point, Fitzpatrick, Indian Point, Palisades, Pilgrim, and Vermont Yankee Nuclear Power St ML0936208952010-01-22022 January 2010 Safety Evaluation,(Non-Proprietary) Order Extending the Effectiveness of the Approval of the Indirect Transfer of Facility Operating Licenses for Big Rock Point, Fitzpatrick, Indian Point,Palisades,Pilgrim, and Vermont Yankee Nuclear Power ML0928706472009-10-29029 October 2009 Request for Threshold Determination Under 10 CFR 50.80-Big Rock Point, James A. FitzPatrick Nuclear Power Plant, Indian Point Nuclear Generating Nos. 1, 2 and 3, Palisades, Pilgrim, Vermont Yankee ML0927301372009-10-0909 October 2009 Safety Evaluation by the Office of Nuclear Reactor Regulation Corporate Restructuring Conversion of Companies and Stock Split-Off by Entergy Nuclear Operations, Inc and Subsidiaries ML0923705492009-09-11011 September 2009 Relief Request (PRR)19, Install a Weld Overlay on Jet Pump Instrumentation Nozzle Weld RPV-N9A-1- Pilgrim Nuclear Power Station ML0911304562009-04-30030 April 2009 Relief Request ISI-2008-1, Use of Later ASME Section XI Code Edition and Addenda for Repair and Replacement, Pressure Testing, and Destructive Testing Activities-Pilgrim Nuclear Power Station ML0906402242009-03-26026 March 2009 License Amendment, Revised Technical Specifications (TS) Section 2.1.2, Safety Limit Minimum Critical Power Ratio (SLMCPR) for Two-Loop and Single-Loop Operation ML0815703662008-11-20020 November 2008 License Amendment, Issuance of Amendment Adoption of TSTF-448, Revision 3, Control Room Envelope Habitability 2021-08-05
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. John A. Dent, Jr. Site Vice President Entergy Nuclear Operations, Inc. Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508 April 21, 2015
SUBJECT:
PILGRIM NUCLEAR POWER PLANT -RELIEF REQUEST PRR-24 REGARDING NOZZLE-TO-VESSEL WELDS AND NOZZLE INNER RADII EXAMINATIONS (TAC NO. MF4187)
Dear Mr. Dent:
By letter dated March 12, 2014, as supplemented by letter dated January 8, 2015, Entergy Nuclear Operations, Inc. (the licensee), submitted Relief Request PRR-24 for authorization of a proposed alternative to certain requirements of the American Society of Mechanical Engineer Boiler and Pressure Vessel Code (ASME Code),Section XI, Examination Category B-D for Pilgrim Nuclear Power Station (Pilgrim).
Specifically, the licensee proposed to use ASME Code Case N-702 which requires examination of a minimum of 25 percent of the nozzle-to-vessel welds and inner radius sections.
Pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) Section 50.55a(a)(3)(i) 1, the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed PRR-24 and concluded that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff has concluded that the licensee addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1
). Therefore, pursuant to 10 CFR 50.55a(z)(1
), the NRC staff has authorized the licensee's proposed alternative, as described in PRR-24, for the remainder of Pilgrim's fourth inservice inspection interval, projected to end on June 30, 2015. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.
1 Retitled Section 50.55a(z)(1), as noticed in the Federal Register on November 5, 2014 (79 FR 65776).
J. Dent If you have any questions, please contact the Pilgrim Project Manager, Nadiyah Morgan, at (301) 415-1016.
Docket No. 50-293
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Michael I. Dudek, Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST PRR-24 REGARDING NOZZLE-TO-VESSEL WELDS AND NOZZLE INNER RADII EXAMINATIONS ENTERGY NUCLEAR OPERATIONS, INC. PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
By letter dated March 12, 2014 (Agencywide Document Access and Management System (ADAMS) Accession No. ML 14077A175), as supplemented by letter dated January 8, 2015 (ADAMS Accession No. ML 15016A 115), Entergy Nuclear Operations, Inc. (the licensee), submitted Relief Request PRR-24 for authorization of a proposed alternative to certain requirements of the American Society of Mechanical Engineer Boiler and Pressure Vessel Code (ASME Code),Section XI, Examination Category B-D for Pilgrim Nuclear Power Station (Pilgrim).
Specifically, the licensee proposed to use ASME Code Case N-702, which requires examination of a minimum of 25 percent of the nozzle-to-vessel welds and inner radius sections.
Pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) Section 50.55a(a)(3)(i) 1 , the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The technical basis for ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactors (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," was documented in the Electric Power Research Institute (EPRI) Technical Report (TR) 1021005, "BWRVIP-241:
BWR Vessel and Internals Project [BWRVIP], Probabilistic Fracture Mechanics
[PFM] Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii." In a safety evaluation dated April 19, 2013 (ADAMS Accession No. ML 13071A240), the U.S. Nuclear Regulatory Commission (NRC) approved the BWRVIP-241 report, which identified plant specific requirements that must be met for applicants proposing to use this alternative.
1 The paragraph headings in 10 CFR 50.55(a) were changed, as noticed in the Federal Register (FR) on November 5, 2014 (79 FR 65776), which became effective on December 5, 2014. See the cross-reference tables, which are cited in the notice and available in ADAMS under Accession Nos. ML 14015A191 and ML 14211A050.
Enclosure
2.0 REGULATORY EVALUATION
The regulation at 10 CFR 50.55a(g)(4) states, in part, that ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
lnservice inspection (ISi) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code, and applicable addenda, as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 1 O CFR 50.55a(g)(6)(i).
The regulations at 10 CFR 50.55a(z) state, in part, that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The regulations require that ISi of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval and subject to the limitations and modifications listed therein. The applicable ASME Code of record for the fourth 10-year ISi interval at Pilgrim is the 1998 Edition through the 2000 Addenda of the ASME Code Section XI. For all reactor pressure vessel (RPV) nozzle-to-vessel shell welds and nozzle inner radii, ASME Code,Section XI, requires 100 percent inspection during each 10-year ISi interval.
However, ASME Code Case N-702 proposes an alternative which reduces the inspection of RPV to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval.
Based on the above, and subject to the technical evaluation which follows, the NRC staff finds that regulatory authority exists to authorize an alternative to items 83. 90 and 823.100 of ASME Code,Section XI, Examination Category B-D, as requested by the licensee.
3.0 TECHNICAL
EVALUATION
3.1 Relief
Request PRR-24 In accordance with 10 CFR 50.55a(z)(1), the licensee proposed an alternative to ASME required volumetric examinations for the ASME Code, Class 1, RPV nozzle-to-shell welds and nozzle inner radius sections listed below in Table 1. The proposed alternative reduces the ASME Code-required volumetric examinations for all RPV nozzle-to-shell welds and inner radii, from a 100 percent to a minimum of 25 percent of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size during each inspection interval.
This alternative is contained in ASME Code Case N-702. The required examination volume for the reduced set of nozzles remains at 100 percent of that depicted in Figures IWB-2500-7 (a) through (d), as applicable.
The licensee stated that, "the twenty-five percent sampling level stated in [ASME] Code Case N-702 provides a significant cost savings and reduction in worker dose exposure." 83.90 RPV-N2A-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N28-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2C-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2D-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2E-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2F-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2G-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2H-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2J-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.90 RPV-N2K-NV 12" Recirculation Inlet Nozzle-to-Vessel Weld 83.100 RPV-N2A-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N28-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2C-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2D-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2E-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2F-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2G-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2H-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2J-NIR 12" Recirculation Inlet Nozzle Inner Radius 83.100 RPV-N2K-NIR 12" Recirculation Inlet Nozzle Inner Radius For three of the reactor recirculation nozzle assemblies (Group N2, recirculation inlet, total number -10, minimum number to be examined -3), both the inner radius region and the nozzle-to-shell weld have already been examined during the fourth interval, three were completed in Refueling Outage 17 in 2009). [The] BWRVIP-241
[report] was developed to propose a relaxation of the criteria in BWRVIP-108:
Boiling Water Reactor Vessel and Internals Project (BWRVIP) Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii allowing BWR's to obtain inspection relief for their Reactor Recirculation inlet and outlet nozzles. The evaluation found that failure probabilities due to a low temperature overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e.,< 1x10-6 for 40 years) with or without inservice inspection.
The report concludes that inspection of 25 percent of each nozzle type is technically justified. The NRC staff's April 19, 2013, safety evaluation states that each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the request for alternative by meeting the criteria discussed in Section 5 of the safety evaluation.
As stated in the PRR-24, the applicability of the BWRVIP-241 report is displayed below by demonstrating Pilgrim's conformance with the three criteria applicable to the recirculation inlet nozzles and inner radius sections.
Criterion 1: The maximum RPV heatup/cooldown rate is less than 115 degrees Fahrenheit
(°F)/hour Therefore, Pilgrim having a maximum RPV Heatup/Cooldown rate that is limited to less than or equal to 100 °F/hour < 115 °F/hour. Criterion 2: For recirculation inlet nozzles, (pr/t)/CRPV
< 1.15 Where: p = RPV normal operating pressure, p = 1035 pounds per square inch (psi) r = RPV inner radius, r = 113.40625 inches t = RPV wall thickness, t = 6.5 inches CRPv =recirculation inlet nozzles (from BWRVIP-108 model)= 19332 psi Therefore, (p*r/t)/CRPV
= 0. 93 < 1.15 Criterion 3: For recirculation inlet nozzles, [p(r 0 2 + ri 2)/ (r 0 2 -ri 2)]/CNozzLE
< 1.47 Where: riN2 = inner radius for Recirculation Inlet N2 nozzles = 5. 75 inches roN2 =outer radius for Recirculation Inlet N2 nozzles = 9.125 inches CiNozzLE = recirculation inlet nozzles (from BWRVIP-108 model) = 1637 psi Therefore, [p(r o 2 + ri 2)/ (r 0 2 -ri 2)]/CNOZZLE
= 1.465 < 1.4 7 3.2 NRC Staff Evaluation By letter dated December 19, 2007 (ADAMS Accession No. ML073600374), the NRC issued a safety evaluation on the acceptability of BWRVIP-108, which specified five plant-specific criteria that licensees must meet in order to demonstrate that BWRVI P-108 results apply to their plants. The five criteria are related to the driving force of the PFM analysis for the recirculation inlet and outlet nozzles. In the safety evaluation, it was stated that the nozzle material fracture toughness-related values used in the PFM analyses were based on data from the entire fleet of BWR RPVs. Therefore, the BWRVIP-108 PFM analyses are bounding with respect to fracture resistance, and only the driving force of the underlying PFM analyses needs to be evaluated.
It was also stated that except for the RPV heat-up/cool-down rate, the plant-specific criteria are for the recirculation inlet and outlet nozzles only because the probabilities of failure for other nozzles are an order of magnitude lower. The NRC staff's April 19, 2013, safety evaluation, related to the BWRVIP-241 report, revised Criterion
- 3. The BWRVIP performed additional PFM analyses in the BWRVIP-241 report using the bounding recirculation inlet and outlet nozzles instead of the typical recirculation inlet and outlet nozzles of the BWRVI P-108 report. The BWRVI P's additional PFM analyses demonstrated that the limits can be higher than 1.15 and the corresponding probability of failures are still below 5 x 10-6/yr. Criterion 3 was modified to be less than or equal to 1.47 and Criterion 5 was modified to be less than or equal to 1.59. The NRC found that these changes result in probabilities of failure that are at least two orders of magnitude lower than the NRC safety goal of 5 x 1 o-6/yr for the pressurized thermal shock concern. As stated, the PFM results in the BWRVIP-241 report are best considered as a supplement to those in the BWRVIP-108 report, not a replacement.
However, it should be made clear that the conditions and limitations specified in Section 5.0 of the April 19, 2013, safety evaluation supersede those of the December 19, 2007, safety evaluation for the BWRVIP-108 report. The licensee stated that Criterion 1 is satisfied because Pilgrim maintains a maximum up/cool-down rate of 100 °F/hour, well below the 115 °F/hour criterion limit. The licensee stated that in accordance with their Technical Specification 3.6.A.2, reactor coolant system heat-up and cool-down rates are limited to a maximum of 100 °F/hour when averaged over any 1-hour period. This addressed whether there have been any events during which the down rate was in excess of 115 °F/hour. This is not a concern as Criterion 1 refers only to normal operations, not typical transients.
For Criterion 2 and 3, the licensee provided and confirmed, in PRR-24 and its January 8, 2015, supplement, Pilgrim's plant-specific data evaluation of the driving force factors, or ratios, against the criteria established in the NRC staff's April 19, 2013, safety evaluation.
The licensee's calculated results showed that Criterion 2 and 3 are satisfied, and the NRC staff confirmed the accuracy of the calculations by performing the calculations independently with the provided radius and thickness values. The licensee noted that ASME Code Case N-702 stipulates that the VT-1 visual examination method may be used in lieu of the volumetric examination method for the inner radius sections.
Despite this allowance, all examination of nozzle inner radii of the selected recirculation inlet nozzles will be volumetric examinations.
The licensee has no intention to use ASME Code Case N-648-1, "Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1." Finally, the licensee indicated that 1 O recordable indications had been detected, three on the recirculation inlet nozzle-to-shell weld RPV-N2G-NV, four on the recirculation inlet nozzle-to-shell weld RPV-N2H-NV, and three on the recirculation inlet nozzle-to-shell weld RPV-N2K-NV.
In all cases, the indications were found to be acceptable per ASME Code,Section XI, IWB-3000.
Based on the above evaluation, the licensee meets Criterion 1, 2, and 3, as specified in the NRC staff's April 19, 2013, safety evaluation.
This plant-specific evaluation forms the technical basis for accepting the proposed alternative specified in ASME Code Case N-702.
4.0 CONCLUSION
As set forth above, the NRC staff has concluded that the proposed alternative provides an acceptable 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)(1
). Therefore, pursuant to 10 CFR 50.55a(z)(1
), the NRC staff authorizes the licensee's proposed alternative, as described PRR-24, for the remainder of Pilgrim's fourth ISi interval, projected to end on June 30, 2015. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor:
T. Mclellan Date: Apr i 1 21 , 201 5 J. Dent If you have any questions, please contact the Pilgrim Project Manager, Nadiyah Morgan, at (301) 415-1016.
Docket No. 50-293
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC LPLl-1 R/F RidsNrrDorlLpl1-1 RidsNrrLAKGoldstein RidsNrrPMPilgrim NrrRidsDorlDpr RidsNrrDeEvib RidsAcrsAcnwMailCenter RidsOgcMailCenter RidsRgn1 MailCenter TMcLellan, NRR Sincerely, IRA/ Michael I. Dudek, Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ADAMS Accession No. ML 15103A069
- See dated memo OFFICE LPLl-1/PM LPLl-1/LA DE/EVIB/BC LPLl-1/BC (A) NMorgan KGoldstein SRosenberg*
MDudek NAME (MHenderson for) DATE 4/16/2015 4/16/2015 3/23/2015 4/21/2015 OFFICIAL RECORD COPY