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Category:Code Relief or Alternative
MONTHYEARML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines ML21299A0032021-10-28028 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 ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 ML20099D9552020-04-17017 April 2020 Request to Use Provisions in the 2013 Edition of the ASME Boiler and Pressure Vessel Code for Performing Non-Destructive Examinations ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19135A4442019-06-21021 June 2019 Issuance of Relief Request I4R-22 Relief from the Requirements of the ASME Code ML19161A2572019-06-0404 June 2019 BWR Fleet Msv/Srv - Testing Frequency Relief Request NRC Pre-Application Meeting June 4, 2019 ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins JAFP-19-0023, Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds2019-02-15015 February 2019 Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds JAFP-18-0076, End of Interval Relief Request Associated with the Fourth Ten-Year Lnservice Inspection (ISI) Interval2018-07-26026 July 2018 End of Interval Relief Request Associated with the Fourth Ten-Year Lnservice Inspection (ISI) Interval ML18039A8542018-05-30030 May 2018 Relief Requests I5R-02, I5R-03, I5R-04 from ASME Code Requirements for Reactor Vessel Internals, Ferretic Piping Repair/Replacement, and RPV Flange Welds, 5th 10-Year Inservice Inspection Interval (MG0116-MG0118; L-2017-LLR-0083 to 0085) JAFP-18-0053, Proposed Alternative to Utilize Code Cases N-878 and N-8802018-05-30030 May 2018 Proposed Alternative to Utilize Code Cases N-878 and N-880 JAFP-18-0052, Proposed Alternative to Utilize Code Case N-8792018-05-30030 May 2018 Proposed Alternative to Utilize Code Case N-879 ML18044A9932018-04-13013 April 2018 Request for Alternatives PRR-01, PRR-02, PRR-04, VRR-02, VRR-03, and VRR-04 from ASME OM Code Requirements for Various Pumps and Valves, Fifth 10-Year Inservice Testing Interval (CAC MG0052-MG0061; EPID L-2017-LLR-0067 to L-2017-LLR-0074) ML17289A0752017-12-12012 December 2017 Issuance of Relief Request for Proposed Alternative to Use ASME Code Case N-789-1 (CAC No. MF9692; EPID L-2017-LLR-0027) Note Correction Safety Evaluation See ML18003B382 ML17219A4282017-12-11011 December 2017 Issuance of Relief Request-Alternative to Certain Requirements of the ASME Code Regarding Use of ASME Code Case N-513-4 (CAC No. MF9641; EPID L-2017-LLR-0023) ML17223A2802017-08-10010 August 2017 Submittal of Relief Requests Associated with the Fifth Lnservice Inspection (ISI) Interval ML17090A1682017-04-12012 April 2017 Alternative to ASME Code Requirements for Weld Overlay Repair ML16355A4292017-01-0606 January 2017 Relief Request for Proposed Alternative for the Implementation of BWRVIP-05 ML16334A4402016-12-0606 December 2016 Relief from the Requirements of the ASME Code Case N-702 and BWRVIP-241 for Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii ML16270A0462016-10-0303 October 2016 Acceptance of Requested Licensing Action Relief Request for Proposed Alternative for the Implementation of BWRVIP-05 ML16253A3412016-09-14014 September 2016 Acceptance of Requested Licensing Action Relief Request for Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 ML16180A2892016-06-29029 June 2016 Inservice Inspection Program Alternative for Safety Relief Valves 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 ML16077A3522016-03-22022 March 2016 Withdrawal of Relief Request No. 19 from the Fourth Inservice Inspection Interval JAFP-15-0122, Withdrawal of Application for Alternative Examination Requirements for James A. FitzPatrick Nuclear Power Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-2412015-11-20020 November 2015 Withdrawal of Application for Alternative Examination Requirements for James A. FitzPatrick Nuclear Power Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 ML15230A3502015-08-18018 August 2015 J.A Fitzpatrick Nuclear Power Plant - Requests Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1), Implementation of BWRVIP-05 (GL 98-05) 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 ML12279A2482012-10-17017 October 2012 Issuance of Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code 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 ML0803902232008-03-13013 March 2008 Relief Request No. 5, Use of Performance Demonstration Initiative in Lieu of ASME Code Section XI, Appendix Viii, Supplement 11 Requirement ML0803003072008-02-28028 February 2008 Relief Request No. RR-6, Implementation of BWRVIP Guidelines in Lieu of ASME Section XI Code Requirements on Reactor Vessel Internals Components Inspection ML0803700802008-02-25025 February 2008 Relief Request No. 2 (RR-2) from the Requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Appendix Viii, Supplement 10 ML0804204272008-02-22022 February 2008 Relief Request No. 3 (RR-3) Risk-Informed Inservice Inservice Inspection Program ML0520700472005-08-0909 August 2005 Relief Request for Temporary Non-Code Repair of a Shutdown Cooling Pipe JAFP-05-0105, Request for Approval of Relief Request No. RR-38, Proposed Alternative to Perform a Temporary Non-Code Repair in Accordance with 10 CFR 50.55a(a)(3)(ii)2005-07-0909 July 2005 Request for Approval of Relief Request No. RR-38, Proposed Alternative to Perform a Temporary Non-Code Repair in Accordance with 10 CFR 50.55a(a)(3)(ii) 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 ML0420301572004-07-20020 July 2004 Relief, Relief Request No. 30 for Third 10-Year Inservice Inspection (ISI) Program Interval ML0410700882004-07-0606 July 2004 Relief Request to Use American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-600 ML0417401842004-07-0606 July 2004 Relief Request, Nos. RR-34 and PRR for the Third 10-Year Inservice Inspection (ISI) Interval, MC1999 and MC2006 ML0405406932004-04-12012 April 2004 Relief Request Review, Relief Request VRR-08 Related to the Third 10-Year Inservice Testing (IST) Ubtervak 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 JAFP-03-0111, Proposed Alternatives in Accordance with 10CFR50.55a(g)(6)(ii)(A)(5) and Relief from ASME Section XI Code Regarding Inspection of RPV Vertical Shell Welds Pursuant to 10 CFR 50.55a (g)(6)(i)2003-08-0404 August 2003 Proposed Alternatives in Accordance with 10CFR50.55a(g)(6)(ii)(A)(5) and Relief from ASME Section XI Code Regarding Inspection of RPV Vertical Shell Welds Pursuant to 10 CFR 50.55a (g)(6)(i) ML0306502552003-04-0101 April 2003 Relief Request Review, Third 10-Year Pump and Valve Inservice Testing Program, Revision of Relief Request VRR-04 ML0231804962002-11-14014 November 2002 Relief, Request for Relief No. RR-28 for the Third 10-Year Inservice Inspection Interval Program Plan for the FitzPatrick Power Plant JAFP-02-0194, Proposed Revision of Relief Request VRR-06 for In-Service Testing Program2002-09-30030 September 2002 Proposed Revision of Relief Request VRR-06 for In-Service Testing Program JPN-02-011, Request for Relief RR-29, Third 10-Year Inservice Inspection Interval Program Plan2002-05-0808 May 2002 Request for Relief RR-29, Third 10-Year Inservice Inspection Interval Program Plan JPN-02-010, Relief Request RR-28, Revision 1 for the Third 10-Year Inservice Inspection Interval Program Plan2002-05-0808 May 2002 Relief Request RR-28, Revision 1 for the Third 10-Year Inservice Inspection Interval Program Plan 2023-12-14
[Table view] Category:Letter
MONTHYEARIR 05000333/20230042024-02-0707 February 2024 Integrated Inspection Report 05000333/2023004 and Independent Spent Fuel Storage Installation Inspection Report 07200012/2023001 ML24037A0102024-02-0606 February 2024 Requalification Program Inspection ML24018A0012024-01-18018 January 2024 Notification of Commercial Grade Dedication Inspection (05000333/2024010) and Request for Information ML24004A2302024-01-0808 January 2024 Project Manager Reassignment ML23356A0832024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0058 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines JAFP-23-0065, License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications2023-12-14014 December 2023 License Amendment Request to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4, and Administrative Changes to the Technical Specifications IR 05000333/20234012023-12-0808 December 2023 Cybersecurity Inspection Report 05000333/2023401 (Cover Letter Only) RS-23-126, Request for Exemption from 10 CFR 2.109(b)2023-12-0707 December 2023 Request for Exemption from 10 CFR 2.109(b) JAFP-23-0069, Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements JAFP-23-0057, and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 and Independent Spent Fuel Storage Installation - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation JAFP-23-0064, Emergency Plan Document Revision2023-11-15015 November 2023 Emergency Plan Document Revision JAFP-23-0063, Registration of Spent Fuel Cask Use2023-11-13013 November 2023 Registration of Spent Fuel Cask Use IR 05000333/20230032023-11-13013 November 2023 Integrated Inspection Report 05000333/2023003 ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums IR 05000333/20230102023-10-26026 October 2023 Biennial Problem Identification and Resolution Inspection Report 05000333/2023010 JAFP-23-0059, Registration of Spent Fuel Cask Use2023-10-24024 October 2023 Registration of Spent Fuel Cask Use IR 05000333/20233012023-10-19019 October 2023 Initial Operator Licensing Examination Report 05000333/2023301 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans JAFP-23-0050, Physical Security Plan, Revision 242023-08-31031 August 2023 Physical Security Plan, Revision 24 JAFP-23-0048, Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-31031 August 2023 Supplemental Information for License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis IR 05000333/20230052023-08-31031 August 2023 Updated Inspection Plan for James A. FitzPatrick Nuclear Power Plant (Report 05000333/2023005) JAFP-23-0047, Correction to the 2022 Annual Radioactive Effluent Release Report2023-08-30030 August 2023 Correction to the 2022 Annual Radioactive Effluent Release Report ML23228A1342023-08-16016 August 2023 Licensed Operator Positive Fitness-For-Duty Test IR 05000333/20230022023-08-0707 August 2023 Integrated Inspection Report 05000333/2023002 RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis2023-08-0303 August 2023 License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis JAFP-23-0043, 10 CFR 50.46 Annual Report2023-07-31031 July 2023 10 CFR 50.46 Annual Report JAFP-23-0038, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance2023-07-28028 July 2023 License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance ML23208A1622023-07-27027 July 2023 Operator Licensing Examination Approval IR 05000333/20234022023-07-26026 July 2023 Security Baseline Inspection Report 05000333/2023402 IR 05000333/20230112023-07-25025 July 2023 Post-Approval Site Inspection for License Renewal - Phase 4 Inspection Report 05000333/2023011 IR 05000333/20235012023-07-20020 July 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000333/2023501 JAFP-23-0033, License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation2023-06-28028 June 2023 License Amendment Request - Technical Specifications (TS) Section 3.3.1.2, Source Range Monitors (SRM) Instrumentation IR 05000333/20234202023-06-26026 June 2023 Security Baseline Inspection Report 05000333 2023420 RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23164A0322023-06-13013 June 2023 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000333/2023010 ML23152A0042023-06-0101 June 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection 05000333/2023401 RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling JAFP-23-0025, 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023 2022 Annual Radiological Environmental Operating Report IR 05000333/20230012023-05-0303 May 2023 Integrated Inspection Report 05000333/2023001 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI JAFP-23-0023, 2022 Annual Radioactive Effluent Release Report2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report IR 05000333/20230122023-04-13013 April 2023 Quadrennial Fire Protection Inspection Report 05000333/2023012 ML23095A3722023-04-0505 April 2023 2023 Updated Final Safety Analysis Report, Technical Specification Bases and Technical Requirements Manual Changes Transmittal RS-23-049, Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-23023 March 2023 Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations IR 05000333/20220042023-03-20020 March 2023 Integrated Inspection Report 05000333/2022004 JAFP-23-0010, 2022 REIRS Transmittal of NRC Form 52023-03-20020 March 2023 2022 REIRS Transmittal of NRC Form 5 ML23061A1632023-03-0303 March 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch 3 2024-02-07
[Table view] Category:Safety Evaluation
MONTHYEARML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML22223A1412022-09-0101 September 2022 Issuance of Amendment No. 353 Adoption of TSTF - 505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B ML22196A0612022-08-23023 August 2022 Issuance of Amendment No. 352 Adoption of 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors ML22166A4302022-07-15015 July 2022 Issuance of Amendment No. 351 Removal of Selected Response Time Testing for Reactor Protection System and Primary Containment Isolation Instrumentation ML22126A1962022-05-27027 May 2022 Issuance of Amendment No. 350 Adoption of TSTF-264, Revision 0 ML22090A0862022-04-29029 April 2022 Amendments to Adopt TSTF-541,Rev.2,Add Exceptions to Surveillance Requirements for Valves,Dampers Locked in Actuated Position ML22094A0012022-04-15015 April 2022 Constellation Energy Generation, LLC - Proposed Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles (Epids L-2021-LLR-0057 and L-2021-LLR-0058) ML21364A0432022-02-28028 February 2022 Issuance of Amendment No. 348 Revising Surveillance Requirement 3.5.1.6 Involving Recirculation Pump Discharge Valves ML21347A0382022-01-13013 January 2022 Issuance of Amendments to Revise Reactor Coolant Leakage Requirements ML21277A2482021-11-16016 November 2021 Letter with Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (Public Version) ML21300A3552021-11-16016 November 2021 Issuance of Amendment No. 345 Adoption of TSTF-582 ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21131A1272021-08-0909 August 2021 Issuance of Amendment No. 343 Modifications to Technical Specification 3.6.1.3, Primary Containment Isolation Valves (Pcivs) ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21166A1682021-06-25025 June 2021 ML21162A0422021-06-14014 June 2021 Issuance of Amendment No. 342, One Time Extension of Completion Times to Support Residual Heat Removal Pump Motor Replacement (Emergency Circumstances) ML21049A3552021-04-28028 April 2021 Issuance of Amendment No. 341 Adoption of TSTF-478, Revision 2, BWR Technical Specification Changes That Implement the Revised Rule for Combustible Gas Control ML21033A5302021-04-0101 April 2021 Issuance of Amendments to Adopt Technical Specifications Task Force TSTF-566, Revise Actions for Inoperable RHR Shutdown Cooling Subsystems ML21028A6732021-02-0303 February 2021 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L-2020-LLR-011 ML20287A1302020-11-0505 November 2020 Review of Quality Assurance Program Changes ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 ML20169A5102020-08-20020 August 2020 Issuance of Amendment No. 339 Changes to Technical Specifications Related to Primary Containment Hydrodynamic Loads ML20140A0702020-07-21021 July 2020 Issuance of Amendment No. 338 Application of Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences ML20141L6362020-07-10010 July 2020 Issuance of Amendments Based on TSTF-427,Allowance for Nontechnical Specification Barrier Degradation on Supported System Operability,Rev 2 ML20134H9402020-07-0808 July 2020 Issuance of Amendments Revising the High Radiation Area Administrative Controls ML20094G9032020-06-0202 June 2020 Issuance of Amendment No. 335 Adoption of TSTF-372, Addition of LCO 3.0.8, 'Inoperability of Snubbers' ML20021A0702020-04-0606 April 2020 Issuance of Amendments to Delete License Conditions for Decommissioning Trusts ML20024C6612020-03-0202 March 2020 Issuance of Amendment No. 332 Adopt TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, Using the Consolidated Line Item Improvement Process ML19295G7832019-12-19019 December 2019 Issuance of Amendment No. 331 Regarding Change to Technical Specifications to Remove Ultimate Heat Sink Bar Rack Heaters ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques 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 ML19176A0332019-08-28028 August 2019 Issuance of Amendments to Adopt TSTF-564, Safety Limit MCPR ML19189A0842019-08-19019 August 2019 Issuance of Amendment No. 326 Adoption of TSTF-522, Revision 0, Revise Ventilation System Surveillance Requirements to Operate for 10 Hours Per Month ML19192A2442019-07-18018 July 2019 Proposed Alternative to Use ASME Code Cases N-878 and N-880 ML19157A2032019-07-11011 July 2019 Issuance of Amendment No. 325 Reactivity Anomalies Surveillance ML19135A4442019-06-21021 June 2019 Issuance of Relief Request I4R-22 Relief from the Requirements of the ASME Code ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins ML18360A6352019-02-25025 February 2019 Safety Evaluation Regarding Implementation of Hardened Containment Vents Capable of Operation Under Severe Accident Conditions Related to Order EA-13-109 (CAC No. MF4464; EPID No. L-2014-JLD-0049) ML18304A3652019-01-16016 January 2019 2. Issuance of Amendments to Revise the Average Power Range Monitor Requirements ML18289A4322018-11-28028 November 2018 Issuance of Amendment No. 323 Revision to the Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 ML18214A7062018-09-19019 September 2018 Issuance of Amendment No. 322, Revise Technical Specification 2.1.1, Reactor Core Sls, to Change Cycle 24 Safety Limit Minimum Critical Power Ratio Numeric Values ML18206A2822018-08-0202 August 2018 Issuance of Amendments to Relocate the Staff Qualification Requirements ML18180A3722018-07-19019 July 2018 Issuance of Amendment No. 319, Revise Technical Specification Surveillance Requirement 3.6.4.1.3 to Allow Opening of Inner and Outer Secondary Containment Access Openings (CAC MG0239; EPID L-2017-LLA-0298) ML18039A8542018-05-30030 May 2018 Relief Requests I5R-02, I5R-03, I5R-04 from ASME Code Requirements for Reactor Vessel Internals, Ferretic Piping Repair/Replacement, and RPV Flange Welds, 5th 10-Year Inservice Inspection Interval (MG0116-MG0118; L-2017-LLR-0083 to 0085) ML18044A9932018-04-13013 April 2018 Request for Alternatives PRR-01, PRR-02, PRR-04, VRR-02, VRR-03, and VRR-04 from ASME OM Code Requirements for Various Pumps and Valves, Fifth 10-Year Inservice Testing Interval (CAC MG0052-MG0061; EPID L-2017-LLR-0067 to L-2017-LLR-0074) ML17289A1752018-03-26026 March 2018 Issuance of Amendment No. 318, Revise Emergency Plan for Emergency Response Organization Requalification Training Frequency Consistent with Exelon Fleet (CAC MG0026; EPID L-2017-LLA-0273) ML18003B3822018-01-0303 January 2018 Correction to the Safety Evaluation to James A. Fitzpatrick Nuclear Power Plant Issuance of Relief Request for Proposed Alternative to Use ASME Code Case N-789-1 (CAC No. MF9692; EPID L-2017-LLR-0027) (Original Safety Evaluation: ML17289A07 ML17342A0062017-12-18018 December 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML17289A0752017-12-12012 December 2017 Issuance of Relief Request for Proposed Alternative to Use ASME Code Case N-789-1 (CAC No. MF9692; EPID L-2017-LLR-0027) Note Correction Safety Evaluation See ML18003B382 2023-05-01
[Table view] |
Text
August 9, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE (TAC NO. MC7544)
Dear Mr. Kansler:
By letter dated July 9, 2005, Entergy Nuclear Operations, Inc. (Entergy) submitted a relief request, which proposed a temporary repair to a shutdown cooling suction pipe in the Residual Heat Removal (RHR) system. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. Entergy requested relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, subsection IWC-4000. The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a(a)(3)(ii). On July 10, 2005, the NRC staff granted verbal authorization.
As documented in the enclosed safety evaluation, the NRC staff reviewed your submittal and concluded that an ASME Code repair under the existing plant conditions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the temporary non-code repair was authorized.
The NRC also notes that on July 12, 2005, the temporary non-code repair was replaced by an ASME Code repair.
If you have any questions regarding this matter, please contact John Boska, the NRC project manager for FitzPatrick, at 301-415-2901.
Sincerely,
/RA/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
As stated cc w/encl: See next page
August 9, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE (TAC NO. MC7544)
Dear Mr. Kansler:
By letter dated July 9, 2005, Entergy Nuclear Operations, Inc. (Entergy) submitted a relief request, which proposed a temporary repair to a shutdown cooling suction pipe in the Residual Heat Removal (RHR) system. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. Entergy requested relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, subsection IWC-4000. The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a(a)(3)(ii). On July 10, 2005, the NRC staff granted verbal authorization.
As documented in the enclosed safety evaluation, the NRC staff reviewed your submittal and concluded that an ASME Code repair under the existing plant conditions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the temporary non-code repair was authorized.
The NRC also notes that on July 12, 2005, the temporary non-code repair was replaced by an ASME Code repair.
If you have any questions regarding this matter, please contact John Boska, the NRC project manager for FitzPatrick, at 301-415-2901.
Sincerely,
/RA/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC RLaufer JUhle CHolden PDI-1 R/F JBoska TMarsh OGC ACRS TChan WBateman SLittle MMitchell BMcDermott, RI JTsao Accession Number: ML052070047 *Safety Evaluation provided OFFICE PDI-1\PM PDI-1\LA EMCB\SC* OGC PDI-1\SC NAME JBoska SLittle TChan TColburn for RLaufer DATE 8/01/05 8/01/05 7/22/05 8/05/05 8/09/05 Official Record Copy
FitzPatrick Nuclear Power Plant cc:
Mr. Gary J. Taylor Resident Inspector's Office Chief Executive Officer James A. FitzPatrick Nuclear Power Plant Entergy Operations, Inc. U. S. Nuclear Regulatory Commission 1340 Echelon Parkway P.O. Box 136 Jackson, MS 39213 Lycoming, NY 13093 Mr. John T. Herron Ms. Charlene D. Faison Sr. VP and Chief Operating Officer Manager, Licensing Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.
440 Hamilton Avenue 440 Hamilton Avenue White Plains, NY 10601 White Plains, NY 10601 Mr. Theodore A. Sullivan Mr. Michael J. Colomb Site Vice President Director of Oversight Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant 440 Hamilton Avenue P.O. Box 110 White Plains, NY 10601 Lycoming, NY 13093 Mr. David Wallace Mr. Kevin J. Mulligan Director, Nuclear Safety Assurance General Manager, Plant Operations Entergy Nuclear Operations, Inc.
Entergy Nuclear Operations, Inc. James A. FitzPatrick Nuclear Power Plant James A. FitzPatrick Nuclear Power Plant P.O. Box 110 P.O. Box 110 Lycoming, NY 13093 Lycoming, NY 13093 Mr. Richard Plasse Mr. Oscar Limpias Manager, Regulatory Compliance Vice President Engineering Entergy Nuclear Operations, Inc.
Entergy Nuclear Operations, Inc. James A. FitzPatrick Nuclear Power Plant 440 Hamilton Avenue P.O. Box 110 White Plains, NY 10601 Lycoming, NY 13093 Mr. Christopher Schwarz Supervisor Vice President, Operations Support Town of Scriba Entergy Nuclear Operations, Inc. Route 8, Box 382 440 Hamilton Avenue Oswego, NY 13126 White Plains, NY 10601 Mr. Charles Donaldson, Esquire Mr. John F. McCann Assistant Attorney General Director, Licensing New York Department of Law Entergy Nuclear Operations, Inc. 120 Broadway 440 Hamilton Avenue New York, NY 10271 White Plains, NY 10601
FitzPatrick Nuclear Power Plant cc:
Regional Administrator, Region I Ms. Stacey Lousteau U.S. Nuclear Regulatory Commission Treasury Department 475 Allendale Road Entergy Services, Inc.
King of Prussia, PA 19406 639 Loyola Avenue Mail Stop L-ENT-15E Oswego County Administrator New Orleans, LA 70113 Mr. Steven Lyman 46 East Bridge Street Ms. Deb Katz, Executive Director Oswego, NY 13126 Nuclear Security Coalition c/o Citizens Awareness Network Mr. Peter R. Smith, President P.O. Box 83 New York State Energy, Research, Shelburne Falls, MA 01370 and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Mr. Travis C. McCullough Assistant General Counsel Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. James H. Sniezek BWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490 Mr. Michael D. Lyster BWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
1.0 INTRODUCTION
Entergy Nuclear Operations, Inc. (Entergy or the licensee), the licensee for the James A.
FitzPatrick Nuclear Power Plant (JAF), was planning a repair to a through-wall crack in the shutdown cooling suction pipe in the Residual Heat Removal (RHR) system in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, subsection IWC-4000. Entergy determined that an ASME Code repair was not possible unless the pipe was drained of water. JAF was in cold shutdown and the shutdown cooling suction pipe was being used for decay heat removal from the reactor fuel. The licensee stated that implementing an alternative method of decay heat removal in cold shutdown to allow draining the pipe would present significant hardship and a challenge to safe plant operation without a compensating increase in the level of quality and safety.
By letter dated July 9, 2005, Entergy submitted a relief request which proposed a temporary repair, which did not fully conform to the ASME Code, to the shutdown cooling suction pipe.
The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.55a(a)(3)(ii). Entergy requested relief from the ASME Code,Section XI, subsection IWC-4000. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. On July 10, 2005, the NRC staff granted verbal authorization for the temporary non-code repair. This evaluation addresses the merits of the requested relief pursuant to 10 CFR 50.55a(a)(3)(ii).
2.0 REGULATORY EVALUATION
The inservice inspection of the ASME Code, Class 1, Class 2, and Class 3 components in nuclear plants is to be performed in accordance with the ASME Code,Section XI, and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Paragraph 10 CFR 50.55a(a)(3) states: "Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f),
(g), and (h) of this section or portions thereof may be used when authorized by the Director of Enclosure
the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that: (i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
Pursuant to 10 CFR 50.55a(g)(4), ASME Code, Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination 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 incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The inservice inspection ASME Code of record for FitzPatricks current 10-year inservice inspection interval is the 1989 edition of the ASME Code,Section XI, no addenda.
3.0 TECHNICAL EVALUATION
3.1 ASME Code Component Affected The licensee requested relief for the common suction header portion of the shutdown cooling piping within the RHR system. It is classified as ASME Class 2, Examination Category C-H, Item C7.30 and C7.40, as shown in Table IWC-2500-1 of the 1989 edition of the ASME Code,Section XI.
3.2 Applicable ASME Code Edition and Addenda The ASME Code of Record for the FitzPatrick Repair/Replacement Program is the 1989 edition, no addenda, of the ASME Code,Section XI.
3.3 Applicable ASME Code Requirement The licensee requested relief from the requirements of Article IWC-4000 of the 1989 edition of the ASME Code,Section XI, which states that the rules of IWA-4000 apply. Article IWA-4000 specifies ASME Code-acceptable repair methods for service-related piping flaws that exceed ASME Code acceptable limits. An ASME Code repair is required to restore the structural integrity of the flawed ASME Code piping, independent of the operational mode of the plant when the flaw is detected.
3.4 Licensees Reason for Request The licensee stated that the plant is in Mode 4, Cold Shutdown, with the RHR system lined up in the shutdown cooling mode. Although the shutdown cooling suction line is degraded and the system has been declared inoperable, it is currently performing the required decay heat removal function. A temporary support has been installed nearby the crack area to reduce the static loading on the affected pipe support. In order to perform an ASME Code repair, the
shutdown cooling suction line would need to be removed from service, thus requiring use of an alternative means of decay heat removal.
The licensee stated that the most preferable alternative means of heat removal in cold shutdown is to use the main steam line drains and the main condenser. Using this method, reactor vessel level is raised to the main steam line elevation and reactor coolant water flows to the main condenser via the main steam line drains. The reactor coolant is cooled by circulating water pumped through the main condenser tubes and returned to the reactor vessel via the condensate and feed water systems. This alternative method presents the fewest challenges because it uses the normal feedwater connections and has little impact on reactor water chemistry. However, this method will not be effective because based on the licensees calculations, it can only remove about half the current decay heat being produced by the reactor fuel assemblies.
Besides the above cooling method, the licensee stated that as required by the plant Technical Specifications (TSs) alternate means of decay heat removal are available if needed to maintain the plant in the safe shutdown condition. However, the use of these alternate methods of decay heat removal presents various levels of operational challenges and hardships.
The first alternative method is to use the core spray system and the low pressure coolant injection system in conjunction with the safety relief valves. This alternative method would effectively remove the decay heat, and is credited in the accident analysis of the Updated Final Safety Analysis Report for this purpose. In this alternative method, the RHR system is lined up in the low pressure coolant injection mode and suction is taken from the torus. Torus water is injected into the reactor vessel and returned to the torus for cooling via the safety relief valves.
However, this method is undesirable due to the upset in normal reactor water chemistry caused by the introduction of torus water into the reactor vessel. This flow path may also introduce particulates from corrosion (commonly known as crud) into the safety relief valves which may result in damage to the valve seats. Industry operating experience demonstrates that the service life of the safety relief valves may be negatively affected with this method. Due to the high flow rate associated with the low pressure coolant injection system, this method also presents a challenge to the plant operators in controlling the heat-up and cooldown rates.
The second alternative is to use the feedwater and condensate systems in conjunction with the safety relief valves to provide a flow of cool water through the reactor vessel. This method has less impact on reactor water chemistry than the previously described method using torus water, but the potential impact on the safety relief valves is the same. In addition, use of this method would require processing and discharging thousands of gallons of liquid radwaste.
The third alternative is to use the decay heat removal system connected to the spent fuel pool.
This method requires removing the reactor vessel head and flooding the refueling cavity, which provides a path for water to flow from the vessel to the spent fuel pool. Use of this method would require extended use of the shutdown cooling system in its degraded condition until the reactor vessel could be flooded up, the vessel head, vessel steam dryer and steam separator removed, the refueling cavity flooded, and conditions for the start-up of decay heat removal established. These conditions include installation of a non-safety related diesel. The disadvantage of this alternative is that it would incur radioactive dose of about 10 to 15 person-rem for reactor disassembly and reassembly.
The licensee concluded that the above alternatives present challenges and hardships due to the impact on equipment and plant operations without any gain in safety or quality, whereas the proposed non-ASME Code repair allows for expedited restoration of the piping integrity with no operational hardships.
3.5 Proposed Alternative and Basis for Use The proposed non-ASME Code repair consists of stress relieving the crack and applying a seal weld to the crack surface. In addition, a temporary pipe support has been installed within 5 feet of the affected pipe support to reduce the static loading on the crack. The licensee has evaluated the adequacy of the affected pipe support, assuming no weld in the repair area, to continue to act as a seismic support and found it to be acceptable for all loading conditions.
After the repair, the licensee will visually inspect the temporary non-ASME Code repair area once a day while the shutdown cooling line is in operation to ensure the leak tightness of the weld and piping is maintained.
The licensee has analyzed the non-ASME Code weld repair, including all loading conditions to demonstrate the operability of the piping in a degraded condition. This will support plant mode changes upon completion of the torus repair and pressure testing. The licensee will then proceed with plant heatup into Mode 3 (Hot Shutdown), thus allowing the shutdown cooling system to be removed from service. Decay heat removal will be accomplished by steaming through the main steam lines to the main condenser. An ASME Code repair in accordance with the requirements of IWC-4000 will be performed when this section of the shutdown cooling line can be removed from service and prior to entering Mode 2 (Startup). Pressure test of the ASME Code repair will be performed in accordance with IWC-5000 requirements.
3.6 Duration of Proposed Alternative The licensee requested this non-ASME Code repair to be a one-time, temporary relief from the requirements of IWC-4000. The non-ASME Code repair would be immaterial as soon as the ASME Code repair is performed on the crack. It is anticipated that the ASME Code repair would be completed a few days after the non-ASME Code repair.
3.7 Staffs Evaluation Based on the above discussion, the NRC staff agrees with the licensee that performing an ASME Code repair of the degraded shutdown cooling piping when the piping is needed for decay heat removal would result in hardship or unusual difficulty.
The affected piping is ASME Code, Class 2, which was designed to the 1967 edition through 1969 addenda of the United States of America Standards Institute (now American National Standards Institute) B31.1 Code. The piping and associated supports are designed for pressure, deadweight, thermal, operating and design-basis earthquake loading, and torus-attached piping dynamic loads. The piping is fabricated with carbon steel. The pipe is 20-inch in nominal diameter with a wall thickness of 0.375 inches. The system operating pressure is 50 psig and design pressure is 150 psig with an operating temperature of 90 to 280 degrees F.
The crack is 6.5 inches in length, and follows the toe of the pipe support trunnion-to-pipe fillet weld. The integral attachment weld is a 3/8-inch fillet weld, continuous around the 6-inch diameter (with 6.625-inch outside diameter) pipe trunnion for a total length of 21 inches with the crack comprising about 31 percent of the total weld length.
The crack extent was determined based on a VT-1 visual examination and confirmed by a fluorescent wet-magnetic surface examination for about 60 percent of the weld based on accessibility. The remaining third of the weld (back side of the trunnion) opposite of the crack was verified to have no indications based on a penetrant examination.
The licensee will drill a hole at both ends of the crack to relieve the stresses in the crack. The NRC staff finds that this is a generally accepted method of crack arrest to prevent the crack from further propagation.
The licensee will use the shielded metal arc welding process with a 3/32-inch E7018 electrode to deposit weld metal on the crack surface. The seal weld deposit is anticipated to require three beads laid adjacent to each other which will produce 1 layer of weld about 1/8-inch thick.
The licensee stated that the purpose of the temporary seal weld is to prevent further leakage until an ASME Code repair can be performed. The NRC staff agrees with the licensee that the seal weld is to prevent leakage. However, it also provides certain bonding on the crack face to minimize any potential crack propagation. The staff noted that the seal weld would not provide the same structural integrity to the affected piping as the ASME Code repair. Nevertheless, the seal weld would provide a certain level of structural integrity to the affected piping to minimize crack propagation.
As part of the non-ASME Code repair, the licensee has modified the existing pipe supports in the affected area of the pipe. The trunnion of the affected pipe support (PFSK-2285) is designed as a two way vertical support. About 2 feet away is a deadweight support (PFSK-2084) that was found to be misadjusted. This resulted in the PFSK-2285 pipe support carrying additional deadweight load for which it was not designed. PFSK-2285 also supports vertical dynamic loads (seismic and torus-attached piping loads) and a small downward thermal load.
The normal piping vibration was attributed as the primary factor in the apparent high cycle fatigue failure near the trunnion-to-pipe attachment weld. The localized vibration at the PFSK-2285 pipe support has been reduced by restoring the adjacent pipe support (PFSK-2084) by installing a shim plate where the support was not load-bearing as designed. The licensee stated that vibration is no longer a concern for the short term at the degraded pipe condition.
The NRC staff agrees with this assessment.
In addition, the properly shimmed PFSK-2084 pipe support will reduce the deadweight load that the PFSK-2855 pipe support has been carrying. As a preventive measure, the licensee installed a temporary deadweight support under the pipe about 5 feet away from the PFSK-2285 support and on the opposite side of the PFSK-2084 support. This means that two pipe supports are straddling the degraded PFSK-2285 support which will reduce the loading on the crack and thus minimize crack propagation.
The licensee stated that its analysis showed that the fillet weld could be reduced 50 percent in length and still provide adequate load capacity for the design conditions. With the additional temporary support, the loading on the fillet weld would be reduced which would provide a favorable condition for the affected piping prior to the ASME Code repair.
The NRC staff finds that the temporary non-ASME Code repair is acceptable because (1) the licensee has performed appropriate crack arrest; (2) the seal weld applied to the crack opening will minimize further crack propagation; and (3) the crack will be stabilized and will not experience much loading during the non-ASME Code repair condition because a temporary support is installed adjacent to the PFSK-2285 support and the existing PFSK-2084 support has been properly shimmed.
In addition, in the July 10, 2005, letter, the licensee provided a regulatory commitment, which states that ENO [Entergy Nuclear Operations, Inc.] will complete the ASME Code repairs and required inspections to the RHR SDC [shutdown cooling] piping prior to startup (entry into Modes 1 or 2) from the current forced outage. This commitment was completed prior to plant startup.
4.0 CONCLUSION
Based on the review of information submitted, the NRC staff has determined that the licensee has demonstrated that compliance with the applicable ASME Code repair/replacement requirements for the degraded shutdown cooling piping of the RHR system would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The staff finds that the licensees non-ASME Code repair as discussed in relief request No.
RR-38, will provide an acceptable level of structural integrity to the degraded shutdown cooling piping prior to the ASME Code repair. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the licensees proposed non-ASME Code repair of the common suction header of the shutdown cooling piping of the RHR system at JAF.
All other requirements of Section XI of the ASME Code for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: John Tsao, John Boska Date: August 9, 2005