Letter Sequence Other |
---|
|
|
MONTHYEARML0700403382007-02-0101 February 2007 Relief, Request for Alternative No. ANO1-ISI-006 to Extend the Third Inservice Inspection Interval for Reactor Vessel Examination Category B-F Weld Examinations Project stage: Other 2007-02-01
[Table View] |
|
---|
Category:Code Relief or Alternative
MONTHYEARML23142A1062023-05-24024 May 2023 Authorization and Safety Evaluation for Alternative Request No. ANO1-ISI-035 ML22073A0952022-04-0707 April 2022 Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID L-2021-LLR-0084) (NON-PROPRIETARY) 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 ML21293A0962021-10-28028 October 2021 Relief ANO2-R&R-012 Concerning Proposed Alternative for Repair of Reactor Vessel Closure Head Penetration 46 ML21168A0562021-07-0909 July 2021 Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code ML21118B0392021-05-19019 May 2021 Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code CNRO-2020-00016, Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2020-08-12012 August 2020 Entergy Nuclear Operations, Inc - Relief Request EN-RR-20-002: Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) ML20121A1932020-05-0606 May 2020 Alternative Requests for the Fifth 10 Year Interval Inservice Testing of Low Pressure Safety Injection Pumps and Service Water Valves ML20119A0732020-04-29029 April 2020 Withdrawal of Requested Alternatives for the Fifth 10-Year Interval Inservice Testing of Service Water Pumps and Visual Examination of Snubbers ML20083J9682020-03-26026 March 2020 Approval of Request for Alternative ANO 2-PT-002 End-of-Internal System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping ML19254A6032019-09-19019 September 2019 Units 2 and 3; Palisades Nuclear Plant; River Bend; and Waterford Steam Electric Station, Unit 3 - Relief Request No. EN-19-RR-1, Use of ASME Code Case N-831-1 2CAN071902, Request for Alternative ANO 2-PT-002 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping2019-07-25025 July 2019 Request for Alternative ANO 2-PT-002 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping ML19184A5542019-07-15015 July 2019 Request for Relief ANO1-ISI-032 from the ASME Code Section XI Visual Examination Requirements for the Fourth 10-Year Inservice Inspection Interval ML19156A4002019-06-11011 June 2019 Request for Alternative ANO2-ISI-021 to Permit Continued Application of the 2007 Edition Through the 2008 Addenda of ASME Code CNRO-2019-00002, Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-01-31031 January 2019 Relief Request Number EN-19-RR-1, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 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 1CAN041801, Requests for Relief from American Society of Mechanical Engineers (ASME) Section XI Volumetric Examination Requirements Fourth 10-Year Interval, Second and Third Period2018-04-23023 April 2018 Requests for Relief from American Society of Mechanical Engineers (ASME) Section XI Volumetric Examination Requirements Fourth 10-Year Interval, Second and Third Period CNRO-2017-00022, Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 12017-11-17017 November 2017 Relief Request Number EN-17-RR-1 - Proposed Alternative to Use ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate-Energy Class 2 or 3 Piping, Section XI, Division 1 ML17174B1442017-07-12012 July 2017 Relief Request for EN-ISI-16-1 Regarding Use of Later Edition and Addenda of the ASME Code ML16130A4712016-12-0909 December 2016 Request for Alternative Test Plan to American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants (CAC Nos. MF7537 and MF7538) ML16237A0822016-08-29029 August 2016 Relief Request No. ANO1-ISI-025, Relief from American Society of Mechanical Engineers Section XI Table IWB-2500-1 Requirements 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 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 ML15070A4282015-03-16016 March 2015 Relief Request ANO2-ISI-017, Alternative to Utilize ASME Code Case N-513-4 for the Fourth 10-Year Inservice Inspection Interval ML15030A2392015-02-0909 February 2015 Correction to Relief Request ANO1-ISI-024 Issued December 23, 2014; Revises 10 CFR 50.55a Citations as a Result of Changes from Final Rule Effective December 5, 2014 ML14330A2072014-12-23023 December 2014 Relief Request ANO1-ISI-024, Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1, Fourth 10-year Inservice Inspection Interval ML14282A4792014-10-29029 October 2014 Relief Request ANO1-ISI-023, Volumetric Examination Frequency Requirements, for the Fourth 10-Year Inservice Inspection Interval ML14150A1632014-06-0606 June 2014 Relief Request ANO2-ISI-004 - Alternative to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations ML14099A4522014-04-15015 April 2014 Relief Request ANO2-ISI-016, from ASME Code Case N-770-1 Successive Examination, for the Fourth 10-Year Inservice Inspection Interval ML13179A1162013-07-24024 July 2013 Request for Relief ANO1-ISI-022 from ASME Code Examination Requirements for Pressure-Retaining Welds in Piping; Fourth 10-Year Inservice Inspection Interval ML13179A1102013-07-16016 July 2013 Request for Relief No. ANO1-ISI-021, from ASME Volumetric Requirements for Full Penetration Welded Nozzles in Vessels - Inspection Program B; Fourth 10-Year ISI Interval ML13161A2412013-06-27027 June 2013 Relief Request ANO2-ISI-015, from Pressure Testing Requirements for Reactor Vessel Flange Seal Leak Detection Piping for the Fourth 10-Year Inservice Inspection Interval ML13071A6342013-03-22022 March 2013 Relief Request ANO2-ISI-014, from Volumetric and Surface Exam Requirements for Pressure Retaining Welds in Austenitic Stainless Steel, Third 10-Year Inservice Inspection Interval ML13071A6632013-03-21021 March 2013 Relief Request ANO2-ISI-010, from Volumetric and Surface Exam Requirements for Pressure Retaining Welds in Piping, Third 10-Year Inservice Inspection Interval ML13032A5732013-02-21021 February 2013 Relief Request Nos. ANO2-ISI-009, ANO2-ISI-011, and ANO2-ISI-012, ASME Code Volumetric and Surface Exam Requirements for Integral Attachment and Pressure-Retaining Welds, Third 10-Year Inservice Inspection Interval ML12319A3672012-11-27027 November 2012 Relief Request ANO2-ISI-007, Alternative to Use ASME Code Case N-770-1 Baseline Examination, Fourth 10-Year Inservice Inspection Interval 1CAN071202, Requests for Relief from American Society of Mechanical Engineers (ASME) Section XI Volumetric Examination Requirements, Fourth 10-Year Interval, First Period2012-07-25025 July 2012 Requests for Relief from American Society of Mechanical Engineers (ASME) Section XI Volumetric Examination Requirements, Fourth 10-Year Interval, First Period 2CAN111101, Use of Alternate ASME Code Case N-770-1 Baseline Examination Request for Alternative ANO2-ISI-0072011-11-30030 November 2011 Use of Alternate ASME Code Case N-770-1 Baseline Examination Request for Alternative ANO2-ISI-007 ML1117107982011-08-10010 August 2011 Relief, ANO1-R&R-016, Request for Use of Non-ASME Code Repair to Service Water Piping in Accordance with Generic Letter 90-05, Fourth 10-Year Interval. ME6107 ML1034301562011-01-10010 January 2011 Relief Request ANO1-R&R-013, Proposed Alternative to Requirements Associated with Repair of Components, for Duration of ANO-1 Spring 2010 Refueling Outage 1R22 ML1035005322011-01-0505 January 2011 Request for Alternative ANO2-ISI-006, Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 for the Fourth 10-Year ISI Interval ML1025001282010-09-23023 September 2010 Relief Request ANO2-ISI-005 to Extend the Inservice Inspection Interval for the Visual Examination of Accessible Interior to Reactor Vessel Attachment Weld, for Fourth 10-Year Interval ML1024506542010-09-21021 September 2010 Relief Request ANO2-ISI-004 to Extend Third Inservice Inspection Interval from 10 to 20 Years for Reactor Vessel Weld Examinations ML1009103502010-08-19019 August 2010 Request for Relief PRR-ANO2-2010-1, Relief from the Requirements of Operation and Maintenance Code Section ISTB-5221 and ITSB-5223 for the Service Water Pumps ML1011701272010-06-0202 June 2010 Request for Alternative ANO1-ISI-014, to Implement a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 ML1011701192010-05-0505 May 2010 Request for Relief Nos. ANO1-ISI-015, ANO1-ISI-016, ANO1-ISI-017, ANO1-ISI-018, ANO1-ISI-019, and ANO1-ISI-020, Relief from Requirements of the ASME Code, Section XI ML1009706592010-04-26026 April 2010 Request for Alternative ANO2-ISI-003 to Use 2001 Edition Through 2003 Addenda in Lieu of the 2004 Edition for Fourth 10-Year Inservice Inspection Interval 2CAN021002, Request for Relief PRR-ANO2-2010-1 from the Requirements of OM Code Section ISTB-5221 and ISTB-5223 for the Service Water Pumps2010-02-17017 February 2010 Request for Relief PRR-ANO2-2010-1 from the Requirements of OM Code Section ISTB-5221 and ISTB-5223 for the Service Water Pumps 2CAN011005, Supplement to Request for Alternative Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-7162010-01-28028 January 2010 Supplement to Request for Alternative Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 ML0923005512009-08-27027 August 2009 Request for Alternative ANO2-ISI-002 to 10 CFR 50.55a(g)(6)(ii)(D) Examination Requirements, for Remainder of Current 10-Year Inservice Inspection Interval and Fourth ISI Interval Until the Head Is Replaced 2023-05-24
[Table view] Category:Letter
MONTHYEARML23326A0392024-01-24024 January 2024 Issuance of Amendment No. 281 Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML24017A1582024-01-17017 January 2024 Submittal of Emergency Plan Revision 50 IR 05000313/20234202024-01-10010 January 2024 Security Baseline Inspection Report 05000313/2023420 and 05000368/2023420 IR 05000313/20234022024-01-0202 January 2024 Security Baseline Inspection Report 05000313/2023402 and 05000368/2023402 ML23354A0022023-12-21021 December 2023 Request for Withholding Information from Public Disclosure ML23349A1672023-12-21021 December 2023 Request for Withholding Information from Public Disclosure ML23348A3572023-12-14014 December 2023 Application to Revise Technical Specifications to Use Online Monitoring Methodology Slides and Affidavit for Pre-Submittal Meeting ML23352A0292023-12-13013 December 2023 Entergy - 2024 Nuclear Energy Liability Evidence of Financial Protection ML23340A1592023-12-13013 December 2023 Entergy Operations, Inc. - Entergy Fleet Project Manager Assignment IR 05000313/20234052023-12-12012 December 2023 Security Baseline Inspection Report 05000313/2023405 and 05000368/2023405 ML23341A0832023-12-11011 December 2023 Material Control and Accounting Program Inspection Report 05000313/368/2023404- Cover Letter ML23305A0922023-12-0707 December 2023 Summary of Regulatory Audit Regarding License Amendment Request to Revise Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times RITSTF Initiative 4b ML23333A1362023-11-29029 November 2023 Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23275A2072023-11-28028 November 2023 Issuance of Amendment No. 280 Removal of Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23325A1412023-11-21021 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000313/20230032023-11-21021 November 2023 Revised - ANO Revised Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/ 2023001 ML23243B0452023-11-13013 November 2023 Request for Withholding Information from Public Disclosure ML23313A0962023-11-13013 November 2023 Integrated Inspection Report 05000313/2023003 and 05000368/2023003 and Independent Spent Fuel Storage Installation Inspection Report 07200013/2023001 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV IR 05000313/20230112023-10-10010 October 2023 Commercial Grade Dedication Inspection Report 05000313/2023011 and 05000368/2023011 IR 05000313/20230052023-08-21021 August 2023 Updated Inspection Plan for Arkansas Nuclear One, Units 1 and 2 (Report 05000313/2023005 and 05000368/2023005) - Mid Cycle IR 05000313/20243012023-08-14014 August 2023 Notification of NRC Initial Operator Licensing Examination 05000313/2024301 IR 05000313/20230022023-08-11011 August 2023 Integrated Inspection Report 05000313/2023002 and 05000368/2023002 and Notice of Violation ML23209A6022023-08-0909 August 2023 Regulatory Audit Plan in Support of License Amendment Request to Revise Technical Specifications to Adopt Risk- Informed Completion Times IR 05000416/20230022023-08-0808 August 2023 Integrated Inspection Report 05000416/2023002 and Independent Spent Fuel Storage Installation Inspection Report 07200050/2023001 ML23208A2132023-08-0303 August 2023 Regulatory Audit Summary Concerning Review of License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23208A2112023-07-27027 July 2023 Entergy Operations Inc., Application to Revise Technical Specifications to Adopt TSTF-205, Revision of Channel Calibration, Channel Functional Test, and Related Definitions ML23188A1732023-07-12012 July 2023 Request for Withholding Information from Public Disclosure ML23142A2022023-06-29029 June 2023 Issuance of Amendment Nos. 279 and 332 Emergency Plan Staffing Requirements IR 05000313/20234032023-06-28028 June 2023 Security Baseline Inspection Report 05000313/2023403 and 05000368/2023403 ML23178A0892023-06-26026 June 2023 August 2023 Emergency Preparedness Program Inspection-Request for Information ML23166B0902023-06-21021 June 2023 Request for Relief ANO1 ISI 036 Regarding Volumetric Examination Requirements ML23180A1082023-06-20020 June 2023 ANO Unit 1 SAR Amendment 31, TRM, TS Bases, 10 CFR 50.59 Report, and Commitment Change Summary Report IR 05000313/20234012023-06-13013 June 2023 Cyber Security Inspection Report 05000313/2023401 and 05000368/2023401 ML23130A2972023-05-17017 May 2023 Notification of Inspection (NRC Inspection Report 05000313/2023003) and Request for Information ML23121A3012023-05-10010 May 2023 Regulatory Audit Plan in Support of License Amendment Request to Revise Technical Specifications to Adopt Risk - Informed Completion Times ML23114A3572023-05-10010 May 2023 Regulatory Audit Summary Concerning Review of Request for Alternative ANO-CISI-002 Regarding Containment Inservice Inspection Frequency ML23130A2732023-05-10010 May 2023 and Waterford 3 Steam Electric Station - Response to Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations IR 05000313/20230012023-05-0909 May 2023 Integrated Inspection Report 05000313/2023001 and 05000368/2023001 ML23118A3202023-05-0808 May 2023 Notification of Biennial Problem Identification and Resolution Inspection and Request for Information (05000313;05000368/2023010) IR 05000368/20233012023-05-0101 May 2023 NRC Examination Report 05000368/2023301 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML23108A1362023-04-24024 April 2023 Approval of Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration No. 46 ML23110A1122023-04-20020 April 2023 Annual Report for Entergy Quality Assurance Program Manual Changes Under 10 CFR 50.54(a)(3), 10 CFR 71.106, and 10 CFR 72.140(d). Notification of Application of Approved Appendix B to 10 CFR 72 Subpart G 2024-01-24
[Table view] Category:Safety Evaluation
MONTHYEARML23326A0392024-01-24024 January 2024 Issuance of Amendment No. 281 Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML23275A2072023-11-28028 November 2023 Issuance of Amendment No. 280 Removal of Technical Specification Condition Allowing Two Reactor Coolant Pump Operation ML23142A2022023-06-29029 June 2023 Issuance of Amendment Nos. 279 and 332 Emergency Plan Staffing Requirements ML23166B0902023-06-21021 June 2023 Request for Relief ANO1 ISI 036 Regarding Volumetric Examination Requirements ML23142A1062023-05-24024 May 2023 Authorization and Safety Evaluation for Alternative Request No. ANO1-ISI-035 ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment ML22342B1602023-01-10010 January 2023 Requests for Relief ANO2-ISI-023, -024, -025, -026, -027, and -028 to Permit Reduction of Inspection Area Due to Interference (EPIDs L-2022-LLR-0022,-0023, -0024, -0025, -0026, and -0027) ML22263A1912022-12-0909 December 2022 Issuance of Amendment No. 278 Revision to Technical Specifications 3.4.12 and 3.4.13 Based on Revised Dose Calculations (EPID L-2021-LLA-0181) - Non-Proprietary ML22236A5452022-08-26026 August 2022 Correction to Issuance of Amendment No. 331 Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors ML22138A4312022-06-23023 June 2022 Issuance of Amendment No. 277 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors ML22104A2222022-05-12012 May 2022 Issuance of Amendments Revise Technical Specifications to Adopt TSTF 554 ML22083A1242022-04-28028 April 2022 Arkansas, Units 1 and 2; Grand Gulf Nuclear Station; River Bend Station; and Waterford Steam Electric Station - Issuance of Amendments Revise Technical Specifications to Adopt TSTF-541, Revision 2 ML22039A2822022-03-15015 March 2022 Issuance of Amendment Nos. 274 and 328 One-Time Change to Support Proactive Upgrade of the Emergency Cooling Pond Supply Piping ML22007A3172022-01-18018 January 2022 1, River Bend Station 1, and Waterford Steam Electric Station 3 - Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI ML21225A0552021-12-17017 December 2021 Issuance of Amendment No. 327 Addition of Technical Specification Limiting Condition for Operation 3.0.6 and Adoption of Safety Function Determination Program ML21341B2262021-12-13013 December 2021 Acceptance of Requested Licensing Action License Amendment Request to Revise Technical Specifications 3.4.12 and 3.4.13 Based on Revised Dose Calculations ML21313A0082021-12-0808 December 2021 Issuance of Amendments to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections ML21208A4492021-09-20020 September 2021 Issuance of Amendment No. 325 Revise Loss of Voltage Relay Allowable Values and Number of Degraded Voltage Channels ML21132A2792021-05-19019 May 2021 Request for Approval of Change to the Entergy Quality Assurance Program Manual ML21088A4332021-04-14014 April 2021 Issuance of Amendment No. 324 Adoption of Technical Specification Task Force Traveler, TSTF-569, Revision 2, Revise Response Time Testing Definition ML21027A4282021-03-23023 March 2021 Issuance of Amendment No. 272 Replacement of Reactor Building Spray Sodium Hydroxide Additive with a Passive Reactor Building Sump Buffering Agent ML21040A5132021-03-10010 March 2021 Issuance of Amendment No. 271 Revision to Loss of Voltage Relay Allowable Values ML20351A1532021-02-0808 February 2021 Issuance of Amendment No. 323 Technical Specification Deletions, Additions, and Relocations ML20288A8242020-11-16016 November 2020 Regulatory Audit Summary Concerning License Amendment Request to Replace the Reactor Building Spray Sodium Hydroxide Additive with a Passive Reactor Building Sump Buffering Agent NUREG-1432, Issuance of Amendment No. 322 Technical Specification Changes Related to Revised Fuel Handling Accident Analysis and Adoption of Technical Specification Improvements Consistent with NUREG-14322020-10-30030 October 2020 Issuance of Amendment No. 322 Technical Specification Changes Related to Revised Fuel Handling Accident Analysis and Adoption of Technical Specification Improvements Consistent with NUREG-1432 ML20226A2722020-08-18018 August 2020 Request to Use a Provision of a Later Edition of the ASME BPV Code, Section XI ML20160A1472020-06-30030 June 2020 Issuance of Amendment No. 270 to Adopt TSTF 563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program ML20135H1412020-06-30030 June 2020 Issuance of Amendment Nos. 269 and 321 Request to Incorporate the Tornado Missile Risk Evaluator Into the Licensing Basis ML20121A1932020-05-0606 May 2020 Alternative Requests for the Fifth 10 Year Interval Inservice Testing of Low Pressure Safety Injection Pumps and Service Water Valves ML20107J3172020-04-23023 April 2020 Relief Request ANO1-ISI-033 Related to ASME Code Case N-729-4 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles ML20087L8032020-04-0808 April 2020 Issuance of Amendment No. 320 to Revise Control Element Assembly Drop Times ML20083J9682020-03-26026 March 2020 Approval of Request for Alternative ANO 2-PT-002 End-of-Internal System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping ML19231A2972020-01-29029 January 2020 Issuance of Amendment No. 318 Addition of Technical Specification Actions to Address Inoperability of the Containment Building Sump ML19220A9382019-09-27027 September 2019 Issuance of Amendment No. 266 Adoption of Technical Specifications Task Force Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues ML19254A6032019-09-19019 September 2019 Units 2 and 3; Palisades Nuclear Plant; River Bend; and Waterford Steam Electric Station, Unit 3 - Relief Request No. EN-19-RR-1, Use of ASME Code Case N-831-1 ML19175A0422019-09-11011 September 2019 Arkansas Units 1 and 2; Grand Gulf, Unit 1; Indian Point 2 and 3; Palisades; River Bend, Unit 1; Waterford Unit 3 - Issuance of Amendments to Adopt TSTF-529, Clarify Use and Application Rules ML19184A5542019-07-15015 July 2019 Request for Relief ANO1-ISI-032 from the ASME Code Section XI Visual Examination Requirements for the Fourth 10-Year Inservice Inspection Interval ML19156A4002019-06-11011 June 2019 Request for Alternative ANO2-ISI-021 to Permit Continued Application of the 2007 Edition Through the 2008 Addenda of ASME Code ML19098A9552019-05-22022 May 2019 Issuance of Amendment No. 264 Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3 ML19063B9482019-04-23023 April 2019 Issuance of Amendment No. 315 Adoption of Technical Specification Task Force Traveler TSTF-425, Revision 3 ML18337A2472019-01-17017 January 2019 Issuance of Amendment Nos. 263 and 314 Revision to Emergency Action Level Scheme ML18346A3142019-01-0707 January 2019 Relief Nos. ANO1-ISI-027, -028, -029, -030, and -031 from ASME Code, Section XI, Examination Requirements for Fourth 10-Year Inservice Inspection Interval (EPID L-2018-LLR-0063, -0064, -0065, -0066, -0067) ML18317A3822018-12-19019 December 2018 Issuance of Amendment No. 313, Revise Technical Specification 3.3.3.6, Post-Accident Instrumentation ML18298A0122018-11-27027 November 2018 Issuance of Amendment No. 311 Updating the Reactor Coolant System Pressure Temperature Limits ML18260A3392018-10-24024 October 2018 Issuance of Amendment No. 261 Revision to Technical Specification Bases Related to Emergency Feedwater Turbine-Driven Pump Steam Supply Valves ML18192A1042018-08-15015 August 2018 Alternative to Examination of Reactor Pressure Vessel Flange Threads ML18173A0192018-06-27027 June 2018 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule (CAC MG0244; EPID L-2017-LLL-0020) ML18115A2822018-06-19019 June 2018 Issuance of Amendment No. 260 Technical Specification Changes Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-412 Revision 3 (CAC No. MG0009; EPID L-2017-LLA-0267) ML18134A2532018-06-19019 June 2018 Issuance of Amendment No. 310 Technical Specification Changes to Provide Actions for Emergency Feedwater Pump Inoperability Consistent with NUREG-1432, Revision 4 (CAC No. MG0016; EPID L-2017-LLA-0268) 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 2024-01-24
[Table view] |
Text
February 1, 2007Mr. Jeffrey S. ForbesSite Vice President Arkansas Nuclear One Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT 1 - REQUEST FOR ALTERNATIVENO. ANO1-ISI-006 TO EXTEND THE THIRD INSERVICE INSPECTION INTERVAL FOR REACTOR VESSEL EXAMINATION CATEGORY B-F WELD EXAMINATIONS (TAC NO. MD1397)
Dear Mr. Forbes:
By letter dated April 24, 2006, Entergy Operations, Inc. (Entergy), the licensee for ArkansasNuclear One, Unit 1 (ANO-1), submitted a request for authorization to extend the third 10-year inservice inspection (ISI) interval for the reactor core flood line nozzle-to-safe end weld examinations. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR)50.55a(a)(3)(ii), the licensee requests approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Paragraph IWB-2412, for the ANO-1 nuclear power plant. In response to the Nuclear Regulatory Commission's (NRC) Request for Additional Information, Entergy resubmitted a revised request for authorization in a letter dated September 8, 2006.Based on the NRC staff's review of the information provided by the licensee in its letters datedApril 24 and September 8, 2006, authorizing the proposed alternative provides reasonabel assurance of structural integrity and is justified on the basis that compliance with the specified requirements would result in hardship without a compensating increase in the level of quality and safety. Therefore, the staff authorizes the proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year ISI interval at ANO-1. The proposed alternative is authorized until the end of the ANO-1 fall 2008 refueling outage.
J. Forbes-2-The NRC staff's safety evaluation is enclosed.Sincerely,/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-313
Enclosure:
Safety Evaluation cc w/encl: See next page J. Forbes-2-The NRC staff's safety evaluation is enclosed.Sincerely,/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-313
Enclosure:
Safety Evaluation cc w/encl: See next pageDISTRIBUTION
- PUBLIC LPLIV r/f RidsAcrsAcnwMailCenter RidsNrrDorlLpl4 RidsNrrLAJBurkhardtRidsNrrPMFSabaRidsOgcRp RidsRgn4MailCenter DCullison, EDO Region IV AShaikh, NRRADAMS Accession No.:
ML070040338OFFICEDORL/LPL4/PMDORL/LPL4/LADCI/CVIB/BCOGC DORL/LPL4/BCNAMEFSabaLFeizollahiKGruss*JRundDTerao DATE1/19/071/19/0711/1/061/30/072/1/07OFFICIAL AGENCY RECORD December 2006Arkansas Nuclear One cc:Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995General Manager PlantOperations Entergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802Director, Nuclear Safety AssuranceEntergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802Manager, LicensingEntergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802Director, Nuclear Safety & LicensingEntergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213-8298Section Chief, Division of HealthRadiation Control Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867Section Chief, Division of HealthEmergency Management Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867Senior Resident InspectorU.S. Nuclear Regulatory Commission P.O. Box 310 London, AR 72847Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064County Judge of Pope County 100 W. Main Street Russellville, AR 72801Vice President, Operations SupportEntergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONREQUEST FOR ALTERNATIVE TO ASME SECTION XIPROPOSED ALTERATIVE TO EXTEND THE THIRD 10 YEAR INSERVICE INSPECTION INTERVAL FOR REACTOR VESSEL EXAMINATION CATEGORY B-F WELD EXAMINATIONSENTERGY OPERATIONS, INC.ARKANSAS NUCLEAR ONE, UNIT 1DOCKET NUMBER 50-31
31.0INTRODUCTION
By letter dated April 24, 2006, Entergy Operations, Inc. (Entergy), the licensee for ArkansasNuclear One, Unit 1 (ANO-1), submitted a request for authorization to extend the third 10-year inservice inspection (ISI) interval for the reactor core flood line nozzle-to-safe end weld examinations. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR),paragraph 50.55a(a)(3)(ii), the licensee requests approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Paragraph IWB-2412, for the ANO-1 nuclear power plant. In response to the Nuclear Regulatory Commission's (NRC) Request for Additional Information, Entergy resubmitted a revised request for authorization in a letter dated September 8, 2006.ANO-1 is currently in its third ISI interval, which began June 1, 1997, and ends May 31, 2007. ASME Code,Section XI, IWA-2430(d) allows for a 1-year extension of an interval, which would extend the current interval to May 31, 2008. (Use of this 1-year extension does not require approval from the NRC.) In order to comply with ASME Code requirements, third interval visual examinations of the Category B-F welds must be performed during ANO-1's spring 2007 refueling outage (1R20). Entergy proposes to perform these examinations during the fall 2008 refueling outage (1R21). 2.0REGULATORY REQUIREMENTSIn accordance with 10 CFR 50.55a, licensees are required to perform periodic inspections ofcomponents. Paragraph 50.55a(g) of 10 CFR requires that licensees perform surveillance testing in accordance with the ASME Code,Section XI requirements. Paragraph 50.55a(a)(3) of 10 CFR states: "Proposed alternatives to the requirements of (c),(d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director of 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." Entergy believes that compliance with the requirements of ASME Code,Section XI, IWA-2432 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, Entergy requests the NRC staff approve this proposed alternative in accordance with 10 CFR 50.55a(a)(3)(ii).
3.0TECHNICAL EVALUATION
3.1Components for which Relief is RequestedThe affected component is the ANO-1 reactor vessel (RV) dissimilar metal weld associated withthe reactor core flood-line nozzle-to-safe end welds. They are ASME Code, Section Xl, Examination Category B-F, and item number B5.10. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 ofthe ASME Code, Section Xl, 1992 Edition.3.2ASME Code RequirementsASME Code,Section XI, IWB-2412, 1992 Edition, Inspection Program B, requires volumetricexamination of RV and piping pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. Paragraph IWA-2430(d) allows inspection intervals to be extended by as much as 1 year if this adjustment does not cause successive intervals to be altered by more than 1 year. Item B5.70 of Table IWB-2500-1 requires that the nozzle-to-safe end butt welds have bothvolumetric and surface examinations every inspection interval. Note 1 of Table IWB-2500 for Examination Category B-F welds states, "[f]or the reactor vessel nozzle safe ends, the examinations may be performed coincident with the vessel nozzle examinations required by Examination Category B-D." 3.3Licensee's Request for ReliefEntergy requests to extend the ISI interval for the identified examination Category B-F,identified in Sections 3.1 of this safety evaluation, to the end of 1R21 (approximately 180 days beyond the currently scheduled interval and the ASME Code-allowed 1-year extension). 3.4Basis for Proposed AlternativeGiven approval to extend the ANO-1's 10-year reactor vessel ISI interval to the end of1R21 as contained in Entergy's Request for Alternative ANO1-ISI-005 (Reference 3),
Entergy also seeks to extend the ASME Code-required core flood-line weld inspections until 1R21. In order to perform the core flood-line Examination Category B-F weld inspections, access to the weld from inside of the reactor vessel as described below is required. Therefore, performing this examination during the upcoming 1R20 refueling outage will result in a hardship without a compensatory increase in the level of quality and safety. The licensee's justification supporting this request is provided below. A.Need to Perform Reactor Core Flood Line Dissimilar Metal Weld Examinationsduring Reactor Vessel ISI The ANO-1 reactor vessel has two (2) ASME Code Class 1 (14-inch) carbonsteel core flood nozzles with stainless steel safe ends. The carbon steel nozzles are lined with stainless steel cladding and have Alloy-82 butter with an Alloy-182 butt-weld to the stainless steel safe ends. The core flood lines connect directly into the reactor vessel and experience temperatures near that of the cold leg temperatures (approximately 577°F) during power operation. As shown in Figure 1 [of the April 24, 2006, submittal], the nozzle butt welds arelocated between the reactor vessel and the concrete shield wall. Above these butt welds is the permanent reactor cavity seal plate used for refueling. This plate is located at the reactor vessel flange which is located at plant elevation 376 feet, 6 inches. The centerline of the core flood lines is at plant elevation 371 feet, 6 inches, which is 5 feet below the seal plate. Just below the seal plate is a series of four (4) concrete shield plugs. The core flood nozzles are located approximately one (1) foot below the shield plugs. Below the shield plugs is metal reflective insulation that surrounds the core flood lines. Access to the core flood nozzles from above would require a modification to thereactor cavity seal plate to allow removal of the reactor cavity shield plugs. The existing man-ways in the reactor cavity seal plate do not allow for removal of the shield plugs. Access to the core flood nozzle outside diameter (OD) from below is not feasible due to the reactor vessel skirt configuration, which does not provide a man-way into the reactor vessel cavity. Access to the core flood nozzle OD from the shield wall penetration is also not feasible since that portion inside the reactor cavity is covered by removable reactor vessel insulation panels that cover the subject welds. Removal of the insulation panels to allow inspections is not feasible due to the distance through the penetration (approximately 6 feet), and the limited clearance between the piping and penetration walls (approximately 4 inches). Therefore, the core flood nozzle examinations per ASME Section XI, IWB-2500,are performed from the inside of the reactor vessel as part of the reactor vessel ISI at the normal interval frequency. As noted above, Entergy has submittedRequest for Alternative ANO1-ISI-005 to defer the reactor vessel ISI (Reference 3).B.Plant-Specific Reactor Vessel ISI History ANO-1 is in its third inservice inspection interval for the reactor vessel; therefore,the preservice and two inservice inspections have been performed on the Examination Category B-F welds. These inspections achieved acceptable coverage; no reportable indications were found. Based on the examination method and coverage obtained, it is reasonable to conclude that the examinations were of sufficient quality to detect any significant flaws that could challenge reactor vessel integrity. The last examination findings are provided in the following table.ANO-1 ISI Results Weld IDASME Weld CategoryDate Last Inspected% CoverageObtained# of Reportable Indications
- of IndicationsCurrently being MonitoredGrowth of IndicationsCurrently being Monitored (in)01-025B-F199510000N/A01-026B-F199510000N/AC.Safety Significance of Core Flood Line Deferral for One Outage General experience has shown that the incidence of primary water stress corrosion cracking (PWSCC) in PWR [pressurized-water reactors] primary coolant systems is due to tensile stresses and the operating temperature. The primary sources of tensile stresses are weld residual stresses and applied operating stresses (such as pressure and temperature). Higher mechanical stresses are typically due to fabrication activities where welding residual stresses cause shrinkage of the material around the weld.
For materials of equal PWSCC susceptibility with equal applied stresses, the time to crack initiation is a function of operating temperature. Locations that operate at higher temperatures, such as pressurizers, typically exhibit cracking much sooner than locations that operate at lower temperatures, such as cold legs. Therefore, Alloy 600 butt welds that operate closer to cold leg temperatures are generally much less susceptible to PWSCC than those that are close to Reactor Coolant System (RCS) operating temperatures. EPRI [Electric Power Research Institute] document MRP-112 (Reference 1)documents the susceptibility of various locations of butt welds on Babcock &
Wilcox (B&W) designed reactors. The assessment included a review of crack orientations and sizes, welding stresses, crack growth rates, limiting flaw sizes and the probability to determine the susceptibility of various butt welds. The B&W core flood line butt welds were specifically evaluated in this report. Thecore flood lines operate at temperatures that are only slightly above the cold leg temperature (577°F). Critical flaw sizes for a through-wall flaw were determined using the methodology of ASME Code Section XI [and Code Case N-513]. The critical flaw sizes were determined for Axial and Circumferential through wall flaw lengths and for a 75% through-wall circumferential flaw depth. The core flood nozzle critical flaw sizes are as follows: Axial Flaw LengthCirc Flaw LengthCirc Flaw Depth22.3 inches20.7 inches0.75%Using the critical flaw data, crack growth analyses were performed under bothPWSCC conditions and fatigue. The results of this report for the B&W core flood lines show that the time forpropagation of a flaw from an initial flaw size that produces an applied stress intensity factor equal to the PWSCC threshold stress intensity factor until 75%
through-wall would take greater than 40 years. The time from identification of a 1-gpm [gallon per minute] leak until it reached a critical flaw size would be in excess of 70 years. The results of this analysis were also reflected in MRP-113 (Reference 2). The core flood lines at ANO-1 were inspected in 1995 during the last 10-year inspection interval and no flaws were identified in these welds.
Therefore, the potential of the core flood line butt welds to represent a safety concern by deferral until 1R21 is very small. In addition, the EPRI MRP performed a risk assessment for failures of Alloy82/182 butt welds in MRP-116 (Reference 4). In this report, probabilistic fracture mechanics (PFM) evaluations were performed for a variety of locations of butt welds in B&W, Combustion Engineering (CE), and Westinghouse reactors. The PFM models are acceptable to the NRC in showing appropriate results.
Specifically for B&W designs, the decay heat line and the surge line were modeled. The decay heat line would be more comparable to the core flood line based on RCS temperatures. The results of the analysis when conducting normal 10-year ISI exams over a 40-year life results in only a 3.75E-09 increase in core damage frequency (CDF). In addition, a sensitivity analysis was also performed that shows that if no ISI is performed over the 40 years, the change in relative risk only increases by 2%. Therefore, the relative risk increase from the decay heat line (comparable to core flood line) butt welds is fully acceptable within the guidance of Regulatory Guide 1.174 (Reference [5]). D.Dose ReductionAs discussed above, Entergy has submitted to the NRC staff a request to deferperforming the complete reactor vessel ISI via Request for Alternative ANO 1-ISI-005 (Reference 3). The complete reactor vessel ISI requires removing the core barrel. Deferring the examination of the core flood line DM
[dissimilar metal] welds until the performance of the complete reactor vessel ISI [in the fall 2008 refueling outage] will consolidate activities and reduce personnelradiological exposure. Specifically, removing and replacing the core barrel in order to perform the weld examination involves approximately 600 mrem dose.
The additional dose involved in performing the weld inspection separate from the remainder of the inspection is expected to be approximately 200 to 400 mrem.
Therefore, performing this examination during the same planned evolution will result in a dose savings of up to 1 Rem since activities to support this examination (i.e., removing and replacing the core barrel and setup and performing the core flood line DM weld examination) will be performed only once rather than twice. E.Evaluation of Core Flood Alloy 600 Mitigation Entergy is evaluating mitigation of the Alloy 600 core flood nozzle DM weld during 1R21. As discussed above, these welds are only readily accessible from the inside diameter (ID) of the reactor vessel after the core barrel has been removed. Due to the relatively new technology to perform ID weld inlay mitigation and given the specific ANO-1 core flood line configuration, Entergy will utilize the additional time to 1R21 to investigate inlay mitigation design, analysis, and tooling. 3.5Staff EvaluationInservice inspection of RV and piping pressure-retaining welds helps to ensure structuralintegrity by identifying flaw growth before flaws become large enough to represent challenges to pressure boundary integrity. The licensee provided a summary of examinations performed on the reactor core flood line nozzle-to-safe end welds in 1995. These examination findings show that with 100% coverage obtained, no reportable indications were found and no indications are currently being monitored. Therefore, the NRC staff agrees with the licensee's assessment that the prior examinations were of sufficient quality to identify any significant flaws that would challenge reactor vessel and piping pressure-retaining weld integrity.The licensee discussed the safety significance of core flood line inspection deferral for oneoutage. They stated that the incidence of PWSCC in PWR primary coolant systems is due to tensile stresses and the operating temperature. The primary sources of tensile stresses are weld residual stresses and applied operating stresses (such as pressure and temperature). For materials of equal PWSCC susceptibility with equal applied stresses, the time to crack initiation is a function of operating temperature. Locations that operate at higher temperatures, such as pressurizers, typically exhibit cracking much sooner than locations that operate at lower temperatures, such as cold legs. Therefore, Alloy 600 butt welds that operate closer to cold-leg temperatures are generally much less susceptible to PWSCC than those that are close to RCS operating temperatures. The core flood lines operate at temperatures that are only slightly above the cold-leg temperature (577 °F). Critical flaw sizes for a through-wall flaw were determined using the methodology of ASME Code,Section XI. The critical flaw sizes were determined for axial and circumferential through-wall flaw lengths and for a 75-percent through-wall circumferential flaw depth. The core flood nozzle critical flaw sizes are as follows:
axial flaw length - 22.3 inches, circumferential flaw length - 20.7 inches, and circumferential flaw depth - 75 percent. Using the critical flaw data, crack growth analyses were performed under both PWSCC conditions and fatigue. The results of these analyses show that the time forpropagation of a flaw from an initial flaw size that produces an applied stress intensity factor equal to the PWSCC threshold stress intensity factor until 75-percent through-wall would take greater than 40 years. The time from identification of a 1-gpm leak until it reached a critical flaw size would be in excess of 70 years. The core flood lines were inspected in 1995 during the last 10-year inspection interval and no flaws were identified in these welds. Therefore, the staff agrees with the licensee's assessment that the potential of the core flood line butt welds to represent a safety concern by deferral until 1R21 is very small. In addition, a risk assessment for failures of Alloy 82/182 butt welds was performed. In thisassessment, PFM evaluations were performed for a variety of locations of butt welds.
Specifically for ANO-1 type designs, the decay-heat line and the surge line were modeled. The decay-heat line would be more comparable to the core flood line based on RCS temperatures.
The results of the analysis when conducting normal 10-year ISI exams over a 40-year life results in only a 3.75E-09 increase in CDF. In addition, a sensitivity analysis was also performed that shows that if no ISI is performed over the 40 years, the change in relative risk only increases by 2 percent. Therefore, the relative risk increase from the decay-heat line (comparable to core flood line) butt welds is fully acceptable within the guidance of Regulatory Guide 1.174 and this assessment is acceptable to the staff.As discussed above, Entergy has submitted to the NRC staff a request to defer performing thecomplete reactor vessel ISI via Request for Alternative ANO 1-ISI-005. The complete reactor vessel ISI requires removal of the core barrel. Deferring the examination of the core flood line dissimilar metal welds until the performance of the complete reactor vessel ISI will consolidate activities and reduce personnel radiological exposure. Specifically, removing and replacing thecore barrel in order to perform the weld examination involves approximately 600 mrem dose.
The additional dose involved in performing the weld inspection separate from the remainder of the inspection is expected to be approximately 200 to 400 mrem. Therefore, performing this examination at the same time will result in a dose savings of up to 1 Rem since activities to support this examination (i.e., removing and replacing the core barrel and setup and performing the core flood line DM weld examination) will be performed only once rather than twice. Staff agrees with the licensee that by performing the same additional work during two separate outages, which could incur additional dose, a hardship condition is created which can be avoided by performing all of the inspections during the one refueling outage (1R21).In summary, the NRC staff concurs with the licensee's assessment that their reactor core floodline nozzle-to-safe end welds have a low likelihood of having significant flaws and that there is a low likelihood of experiencing a severe event during the proposed extension period. Therefore, the staff believes that performance of ISI of the reactor core flood line nozzle-to-safe end welds during the next refueling outage (1R20) would not provide an additional level of safety or quality in comparison to deferring the examination for one refueling cycle (1R21). The staff finds that the risk associated with the one-cycle extension of the examination interval is sufficiently smalland coupled with the dose savings that would result, the alternative provides reasonable assurance of structural integrity and that compliance with the specified requirements of ASME Code,Section XI, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0CONCLUSION
Based on the above evaluation, the NRC staff concludes that the licensee's proposedalternative provides reasonable assurance of structural integrity and that compliance with the specified requirements of ASME Code,Section XI, 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 NRC staff authorizes the extension of the third 10-year ISI interval not to exceed the end of the fall 2008 refueling outage, to complete the ANO-1 reactor core flood-line nozzle-to-safe end welds examinations.All other requirements of the ASME Code for which relief has not been specifically requestedremain applicable including third-party review by the Authorized Nuclear Inservice Inspector.
5.0REFERENCES
1.Alloy 82/182 Pipe Butt Weld Safety Assessment for US PWR Plant Designs: Babcock & Wilcox Design Plants (MRP-112), EPRI (October 2004).2.Alloy 82/182 Pipe Butt Weld Safety Assessment for U.S. PWR Plant Designs(MRP-113), EPRI (July 2004).3.Entergy Operations, Inc., letter CNRO-2006-00024, Request for AlternativeANO-ISI-005 - Proposed Alternative to Extend the Third Inservice Inspection Interval for Reactor Vessel Inservice Examinations, dated April 24, 2006.4.Probabilistic Risk Assessment of Alloy 82/182 Piping Butt Welds (MRP-116),EPRI (August 2004).5.NRC Regulatory Guide 1.174, Revision 1, An Approach for Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plan Specific Changes to the Licensing Basis, November 2002.Principal Contributor: Edward V. Andruszkiewicz Date: February 1, 2007