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Category:Inservice/Preservice Inspection and Test Report
MONTHYEARGO2-23-097, In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26)2023-09-0606 September 2023 In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26) ML22278B0972022-10-0505 October 2022 Fourth Ten-Year Interval Inservice Inspection Program Relief Request 4ISl-11 GO2-21-107, In-Service Inspection Summary Report for the Twenty-Fifth Refueling (R25)2021-08-26026 August 2021 In-Service Inspection Summary Report for the Twenty-Fifth Refueling (R25) GO2-21-020, General Station - Fourth Ten-Year Interval Inservice Inspection (ISI) Program 4ISI-10, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2021-03-0101 March 2021 General Station - Fourth Ten-Year Interval Inservice Inspection (ISI) Program 4ISI-10, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI GO2-20-146, Erratum to Fourth Ten-Year Inservice Inspection (ISI) Program Relief Request 4ISI-092020-11-17017 November 2020 Erratum to Fourth Ten-Year Inservice Inspection (ISI) Program Relief Request 4ISI-09 ML20114E2352020-04-22022 April 2020 Columbia,Fourth Ten-Year Interval Inservice Inspection Program Relief Request 4ISI-09 ML20057E9542020-02-26026 February 2020 Fourth Ten-Year Interval Inservice Inspection Program Relief Request 4ISI-08 GO2-19-127, In-Service Inspection Summary Report for the Twenty-Fourth Refueling Outage (R24)2019-09-11011 September 2019 In-Service Inspection Summary Report for the Twenty-Fourth Refueling Outage (R24) GO2-17-153, Inservice Inspection Summary Report for the Twenty-Third Refueling Outage (R23)2017-09-12012 September 2017 Inservice Inspection Summary Report for the Twenty-Third Refueling Outage (R23) GO2-16-068, Re-Submittal of Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISI-072016-04-26026 April 2016 Re-Submittal of Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISI-07 GO2-16-008, Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISIS-042016-02-0404 February 2016 Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISIS-04 GO2-15-157, Fourth Ten-Year Interval Inservice Inspection (ISI) Program Request 4ISI-022015-12-10010 December 2015 Fourth Ten-Year Interval Inservice Inspection (ISI) Program Request 4ISI-02 GO2-15-136, In-Service Inspection Summary Report for the Twenty-Second Refueling Outage (R22)2015-09-22022 September 2015 In-Service Inspection Summary Report for the Twenty-Second Refueling Outage (R22) GO2-15-055, Submittal of Fourth 10-Year Inservice Testing (IST) Program Plan2015-04-23023 April 2015 Submittal of Fourth 10-Year Inservice Testing (IST) Program Plan GO2-15-024, Withdrawal of Inservice Inspection (ISI) Program Request 3ISI-162015-02-18018 February 2015 Withdrawal of Inservice Inspection (ISI) Program Request 3ISI-16 GO2-14-152, Erratum for Relief Requests for the Fourth Ten-Year Interval Inservice Testing Program2014-10-23023 October 2014 Erratum for Relief Requests for the Fourth Ten-Year Interval Inservice Testing Program GO2-14-017, Inservice Inspection (ISI) Program Request 3 ISI-142014-02-0606 February 2014 Inservice Inspection (ISI) Program Request 3 ISI-14 GO2-13-133, In-Service Inspection Summary Report for the Twenty-First Refueling Outage (R21)2013-09-18018 September 2013 In-Service Inspection Summary Report for the Twenty-First Refueling Outage (R21) ML13249A1602013-08-28028 August 2013 Submittal of Revision 1 of the Third Ten-Year Interval Pump and Valve Inservice Testing (IST) Program Plan GO2-11-193, In-Service Inspection Summary Report for the Twentieth Refueling Outage (R20)2011-12-14014 December 2011 In-Service Inspection Summary Report for the Twentieth Refueling Outage (R20) GO2-09-134, Inservice Inspection Summary Report for Refueling Outage R-192009-09-22022 September 2009 Inservice Inspection Summary Report for Refueling Outage R-19 GO2-07-141, Inservice Inspection Summary Report for Refueling Outage R-182007-09-18018 September 2007 Inservice Inspection Summary Report for Refueling Outage R-18 ML0708707152007-03-22022 March 2007 Response to Request for Additional Information Regarding the Third Ten-Year Interval Inservice Testing Program ML0636201472006-12-14014 December 2006 Columbia Generating Station, Second Ten-Year Interval ISI Request 2ISI-32 GO2-06-062, Correction to Third Ten-Year Interval Inservice Inspection Program Plan2006-04-27027 April 2006 Correction to Third Ten-Year Interval Inservice Inspection Program Plan GO2-05-196, Submittal of the Third Ten-Year Interval Inservice Inspection Program Plan and 10 CFR 50.55a Requests 3ISI-01 Through 3ISI-072005-12-15015 December 2005 Submittal of the Third Ten-Year Interval Inservice Inspection Program Plan and 10 CFR 50.55a Requests 3ISI-01 Through 3ISI-07 ML0527103602005-09-15015 September 2005 Enclosure, Columbia Generating, Analytical Evaluation of Inservice Inspection Examination Results GO2-05-153, Analytical Evaluation of Inservice Inspection Examination Results2005-09-15015 September 2005 Analytical Evaluation of Inservice Inspection Examination Results ML0525801582005-09-0606 September 2005 Inservice Inspection Summary Report for Refueling Outage R-17 GO2-04-065, Request to Incorporate ASME Code Case N-663 in Columbia'S Second Ten-Year Interval Inservice Inspection Program Plan2004-04-0808 April 2004 Request to Incorporate ASME Code Case N-663 in Columbia'S Second Ten-Year Interval Inservice Inspection Program Plan GO2-03-147, Inservice Inspection Summary Report for Refueling Outage R-162003-09-16016 September 2003 Inservice Inspection Summary Report for Refueling Outage R-16 2023-09-06
[Table view] Category:Letter type:GO
MONTHYEARGO2-24-004, License Amendment Request to Revise Emergency Plan2024-01-30030 January 2024 License Amendment Request to Revise Emergency Plan GO2-24-003, Relief Requests for the Columbia Generating Station Fifth Ten-Year Interval Inservice Testing2024-01-29029 January 2024 Relief Requests for the Columbia Generating Station Fifth Ten-Year Interval Inservice Testing GO2-24-005, Docket No. 50-397 Supplement to Reply to a Notice of Violation; EA-21-1702024-01-0808 January 2024 Docket No. 50-397 Supplement to Reply to a Notice of Violation; EA-21-170 GO2-23-136, Guarantee of Payment of Deferred Premium2023-12-20020 December 2023 Guarantee of Payment of Deferred Premium GO2-23-135, Notice of Readiness for Supplemental Inspection2023-12-14014 December 2023 Notice of Readiness for Supplemental Inspection GO2-23-130, Reply to a Notice of Violation; EA-23-0542023-12-14014 December 2023 Reply to a Notice of Violation; EA-23-054 GO2-23-105, Licensing Basis Document Update and Biennial Commitment Change Report2023-12-12012 December 2023 Licensing Basis Document Update and Biennial Commitment Change Report GO2-23-107, Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation2023-12-0505 December 2023 Application to Revise Technical Specifications to Adopt TSTS-584, Eliminate Automatic RWCU System Isolation on SLC Initiation GO2-23-126, Docket No. 50-397 Voluntary Response to Regulatory Issue Summary 2023-012023-11-28028 November 2023 Docket No. 50-397 Voluntary Response to Regulatory Issue Summary 2023-01 GO2-23-121, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-27027 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation GO2-23-097, In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26)2023-09-0606 September 2023 In-Service Inspection Summary Report for the Twenty-Sixth Refueling Outage (R26) GO2-23-100, Technical Specification Section 5.6.4 Post Accident Monitoring Instrumentation 14-Day Report for Inoperable Suppression Pool Level Indication2023-08-31031 August 2023 Technical Specification Section 5.6.4 Post Accident Monitoring Instrumentation 14-Day Report for Inoperable Suppression Pool Level Indication GO2-23-056, Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2023-08-29029 August 2023 Application to Revise Technical Specifications to Adopt TSTF-230, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling GO2-23-093, Supplement to Reply to a Notice of Violation: EA-21-1702023-07-27027 July 2023 Supplement to Reply to a Notice of Violation: EA-21-170 GO2-23-068, Notification of Completion of Commitments Required Prior to Entry Into the Period of Extended Operation2023-07-19019 July 2023 Notification of Completion of Commitments Required Prior to Entry Into the Period of Extended Operation GO2-23-090, Reply to a Notice of Violation2023-07-12012 July 2023 Reply to a Notice of Violation GO2-23-073, Notification of NPDES Permit Issuance2023-06-26026 June 2023 Notification of NPDES Permit Issuance GO2-23-072, Cycle 27 Core Operating Limits Report2023-06-0505 June 2023 Cycle 27 Core Operating Limits Report GO2-23-055, Independent Spent Fuel Storage Installation, 2022 Annual Radiological Environmental Operating Report2023-05-15015 May 2023 Independent Spent Fuel Storage Installation, 2022 Annual Radiological Environmental Operating Report GO2-23-006, License Amendment Request to Clean-Up Operating License and Appendix a Technical Specifications2023-05-0101 May 2023 License Amendment Request to Clean-Up Operating License and Appendix a Technical Specifications GO2-23-053, 2022 Annual Environmental Operating Report2023-04-26026 April 2023 2022 Annual Environmental Operating Report GO2-23-057, 2022 Annual Radioactive Effluent Release Report2023-04-25025 April 2023 2022 Annual Radioactive Effluent Release Report GO2-23-012, Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position2023-03-27027 March 2023 Application to Revise Technical Specifications to Adopt Tstf-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position GO2-23-041, Generation Station - Level of Financial Protection - Annual Reporting Requirement2023-03-15015 March 2023 Generation Station - Level of Financial Protection - Annual Reporting Requirement GO2-23-038, Plant Decommissioning Fund Status Report2023-03-15015 March 2023 Plant Decommissioning Fund Status Report GO2-23-034, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.462023-03-13013 March 2023 Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.46 GO2-23-035, Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-03-0909 March 2023 Supplement to License Amendment Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling GO2-23-019, Emergency Plan - Summary of Changes and Analysis for Revision 68 of the EP-012023-02-0808 February 2023 Emergency Plan - Summary of Changes and Analysis for Revision 68 of the EP-01 GO2-22-131, Guarantee of Payment of Deferred Premium2022-12-19019 December 2022 Guarantee of Payment of Deferred Premium GO2-22-122, Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-11-28028 November 2022 Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-076, Status of License Renewal Commitments2022-10-31031 October 2022 Status of License Renewal Commitments GO2-22-110, Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactor2022-10-17017 October 2022 Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactor GO2-22-096, Supplement to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-10-0404 October 2022 Supplement to License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-085, Docket No. 50-397, Evacuation Time Estimate Analysis2022-08-30030 August 2022 Docket No. 50-397, Evacuation Time Estimate Analysis GO2-22-075, Response to Request for Additional Information Related to Revised Pressure-Temperature Limit Curves2022-07-11011 July 2022 Response to Request for Additional Information Related to Revised Pressure-Temperature Limit Curves GO2-22-072, Summary of Changes and Analysis for Revision 57 of PPM 13.14.4 - Emergency Equipment Maintenance and Testing2022-07-11011 July 2022 Summary of Changes and Analysis for Revision 57 of PPM 13.14.4 - Emergency Equipment Maintenance and Testing GO2-22-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process2022-05-25025 May 2022 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Colling Using the Consolidated Line Item Improvement Process GO2-22-062, Registration of Spent Fuel Cask Use2022-05-19019 May 2022 Registration of Spent Fuel Cask Use GO2-22-049, Supplement to Licensing Amendment Request to Adopt 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Rectors2022-05-10010 May 2022 Supplement to Licensing Amendment Request to Adopt 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Rectors GO2-22-054, Independent Spent Fuel Storage Installation, 2021 Annual Radiological Environmental Operating Report2022-05-0909 May 2022 Independent Spent Fuel Storage Installation, 2021 Annual Radiological Environmental Operating Report GO2-22-056, Registration of Spent Fuel Cask Use2022-04-21021 April 2022 Registration of Spent Fuel Cask Use GO2-22-055, 2021 Annual Environmental Operating Report2022-04-20020 April 2022 2021 Annual Environmental Operating Report GO2-22-048, 2021 Annual Radioactive Effluent Release Report2022-04-15015 April 2022 2021 Annual Radioactive Effluent Release Report GO2-22-045, Registration of Spent Fuel Cask Use2022-04-0606 April 2022 Registration of Spent Fuel Cask Use GO2-22-044, Level of Financial Protection - Annual Reporting Requirement2022-03-21021 March 2022 Level of Financial Protection - Annual Reporting Requirement GO2-22-038, Docket No. 50-397, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.462022-03-0909 March 2022 Docket No. 50-397, Report of Changes or Errors in Emergency Core Cooling System Loss of Coolant Accident Analysis Models Pursuant to 10 CFR 50.46 GO2-22-031, Response to Audit Plan and Online Reference Portal Request2022-02-15015 February 2022 Response to Audit Plan and Online Reference Portal Request GO2-22-011, Docket No. 72-35; Independent Spent Fuel Storage Installation, 2021 Annual Effluent Release Report2022-02-0808 February 2022 Docket No. 72-35; Independent Spent Fuel Storage Installation, 2021 Annual Effluent Release Report GO2-22-001, License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-02-0303 February 2022 License Amendment Request to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b GO2-22-012, Independent Spent Fuel Storage Installation - Biennial 50.59/72.48 Report2022-01-24024 January 2022 Independent Spent Fuel Storage Installation - Biennial 50.59/72.48 Report 2024-01-08
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X Alex Javorik ENERGY Vice P.O. BoxPresident, 968, MailEngineering Drop PE20 J NORTHW EST Richland, WA 99352-0968 Ph. 509-377-8555 F. 509-377-4317 aljavorik@energy-northwest.com February 6, 2014 G02-14-017 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397; INSERVICE INSPECTION (ISI) PROGRAM REQUEST 3 IS1-14
References:
(1) Letter dated April 19, 2013, Sher Bahadur (NRC) to Dennis Madison (BWRVIP), "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP-241) Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii" (2) Letter dated December 19, 2007, Matthew A. Mitchell (NRC), to Rick Libra (BWRVIP), "Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius"'
Dear Sir or Madam:
Section 50.55a of Title 10 of the Code of Federal Regulations requires that In Service Inspection (ISI) of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 piping be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. Pursuant to 10 CFR 50.55a(a)(3)(i) Energy Northwest hereby requests NRC approval of the proposed alternate to ASME Section XI, Sub Article IWB-2500 to allow reduced percentage requirements for nozzle to vessel weld and inner radius examinations while still providing an acceptable level of quality and safety. This alternative is requested for the third ten-year interval ISI program at Columbia Generating Station. The details of the 10 CFR 50.55a request are enclosed as Attachment 1.
Approval of request 31S1-14 would allow reduced examination requirements through application of American Society of Mechanical Engineers (ASME) Code Case N-702. The NRC provided a Safety Evaluation approving the generic technical bases and acceptability criteria for application of Code Case N-702, which Energy Northwest has followed as detailed in the attached request.
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 Page 2 Energy Northwest requests approval by February 28, 2015 to accommodate application of the request during the next refueling outage. If approved, the use of Code Case N-702 at Columbia Generating Station would result in significantly reduced radiological dose to personnel while providing acceptable level of quality and safety.
There are no new commitments made in this submittal. If you have any questions or require additional information, please contact Lisa Williams at 509-377-8148.
Respectfully, A. L. Javorik Vice President, Engineering
Attachment:
(1) 10 CFR 50.55a Request Number 31S1-14 cc: NRC RIV Regional Administrator CF Lyon, NRC NRR Project Manager NRC Sr. Resident Inspector - 988C AJ Rapacz - BPA/1 399 WA Horin - Winston & Strawn
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 1 of 5 10 CFR 50.55a Request Number 31S1-1 4 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
--Alternative Provides Acceptable Level of Quality and Safety--
- 1. ASME Code Component(s) Affected The components in the following table are affected by this request.
Table 1 Identification Code Item Number Description Category Number N1-0 RRC NOZZLE TO VESSEL WELD @ 0 Deg B-D B3.90 N1 IR RRC NOZZLE INNER RADIUS @ 0 Deg B-D B3.1 00 N1 -180 RRC NOZZLE TO VESSEL WELD @ 180 Deg B-D B3.90 N1 -180-IR RRC NOZZLE INNER RADIUS @ 180 Deg B-D B3.100 N2-30 RRC NOZZLE TO VESSEL WELD @ 30 Deg B-D B3.90 N2-30-IR RRC NOZZLE INNER RADIUS @ 30 Deg B-D B3.100 N2-60 RRC NOZZLE TO VESSEL WELD @ 60 Deg B-D B3.90 N2-60-IR RRC NOZZLE INNER RADIUS @ 60 Deg B-D B3.100 N2-90 RRC NOZZLE TO VESSEL WELD @ 90 Deg B-D B3.90 N2-90-IR RRC NOZZLE INNER RADIUS @ 90 Deg B-D B3.100 N2-120 RRC NOZZLE TO VESSEL WELD @ 120 Deg B-D B3.90 N2-120-IR RRC NOZZLE INNER RADIUS @ 120 Deg B-D B3.100 N2-150 RRC NOZZLE TO VESSEL WELD @ 150 Deg B-D B3.90 N2-150-IR RRC NOZZLE INNER RADIUS @ 150 Deg B-D B3.100 N2-210 RRC NOZZLE TO VESSEL WELD @ 210 Deg B-D B3.90 N2-210-IR RRC NOZZLE INNER RADIUS @ 210 Deg B-D B3.100 N2-240 RRC NOZZLE TO VESSEL WELD @ 240 Deg B-D B3.90 N2-240-IR RRC NOZZLE INNER RADIUS @ 240 Deg B-D B3.100 N2-270 RRC NOZZLE TO VESSEL WELD @ 270 Deg B-D B3.90 N2-270-IR RRC NOZZLE INNER RADIUS @ 270 Deg B-D B3.100 N2-300 RRC NOZZLE TO VESSEL WELD @ 300 Deg B-D B3.90 N2-300-IR RRC NOZZLE INNER RADIUS @ 300 Deg B-D B3.100 N2-330 RRC NOZZLE TO VESSEL WELD @ 330 Deg B-D B3.90 N2-330-IR RRC NOZZLE INNER RADIUS @ 330 Deg B-D B3.100 RRC Reactor Recirculation
2. Applicable Code Edition and Addenda
Columbia Generating Station (Columbia) ASME Section Xl Code is the 2001 Edition through the 2003 Addenda (Reference 3). Additionally, for ultrasonic examinations, Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xv).
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 2 of 5
3. Applicable Code Requirement
Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Subsection IWB, Table IWB-2500-1, "Examination Category B-D Full Penetration Welds of Nozzles in Vessels - Inspection Program B," Item Numbers B3.90 and B3.100, respectively. The method of examination is volumetric. For the extent of examination, all nozzles with full penetration weld to the vessel shell (or head) and integrally cast nozzles must be examined each interval.
4. Reason for Request
The identified nozzles (see Table 1) are scheduled for examination prior to the end of the current inspection interval for Columbia. The interval began in December 2005 and will end in December 2015. The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 1250 mRem for the next refuel outage.
- 5. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzle assemblies in Table 1 above. As an alternative, for all welds and inner radii identified in Table 1, Columbia proposes to examine a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each type and nominal pipe size, in accordance with Code Case N-702 (Reference 4). The proposed schedule to complete the minimum of 25% from each group is presented in Table 2 below. Columbia has completed three assembly (nozzle and inner radii) examinations already this interval using automated ultrasonic testing techniques. Greater than 97% coverage was achieved on each location and no relevant indications were reported. With the acceptance of the relief request only one nozzle-to-vessel weld and one inner radius would be required to meet the 25% for the current interval.
Table 2 Total Number to Group Number be Examined Comments RRC Outlet (N1) 2 1 1 Completed in R19 RRC Inlet (N2) 10 3 2 Completed in R19, 1 scheduled in R22 R19 Refuel Outage 19 (2009)
R22 Refuel Outage 22 (2015)
Basis for Use Boiling Water Reactor (BWR) Vessel Internals Project (BWRVIP) topical report BWRVIP-241, "BWR Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (hereafter referred to as BWRVIP-241)
(Reference 2), documents supplemental analyses for BWR reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation (SE) for the BWRVIP-1 08NP report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii" (Reference 1). The BWRVIP-1 08NP report contains the technical basis
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 3 of 5 supporting American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. Based on the two evaluations (BWRVIP-241 and BWRVIP-1 08NP), the failure probabilities due to a low temperature over pressure (LTOP) event at the nozzle blend radius region and the nozzle-to-vessel shell weld for Columbia recirculation nozzles are very low and meet the NRC safety goal.
Based on the results of this evaluation, the report concluded that the inspection of 25% of each nozzle type is technically justified as per Code Case N-702.
EPRI report BWRVIP-241 received a final NRC SE on April 19, 2013 (ML13071A240)
(Reference 6). In the SE, Section 5.0 "Conditions and Limitations" indicates that each licensee who plans to request relief from the ASME Code, Section Xl requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:
(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115OF per hour.
The Columbia Technical Specification limits the heatup/cooldown rate to less than or equal to 100 OF in any one hour period.
For the Recirculation Inlet Nozzles (N2) the following criteria must be met:
(2) (pr/t)/CRPV*1l.15.
(3) [p(r02 +ri2)/(ro2-rI2)]/CNozzLE<1.47.
For the Recirculation Outlet Nozzles (N1) the following criteria must be met:
(4) (pr/t)/CRPV<1.15.
(5) [p (ro2+ri 2)/(r0 2-ri2 )]/CNozZLE<I .59.
The terms to be used in the NRC SE Section 5 applicability evaluations criteria 2-4 are:
CaPv = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 19332 CNOZZLE = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 1637 CRPV = recirculation outlet nozzles N1 (from BWRVIP-241 model) = 16171 CNOZZLE = recirculation outlet nozzles N1 (from BWRVIP-241 model) = 1977 p = RPV normal operating pressure (psi) r = RPV inner radius (inch) t = RPV wall thickness (inch) r= Nozzle inner radius (inch) ro = Nozzle outer radius (inch)
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 4 of 5 Table 3 below summarizes these results.
Table 3 Outlet Nozzles (N1)
Criteria (4) Criteria (5)
Cnozzle CRPV p r t(min) ri ro <1.15 <1.59 1977 16171 1020 127 9.5 10.8 15.4 0.84 1.51 Inlet Nozzles (N2)
Criteria (2) Criteria (3)
Cnozzle CRPV p r t(min) ri r, <1.15 <51.47 1637 19332 1020 127 9.5 5.8 10.0 0.71 1.25 Based upon the above information, the RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1 meet the SE criteria and therefore Code Case N-702 is applicable. In addition, the reactor vessel is low alloy steel plate specification SA-533 grade B class I. The nozzle is low alloy steel forging specification SA-508 class 2. The weld metal used in the welds specified in Table 1 is carbon/low alloy steel and the welds are outside the beltline fluence region. Therefore, use of Code Case N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) for all RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1.
- 6. Duration of Proposed Alternative The duration of this request is for the third inservice inspection interval ending December 12, 2015.
- 7. Precedents None
- 8. References
- 1. BWRVIP-IO8NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007.1016123.
- 2. BWRVIP-241: BWR Vessel and Internals Project, ProbabilisticFractureMechanics Evaluation for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2010. 1021005.
- 3. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plants," 2001 Edition through 2003 Addenda.
- 4. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 5 of 5
- 5. Matthew A. Mitchell, Office of Nuclear Reactor Regulation, to Rick Libra, BWRVIP Chairman, "Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-1 08)',"
December 19, 2007.
- 6. Sher Bahadur, Office of Nuclear Reactor Regulation, to Dennis Madison, BWRVIP Chairman, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP-241) Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii (TAC NO.
ME6328)" April 19, 2013.