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Category:Inservice/Preservice Inspection and Test Report
MONTHYEARDCL-23-009, Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage2023-02-22022 February 2023 Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage DCL-22-091, Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage2022-12-20020 December 2022 Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage DCL-22-060, Owner'S Activity Report for Unit 1 Twenty-Third Refueling Outage2022-07-21021 July 2022 Owner'S Activity Report for Unit 1 Twenty-Third Refueling Outage ML22138A4362022-05-18018 May 2022 ASME Section XI Inservice Inspection Program Plan - Fourth 10-Year Inspection Interval, Revision 1 ML21307A0012021-11-15015 November 2021 Review of the Fall 2020 Steam Generator Tube Inservice Inspection Report DCL-21-055, Owner'S Activity Report for Unit 2 Twenty-Second Refueling Outage2021-07-19019 July 2021 Owner'S Activity Report for Unit 2 Twenty-Second Refueling Outage DCL-21-008, Owner'S Activity Report for Unit 1 Twenty-Second Refueling Outage2021-01-27027 January 2021 Owner'S Activity Report for Unit 1 Twenty-Second Refueling Outage DCL-20-039, One Hundred Eighty Day Steam Generator Report for Diablo Canyon Power Plant Unit 2 Twenty-First Refueling Outage2020-05-13013 May 2020 One Hundred Eighty Day Steam Generator Report for Diablo Canyon Power Plant Unit 2 Twenty-First Refueling Outage ML20064F5782020-03-0404 March 2020 Owner'S Activity Report for Unit 2 Twenty-First Refueling Outage DCL-19-084, ASME Section XI Lnservice Inspection Program Request for Alternative NDE-RCSSE-2R22 Use of Alternate Sizing Qualification Criteria Through a Protective Clad Layer2019-10-31031 October 2019 ASME Section XI Lnservice Inspection Program Request for Alternative NDE-RCSSE-2R22 Use of Alternate Sizing Qualification Criteria Through a Protective Clad Layer DCL-19-049, Owner'S Activity Report for Unit 1 Twenty-first Refueling Outage2019-06-13013 June 2019 Owner'S Activity Report for Unit 1 Twenty-first Refueling Outage DCL-18-105, Submittal of the Fourth Ten-Year Interval Inservice Testing (1ST) Program Plan, Revision 12018-12-0505 December 2018 Submittal of the Fourth Ten-Year Interval Inservice Testing (1ST) Program Plan, Revision 1 DCL-18-048, Owner'S Activity Report for Unit 2 Twentieth Refueling Outage2018-06-19019 June 2018 Owner'S Activity Report for Unit 2 Twentieth Refueling Outage DCL-17-028, ASME Section XI Inservice Inspection Program Plan - Fourth 10-Year Inspection Interval2017-04-18018 April 2017 ASME Section XI Inservice Inspection Program Plan - Fourth 10-Year Inspection Interval DCL-16-116, ASME Section XI Inservice Inspection Program Relief Request NDE-SIF-U2 Due to Impracticality of Full Examination Volume Coverage Requirements2016-11-10010 November 2016 ASME Section XI Inservice Inspection Program Relief Request NDE-SIF-U2 Due to Impracticality of Full Examination Volume Coverage Requirements DCL-16-115, ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternative Depth Sizing Criteria2016-11-10010 November 2016 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternative Depth Sizing Criteria DCL-16-086, Owner'S Activity Report for Unit 2 Nineteenth Refueling Outage2016-08-31031 August 2016 Owner'S Activity Report for Unit 2 Nineteenth Refueling Outage DCL-16-056, One Hundred Eighty-Day Steam Generator Report for Nineteenth Refueling Outage2016-05-0202 May 2016 One Hundred Eighty-Day Steam Generator Report for Nineteenth Refueling Outage DCL-16-024, Submittal of the Fourth Ten-Year Interval Inservice Testing (IST) Program Plan2016-03-0202 March 2016 Submittal of the Fourth Ten-Year Interval Inservice Testing (IST) Program Plan DCL-16-018, Owner'S Activity Report for Nineteenth Refueling Outage2016-02-0303 February 2016 Owner'S Activity Report for Nineteenth Refueling Outage DCL-15-116, Submittal of 10 CFR 50.55a Request FLIG-U1, Request for Extension of Third Lnservice Inspection Interval for Performing Reactor Vessel Stud Hole Ligament Examinations2015-10-0707 October 2015 Submittal of 10 CFR 50.55a Request FLIG-U1, Request for Extension of Third Lnservice Inspection Interval for Performing Reactor Vessel Stud Hole Ligament Examinations DCL-15-106, ASME Section XI Inservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements2015-09-0303 September 2015 ASME Section XI Inservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements DCL-15-054, One Hundred Eighty-Day Steam Generator Report for Diablo Canyon Power Plant, Unit 2, Eighteenth Refueling Outage2015-04-29029 April 2015 One Hundred Eighty-Day Steam Generator Report for Diablo Canyon Power Plant, Unit 2, Eighteenth Refueling Outage DCL-15-048, ASME Section XI Inservice Inspection Program Request for Relief NDE-PNS-U2A to Allow Use of Alternate Examination Volume Coverage Requirements2015-04-0909 April 2015 ASME Section XI Inservice Inspection Program Request for Relief NDE-PNS-U2A to Allow Use of Alternate Examination Volume Coverage Requirements DCL-14-053, Owner'S Activity Report for Eighteenth Refueling Outage2014-06-11011 June 2014 Owner'S Activity Report for Eighteenth Refueling Outage DCL-14-027, Request for Relief from the Requirements of Appendix IX of ASME Section XI, 2001 Edition with 2003 Addendum2014-03-28028 March 2014 Request for Relief from the Requirements of Appendix IX of ASME Section XI, 2001 Edition with 2003 Addendum DCL-13-063, Inservice Inspection Report for Seventeenth Refueling Outage2013-06-13013 June 2013 Inservice Inspection Report for Seventeenth Refueling Outage DCL-12-089, Inservice Inspection Report for Unit 1 Seventeenth Refueling Outage2012-09-13013 September 2012 Inservice Inspection Report for Unit 1 Seventeenth Refueling Outage DCL-12-007, Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirement for Class 1 and 2 Piping Welds2012-01-20020 January 2012 Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirement for Class 1 and 2 Piping Welds ML12025A3042011-12-28028 December 2011 Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 201 Through Drawing No. 102028, Sheet 38 ML12025A3052011-12-28028 December 2011 Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Drawing No. 102028, Sheet 39, Through Drawing No. 104628, Sheet 46 ML12025A3032011-12-28028 December 2011 Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 101 Through Page 200 ML12025A3022011-12-28028 December 2011 Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Cover Page Through Page 100 DCL-12-006, Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Drawing No. 102028, Sheet 39, Through Drawing No. 104628, Sheet 462011-12-28028 December 2011 Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Drawing No. 102028, Sheet 39, Through Drawing No. 104628, Sheet 46 DCL-12-006, Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 201 Through Drawing No. 102028, Sheet 382011-12-28028 December 2011 Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 201 Through Drawing No. 102028, Sheet 38 DCL-12-006, Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 101 Through Page 2002011-12-28028 December 2011 Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Page 101 Through Page 200 DCL-12-006, Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Cover Page Through Page 1002011-12-28028 December 2011 Diablo Canyon - Inservice Inspection Program Plan, Third 10-Year Inspection Interval, Revision 1, Cover Page Through Page 100 DCL-11-093, Inservice Inspection Report for Sixteenth Refueling Outage2011-09-0101 September 2011 Inservice Inspection Report for Sixteenth Refueling Outage DCL-11-053, One Hundred Eighty-Day Steam Generator Report - Sixteenth Refueling Outage2011-04-21021 April 2011 One Hundred Eighty-Day Steam Generator Report - Sixteenth Refueling Outage DCL-10-157, Snubber Visual Examination and Functional Testing Related to the Inservice Inspection Program Third 10-Year Interval2010-12-21021 December 2010 Snubber Visual Examination and Functional Testing Related to the Inservice Inspection Program Third 10-Year Interval DCL-10-051, ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R16 to Allow Use of Alternate Sizing Qualification Criteria2010-05-17017 May 2010 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R16 to Allow Use of Alternate Sizing Qualification Criteria ML1005005362010-02-0808 February 2010 Inservice Inspection Report for Fifteenth Refueling Outage DCL-09-046, Submittal of Fifteenth Refueling Outage Inservice Inspection Report2009-06-22022 June 2009 Submittal of Fifteenth Refueling Outage Inservice Inspection Report DCL-08-103, ASME Section XI Inservice Inspection Program Relief Request NDE-Leak Path for the Unit 1, Fifteenth Refueling Outage, Third Ten-Year Inspection Interval to Allow Use of the Rules of the NRC First Revised Order, EA-03-009.2008-12-0404 December 2008 ASME Section XI Inservice Inspection Program Relief Request NDE-Leak Path for the Unit 1, Fifteenth Refueling Outage, Third Ten-Year Inspection Interval to Allow Use of the Rules of the NRC First Revised Order, EA-03-009. DCL-08-058, Transmittal of Inservice Inspection Report, Fourteenth Refueling Outage2008-07-10010 July 2008 Transmittal of Inservice Inspection Report, Fourteenth Refueling Outage DCL-07-099, ASME Section XI Inservice Inspection Program Relief Request REP-1 U2, Revision 1, and Response to Request for Additional Information2007-10-22022 October 2007 ASME Section XI Inservice Inspection Program Relief Request REP-1 U2, Revision 1, and Response to Request for Additional Information DCL-07-084, Inservice Inspection Report for Fourteenth Refueling Outage2007-08-27027 August 2007 Inservice Inspection Report for Fourteenth Refueling Outage DCL-06-099, ASME Section XI Inservice Inspection Program Relief Requests NDE-SLH U2 and NDE-LSL U22006-08-24024 August 2006 ASME Section XI Inservice Inspection Program Relief Requests NDE-SLH U2 and NDE-LSL U2 DCL-06-101, Inservice Inspection Report for Plant Thirteenth Refueling Outage2006-08-23023 August 2006 Inservice Inspection Report for Plant Thirteenth Refueling Outage DCL-06-031, Inservice Inspection Report for Unit 1 Thirteenth Refueling Outage2006-03-0303 March 2006 Inservice Inspection Report for Unit 1 Thirteenth Refueling Outage 2023-02-22
[Table view] Category:Letter type:DCL
MONTHYEARDCL-24-010, Nuclear Material Transaction Report for New Fuel2024-01-29029 January 2024 Nuclear Material Transaction Report for New Fuel DCL-24-009, Nuclear Material Transaction Report for New Fuel2024-01-17017 January 2024 Nuclear Material Transaction Report for New Fuel DCL-24-008, Schedule Considerations for Review of the DCPP License Renewal Application2024-01-17017 January 2024 Schedule Considerations for Review of the DCPP License Renewal Application DCL-24-004, Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2024-01-15015 January 2024 Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b DCL-23-129, Nuclear Material Transaction Report for New Fuel2023-12-27027 December 2023 Nuclear Material Transaction Report for New Fuel DCL-23-122, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-14014 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation DCL-23-128, Emergency Plan Update2023-12-13013 December 2023 Emergency Plan Update DCL-23-125, Core Operating Limits Report for Unit 1 Cycle 252023-12-0606 December 2023 Core Operating Limits Report for Unit 1 Cycle 25 DCL-23-121, Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System2023-11-16016 November 2023 Supplement to License Amendment Request 23-03, Revision to Technical Specification3.7.8, Auxiliary Saltwater System DCL-23-120, License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System2023-11-14014 November 2023 License Amendment Request 23-03 Revision to Technical Specification 3.7.8, Auxiliary Saltwater (Asw) System DCL-23-118, License Renewal Application2023-11-0707 November 2023 License Renewal Application DCL-2023-520, Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP)2023-10-19019 October 2023 Discharge Self-Monitoring at Diablo Canyon Power Plant (DCPP) DCL-23-103, Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System2023-10-13013 October 2023 Independent Spent Fuel Storage Installation - Withdrawal of License Amendment Request 22-01, Request for Approval of Alternative Security Measures for Early Warning System DCL-23-092, Material Status Report for the Period Ending August 31, 20232023-09-28028 September 2023 Material Status Report for the Period Ending August 31, 2023 DCL-23-077, License Amendment Request 23-02 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-27027 September 2023 License Amendment Request 23-02 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors DCL-23-083, Upcoming Meeting with Nuclear Security and Incident Response Staff2023-09-13013 September 2023 Upcoming Meeting with Nuclear Security and Incident Response Staff DCL-23-078, Nuclear Material Transaction Report for New Fuel2023-09-0606 September 2023 Nuclear Material Transaction Report for New Fuel DCL-23-070, Withdrawal of Request Regarding Senior Reactor Operator License Application2023-08-16016 August 2023 Withdrawal of Request Regarding Senior Reactor Operator License Application DCL-23-068, Nuclear Material Transaction Report for New Fuel2023-08-0909 August 2023 Nuclear Material Transaction Report for New Fuel DCL-23-066, Nuclear Material Transaction Report for New Fuel2023-08-0303 August 2023 Nuclear Material Transaction Report for New Fuel DCL-23-065, Letter DCL-23-065 Request That Exam Material Be Withheld from Public Disclosure2023-07-27027 July 2023 Letter DCL-23-065 Request That Exam Material Be Withheld from Public Disclosure DCL-23-064, Nuclear Material Transaction Report for New Fuel2023-07-26026 July 2023 Nuclear Material Transaction Report for New Fuel DCL-23-054, License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2023-07-13013 July 2023 License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b DCL-23-055, Nuclear Material Transaction Report for New Fuel2023-07-0606 July 2023 Nuclear Material Transaction Report for New Fuel DCL-23-042, Independent Spent Fuel Storage Installation, Submittal of Quality Assurance Program Description, Revision 02023-05-24024 May 2023 Independent Spent Fuel Storage Installation, Submittal of Quality Assurance Program Description, Revision 0 DCL-23-032, Independent Spent Fuel Storage Installation Docket No. 72-27, Materials License Number SNM-2514 Humboldt Bay Independent Spent Fuel Storage Installation - NRC Requested Notification of Withdrawal from Voluntary Program2023-05-22022 May 2023 Independent Spent Fuel Storage Installation Docket No. 72-27, Materials License Number SNM-2514 Humboldt Bay Independent Spent Fuel Storage Installation - NRC Requested Notification of Withdrawal from Voluntary Program DCL-23-041, 10 CFR 50.46 Annual Report of Emergency Core Cooling System Evaluation Model Changes for Peak Cladding Temperature for 20222023-05-22022 May 2023 10 CFR 50.46 Annual Report of Emergency Core Cooling System Evaluation Model Changes for Peak Cladding Temperature for 2022 DCL-23-038, Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule2023-05-15015 May 2023 Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule DCL-23-034, 2022 Annual Nonradiological Environmental Operating Report2023-05-0101 May 2023 2022 Annual Nonradiological Environmental Operating Report DCL-23-036, 2022 Annual Radiological Environmental Operating Report2023-05-0101 May 2023 2022 Annual Radiological Environmental Operating Report DCL-23-025, 2022 Annual Radioactive Effluent Release Report2023-05-0101 May 2023 2022 Annual Radioactive Effluent Release Report DCL-2023-512, Submittal of Receiving Water Monitoring Program 2022 Annual Report2023-04-27027 April 2023 Submittal of Receiving Water Monitoring Program 2022 Annual Report DCL-23-035, Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application2023-04-24024 April 2023 Response to Request for Additional Information on Request Regarding Senior Reactor Operator License Application DCL-23-028, Technical Specification Bases, Revision 142023-04-24024 April 2023 Technical Specification Bases, Revision 14 DCL-2023-511, Report on Discharge Monitoring for the First Quarter of 20232023-04-20020 April 2023 Report on Discharge Monitoring for the First Quarter of 2023 DCL-23-031, Submittal of Annual Report of Occupational Radiation Exposure for 20222023-04-19019 April 2023 Submittal of Annual Report of Occupational Radiation Exposure for 2022 DCL-23-024, Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System2023-04-0404 April 2023 Independent Spent Fuel Storage Installation, Response to Nrg Request for Additional Information for Revision to License Amendment Request 22-01 Request for Approval of Alternative Security Measures for Early Warning System DCL-23-022, 2023 Annual Statement of Insurance for Pacific Gas and Electric Companys Diablo Canyon Power Plant2023-03-29029 March 2023 2023 Annual Statement of Insurance for Pacific Gas and Electric Companys Diablo Canyon Power Plant DCL-23-023, Decommissioning Funding Report for Diablo Canyon Power Plant, Units 1 and 22023-03-28028 March 2023 Decommissioning Funding Report for Diablo Canyon Power Plant, Units 1 and 2 DCL-23-021, Independent Spent Fuel Storage Installation, Emergency Plan Update2023-03-23023 March 2023 Independent Spent Fuel Storage Installation, Emergency Plan Update DCL-23-020, Responses to NRC Questions Regarding License Renewal Efforts2023-03-17017 March 2023 Responses to NRC Questions Regarding License Renewal Efforts DCL-23-009, Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage2023-02-22022 February 2023 Owner'S Activity Report for Unit 2 Twenty-Third Refueling Outage DCL-23-008, Request Regarding Senior Reactor Operator License Application2023-02-22022 February 2023 Request Regarding Senior Reactor Operator License Application DCL-2023-503, Annual Sea Turtle Report2023-01-26026 January 2023 Annual Sea Turtle Report DCL-2023-502, (DCPP) 4th Quarter 2022 Report on Discharge Self-Monitoring2023-01-18018 January 2023 (DCPP) 4th Quarter 2022 Report on Discharge Self-Monitoring DCL-22-091, Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage2022-12-20020 December 2022 Revised Steam Generator Tube Inspection Report for Twenty-First Refueling Outage DCL-22-089, Core Operating Limits Report for Unit 2 Cycle 242022-12-20020 December 2022 Core Operating Limits Report for Unit 2 Cycle 24 DCL-22-085, Request to Resume Review of the License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application2022-10-31031 October 2022 Request to Resume Review of the License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application DCL-22-041, Site-Specific Decommissioning Cost Estimate, Revision 12022-10-12012 October 2022 Site-Specific Decommissioning Cost Estimate, Revision 1 DCL-22-042, Post-Shutdown Decommissioning Activities Report, Revision 12022-10-12012 October 2022 Post-Shutdown Decommissioning Activities Report, Revision 1 2024-01-29
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Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Vice President, Nuclear Services Mail Code 104/6 P. 0. Box 56 September 3, 2015 Avila Beach, CA 93424 805.545.4888 PG&E Letter DCL-15-106 Internal: 691.4888 Fax: 805.545 .6445 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant (DCPP) Unit 1 and Unit 2 ASME Section Xllnservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements
Dear Commissioners and Staff:
Pursuant to 10 CFR 50.55a(g)(5)(iii), Pacific Gas and Electric Company (PG&E) hereby requests NRC approval of lnservice Inspection Request for Relief NDE-FWNS-U1/U2 for the Diablo Canyon Power Plant Unit 1 and Unit 2 third lnservice Inspection Interval.
Relief is requested from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, for examination coverage of Class 2 feedwater nozzle-to-vessel welds. The details of the proposed request are enclosed.
PG&E requests approval of NDE-FWNS-U1/U2 by September 3, 2016.
PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.
This letter includes no revisions to existing regulatory commitments.
If you have any questions or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.
Sincerely, s1.7.n 5 ' 4a-rntt/4231/50033145 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator John P. Reynoso, NRC Acting Senior Resident Inspector Siva P. Lingam, NRR Project Manager Gonzalo L. Perez, Branch Chief, California Department of Public Health State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
Enclosure PG&E Letter DCL-15-1 06 10 CFR 50.55a Request NDE-FWNS-U1/U2 Relief Request in Accordance with 10 CFR 50.55a(g)(5)(iii)
--lnservice Inspection Impracticality--
- 1. ASME Code Component(s) Affected Diablo Canyon Power Plant (DCPP), Unit 1 and Unit 2, ASME Section XI , Code Class 2 steam generator (SG) feedwater nozzle-to-shell welds (two welds) are affected:
Line Code Item Outage Description Weld Number Size Cat. No. Examined (inch)
Unit 1 Feedwater Nozzle-to-C-B C2.21 FW-13.03.01 16 1R17 Shell Weld Unit 2 Feedwater Nozzle-to-C-B C2.21 FW-13.03.01 16 2R17 Shell Weld
- 2. Applicable Code Edition and Addenda The DCPP Unit 1 and Unit 2 third intervallnservice Inspection (lSI) Program Plan is based on the American Society of Mechanical Engineering (ASME) Boiler and Pressure Vessel Code,Section XI, 2001 Edition, with 2003 Addenda.
- 3. Applicable Code Requirement ASME Section XI, Table IWC-2500-1, Category C-B, Item No. C2.21 requires that SG feedwater nozzle-to-shell welds be volumetrically examined once during the lSI interval. Essentially 100 percent of the inner one-third volume of the weld, and adjacent base material, is to be examined in accordance with the requirements of Appendix I, 1-2120. The applicable examination volume is defined by Figure IWC-2500-4(a) and the examination is performed per the rules of ASME Section V, Article 4, as supplemented by Table 1-2000-1 .
- 4. Impracticality of Compliance The Unit 1 and Unit 2 SG feedwater nozzle-to-shell weld configurations are such that essentially 100 percent coverage of the ASME Code required examination 1
Enclosure PG&E Letter DCL-15-1 06 volume from the outside diameter (OD) is not feasible, as determined during the third interval examinations conducted in the Diablo Canyon Unit 1 and Unit 2 seventeenth refueling outages (1 R17 and 2R17).
Background Information The DCPP replacement SGs shell and nozzle forgings are fabricated from SA-508 Grade 3 Class 2 material with a nominal shell thickness of approximately 3.50 inches. The feedwater nozzles intersect the shell cylinder at a right angle and are joined by a weld extending concentrically around the nozzle forging and through the full thickness of the shell. The weld joint design is an unequal depth double U - groove design with an included groove angle of 7 degrees. The nozzle-to-shell welds were made using high strength filler metals (i.e., E9018M, ER80S-2, and S3NiMo1) whose composition and mechanical properties are similar to the joined base metals.
The subject welds were examined in 1R 17 and 2R 17 to the extent practicable using a combination of 35, 45, and 60 degree angled shear waves and zero degree longitudinal waves. The 35 and 45 degree angles were used for radial-out examinations in order to achieve the maximum possible coverage of the Code specified examination volume (Note: 60 degrees is not able to interrogate the Code examination volume in the radial-out direction due to the restricted setback due to the nozzle boss configuration). Forty-five and sixty degree angles were used for radial-in and circumferential scan examinations. No flaws were detected in any of the examinations of the subject welds.
The following table summarizes the exam volume coverage attained for both welds in each of the four scan directions and a combined average value.
Although the entire Code exam volume was interrogated by the zero degree longitudinal wave scan, coverage values are based on 35, 45, and 60 degree angles since they would be expected to detect service induced planar flaws emanating from the inside surface. The sketches that are presented at the end of this Enclosure illustrate the coverage for each of the inspection angles and directions used to determine coverage values.
Radial-In Radial-Out Circ. Scan Circ. Scan Steam Scan Scan Exam Volume Exam Volume Combined Generator Volume Volume CoverageCW Coverage CCW Coverage2 Coverage Coverage 1 Direction Direction 1-3 (Unit 1) 100%> 39%> 100%> 100°/o 84%>
2-1 (Unit 2) 100°/o 17%> 100%> 100°/o 79o/o 1
Combined coverage average for 35 and 45 degree angles.
2 The reported combined coverage value is an equal weighted average of the coverage values from each of the four scan directions.
2
Enclosure PG&E Letter DCL-15-1 06 Impracticality The compound curvature of the nozzle forging boss to shell contour transition zone and the forging design diameter constitute geometric restrictions that preclude full examination volume coverage from the outside surface. The coverage limitations are associated with the radial-out oriented scans.
An inherent design characteristic of the flange type nozzle configuration is that there is often insufficient setback distance for the radial-out scan beams to cover the entire Code specified exam area at the inside surface.
- 5. Burden Caused by Compliance "Essentially 100 percent" coverage of the exam volume from the outside surface would require redesign of the SG to move the weld farther back from the nozzle reinforcement or eliminate the weld by integrally incorporating the nozzle into the shell. Either of these two modifications would effectively result in performing major redesign and rework or replacement of the entire SG to accommodate full coverage of the exam are~ as specified by Code.
Performing examinations from the inside diameter of the SGs would require accessing the secondary side of the generators which involves substantial effort to remove the manway cover, making provisions for personnel access into a confined space, and work in a high risk foreign material exclusion area.
These efforts required to attain a small incremental increase in coverage would incur increased personnel radiation exposure and an increase in personnel safety risk due to work in a difficult to access and highly constrained work environment without a commensurate increase in examination effectiveness.
- 6. Proposed Alternative and Basis for Use PG&E proposes that the alternative ultrasonic examinations conducted to the maximum extent practicable from the outside surface provide reasonable assurance that the structural integrity of the subject welds remains intact.
The 1R 17 and 2R 17 examinations were implemented to the extent practicable using manual scan techniques and small footprint search units in effort to attain the greatest possible coverage of the required examination volume. The volume examined on both of the subject feedwater nozzle-to-shell welds includes the weld and surrounding base material near the inside surface of the weld joint, which are typically the highest stress regions and where degradation would likely manifest, should it occur.
The radial-in and circumferentially oriented angle beam scans fully interrogated the ASME XI Code exam volume, whereas, the radial-out scans covered a portion of the volume. Examination of ferritic materials from a single side of the weld has been demonstrated as effective in studies (Reference 1) and by 3
Enclosure PG&E Letter DCL-15-1 06 successful single side ferritic ASME XI, Appendix VIII qualifications per the Performance Demonstration Initiative program. Therefore, it is expected that the ultrasonic techniques employed on the DCPP feedwater nozzle-to-shell welds would have detected structurally significant flaws if extant within the examination area.
The 1R 17 and 2R 17 ultrasonic examinations with combined coverage values of approximately 84 percent and 79 percent for the selected Unit 1 and Unit 2 subject welds, respectively, provide reasonable assurance that the structural integrity of these welds remains intact and provide an acceptable level of quality and safety.
Potential Failure Consequences A failure of the feedwater nozzle-to-shell weld could result in a loss of feedwater to a SG. Depending on the size of the postulated break (leak) the specific consequences will vary. At the smallest end of the break size spectrum, the feedwater system would be capable of maintaining SG level through normal makeup. Larger break sizes would result in depressurization of the SG and loss of heat transfer capability. The worst case consequence would occur if the nozzle-to-shell weld was to suffer 360 degree circumferential cracking. In this case, the break is bounded by the feedwater line break assumed in the DCPP design basis analysis.
Essentially no change to overall plant safety is expected due to implementation of the proposed alternative in lieu of the Code requirement. This assumption is based on the effectiveness of ultrasonic examination on ferritic material as previously described, and little or no historical occurrence of large service induced planar flaws in this type of weldment.
- 7. Duration of Proposed Alternative Relief is requested for the remainder of the DCPP Unit 1 and Unit 2 third lSI intervals. The DCPP Unit 1 third inspection interval nominally ended on May 6, 2015. The DCPP Unit 2 third inspection interval is nominally scheduled to end on March 12, 2016. Actual end dates of the interval are dependent on the completion dates of the 19th refueling outages for each unit, in accord with ASME Section XI, paragraph IWA-2430(d)(1).
As stated in the relief request, the third interval for Unit 1 nominally ended May 8, 2015, the 30th anniversary of the commercial operation date for the unit. However, per Section XI paragraph IWA-2430(d)(1 ), "Each inspection interval may be reduced or extended by as much as one year." Paragraph IWA-2430(d)(3) states, "That portion of an inspection interval described as an inspection period may be reduced or extended by as much as one year to enable an inspection to coincide with a plant outage." For Unit 1, the interval is being extended past the nominal end date to November 6, 2015, to coincide with the dates of 1R 19, and may be further continued if necessary until May 8, 2016.
4
Enclosure PG&E Letter DCL-15-1 06 Therefore, this submittal is timely. All subject examinations or their approved alternatives would be credited for the third inspection interval only.
For alternative requests based on impracticality, submittals must be made no later than 12 months after the expiration of the interval for which relief is sought per 10 CFR 50.55a(g)(5)(iii)/(iv).
- 8. Precedents This request is essentially the same as DCPP request NDE-25.2R8 approved in an NRC letter dated September 29, 1999, for the second lSI interval. It is also similar to Relief ISI-6 for Calvert Cliffs Nuclear Power Plant (TAC numbers MA-9404 and MA-9405). ISI-6 was approved by the NRC in a letter dated July 27, 2001.
- 9. References
- 1. P.G. Heasler and S.R. Doctor, 1996. Piping Inspection Round Robin, NUREG/CR-5068, PNNL-1 0475, U.S. Nuclear Regulatory Commission, Washington, DC 5
Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration RSG 1-3 & 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 45° and 60° Radial-In Scans Nozzle Forging 100% Coverage 45° and 60° Radial-In Scans
~ = Volume Examined 6
Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 1-3 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 35° and 45° Radial-Out Scans 35° and 45° (35° Position Illustrated)
Nozzle Forging 67°/o Coverage 35° Radial-Out Scan 11 °/o Coverage 45° Radial-Out Scan
~ = Volume Examined With 35°
- = Volume Examined With 35° & 45° 7
Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 35° and 45° Radial-Out Scans 35° and 45° (35° Position Illustrated)
Nozzle Forging 34°/o Coverage 35° Radial-Out Scan 0°/o Coverage 45° Radial-Out Scan
~ = Volume Examined With 35° 8
Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 1-3 & 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 45° and 60° Clockwise and Counter-Clockwise Scans Nozzle Forging 100% Coverage 45° and 60° Clockwise and Counter-Clockwise Scans
~=Volume Examined 9