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MONTHYEARDCL-13-038, ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria2013-04-11011 April 2013 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria Project stage: Request ML13123A0412013-05-0202 May 2013 Acceptance Review Email, Relief Request NDE-RCS-SE-1R18 - Alternative to Requirements to ASME Code, Section XI, Appendix Viii, Suppl. 2 and 10 as Modified by CC N-695 and CC N-696 to Allow Use of Alternative Depth-Sizing Criteria Project stage: Acceptance Review ML13191A0832013-07-0909 July 2013 Request for Additional Information Email, Relief Request NDE-RCS-SE-1R18, Alternative to ASME Code, Section XI, App. Viii, Suppl. 2 and 10 as Modified by CC N-695 and CC N-696, to Allow Use of Alternative Depth-Sizing Criteria Project stage: RAI DCL-13-078, Response to NRC Request for Additional Information Regarding Request for Relief NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria2013-08-0606 August 2013 Response to NRC Request for Additional Information Regarding Request for Relief NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria Project stage: Response to RAI ML13350A1512014-01-0303 January 2014 Relief Request NDE-RCS-SE-1R18, Alternative to ASME Code, Section XI, App. Viii, Suppl. 2 and 10 as Modified by CC N-695 and CC N-696, to Allow Use of Alternative Depth-Sizing Criteria Project stage: Other 1CAN041603, Request for Relief from ASME Section XI, Table IWB-2500-1 Requirements2016-04-22022 April 2016 Request for Relief from ASME Section XI, Table IWB-2500-1 Requirements Project stage: Request DCL-16-102, ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternative Depth Sizing Criteria2016-10-10010 October 2016 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternative Depth Sizing Criteria Project stage: Request 2013-08-06
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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
[Table view] Category:Code Relief or Alternative
MONTHYEARML20249C8502020-10-22022 October 2020 Proposed Alternative NDE-RCS-SE-2R22 to the Requirements of the ASME Code ML19176A0082019-06-28028 June 2019 Request for Relief from the Inservice Inspection Program Requirements of ASME Code, Section XI, for Steam Generator and Pressurizer Nozzle Inside Radius Sections ML19023A0262019-01-25025 January 2019 Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4 ML17263A0592017-10-0606 October 2017 Diablo Canyon Power Plant, Unit 2 - Relief Requests to Allow Use of Proposed Alternatives to Implement Requirements of ASME Code, for Reactor Pressure Vessel Welds for the Third ISI Interval (CAC Nos. MF9386, MF9387, MF9388, MF9389, and MF9390) ML17138B1152017-05-18018 May 2017 10 CFR 50.55a Request for Approval of Alternative - RI-ISI-2 ML17037D2672017-03-0909 March 2017 Relief Request NDE-RCS-SE-1R20 to Allow Use of Alternate Depth Sizing Criteria for the Fourth Inservice Inspection ML16207A3972016-07-27027 July 2016 Relief Request NDE-SIF-U1, Allow Use of Alternate Examination Volume Coverage Requirements, Class 2 Seal Injection Welds, for the Third 10-Year Inservice Inspection Interval ML16056A5132016-03-0707 March 2016 Relief Request FLIG-U1, Request to Extend Third 10-Year Inservice Inspection Interval for Reactor Vessel Stud Hole Ligament Examinations ML16055A2812016-02-26026 February 2016 Relief Request NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements, Third 10-Year Inservice Inspection Interval ML16013A3402016-01-27027 January 2016 Relief Request NDE-PNS-U2A, ASME Code Section XI Requirements to Allow Use of Alternate Exam Volume Coverage Requirements, for the Third 10-year Inservice Inspection Interval ML15337A0212015-12-0808 December 2015 Relief Request NDE-SG-MS-IR - Relief from Exam Requirements for Main Steam Nozzles in Replacement Steam Generators, Third 10-Year Inservice Inspection Interval ML15299A0342015-11-0404 November 2015 Relief Request NDE-RCS-SE-2R19, Use of Alternate Sizing Qualification Criteria Through a Protective Clad Layer, Third 10-Year Inservice Inspection Interval DCL-15-118, Response to NRC Request for Additional Information Regarding Relief Request FLIG-U12015-10-12012 October 2015 Response to NRC Request for Additional Information Regarding Relief Request FLIG-U1 ML15261A0072015-09-18018 September 2015 Request for Relief PRS-3, Proposed Alternative to Requirements for System Leakage Testing Class 1 Piping Vent, Drain, and Test Isolation Devices, for the Third 10-year Inservice Inspection Interval (TAC MF6520-MF6521) DCL-15-086, ASME Section XI Inservice Inspection Program Request for Relief PRS-32015-07-28028 July 2015 ASME Section XI Inservice Inspection Program Request for Relief PRS-3 ML15187A0352015-07-15015 July 2015 Relief Request REP-SI, Revision 2, Proposed Alternative for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments, Remaining Useful Life +20 Years ML15168A0242015-06-19019 June 2015 Request for Alternative RPV-U1 - Extension to Allow Use of Alternate Reactor Inspection Interval Requirements for the Third 10-Year Inservice Inspection Interval DCL-15-075, Relief Request REP-SI - Supplement2015-06-18018 June 2015 Relief Request REP-SI - Supplement ML14251A4682014-09-16016 September 2014 Relief Request NDE-SFW-13, ASME Code Section XI, Request for Alternative to Allow Use of Alternate Examination Requirements, Third 10-Year Inservice Inspection Interval ML14196A0952014-08-0404 August 2014 Request for Relief from ASME Code Requirements for Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary ML14104B6132014-05-14014 May 2014 Relief Request NDE-PNS-U2, Requirements for Examination Coverage of ASME Class 1 Pressurizer Nozzle-to-Head Welds, Third 10-Year Inservice Inspection Interval ML13350A1512014-01-0303 January 2014 Relief Request NDE-RCS-SE-1R18, Alternative to ASME Code, Section XI, App. Viii, Suppl. 2 and 10 as Modified by CC N-695 and CC N-696, to Allow Use of Alternative Depth-Sizing Criteria ML13192A3542013-09-12012 September 2013 Relief Request RVFLNG-INT3 - Alternative to ASME Code Pressure Test Requirements for Class 1 Reactor Vessel Flange Leakoff Lines, Third 10-Year Inservice Inspection Interval (TAC MF0408-MF0409) ML13232A3082013-08-28028 August 2013 Relief Request REP-1 U2, Revision 2, Alternative to Code Requirements for Pressurizer Structural Weld Overlays, for the Third 10-Year Inservice Inspection Interval DCL-13-038, ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria2013-04-11011 April 2013 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1R18 to Allow Use of Alternative Depth Sizing Criteria ML1106003562011-03-29029 March 2011 Relief Request NDE-RCS-SE-2R16, Alternative to Requirements of ASME Code, Section XI, Supplement 10 as Modified by Code Case N-695, Qualification Requirements for Dissimilar Metal Piping Welds ML0801100012008-02-0606 February 2008 Approval of Relief Request REP-1 U2, Revision 1, for the Application of Weld Overlay on Dissimilar Metal Welds of Pressurizer Nozzles ML0716200722007-06-25025 June 2007 Relief Requests NDE-SLH U2 and NDE-LSL U2 for Second 10-year Inservice Inspection Interval Re Reactor Vessel Shell Weld Inspections ML0707803692007-03-29029 March 2007 Relief Request NDE-SBR for the Second 10-Year Interval Inservice Inspection and Examination Program for Snubbers DCL-07-032, Response to NRC Request for Additional Information Regarding ASME Section XI Inservice Inspection Program Relief Requests NDE-SLH U2 and NDE-LSL U22007-03-23023 March 2007 Response to NRC Request for Additional Information Regarding ASME Section XI Inservice Inspection Program Relief Requests NDE-SLH U2 and NDE-LSL U2 ML0621200382006-09-20020 September 2006 Relief, Relief Request for Relief NDE-SLH (TAC No. MC9474) - Reactor Pressure Vessels Welds DCL-06-069, Residual Heat Removal Weld RB-119-11 - Flaw Analytical Evaluation Results2006-06-0606 June 2006 Residual Heat Removal Weld RB-119-11 - Flaw Analytical Evaluation Results ML0608203592006-04-0404 April 2006 Relief, Relief Request for Second Inservice Inspection Interval for Steam Generator Welds, MC9473 ML0603302512006-01-30030 January 2006 Relief Requests 1, 2, and 3 Approval ML0526603312005-10-26026 October 2005 Diablo Canon Pp, Units 1 & 2 - Approval of Requests for Relief from the (ASME) Boiler and Pressure Vessel Code (Code) Section XI Inservice Inspection Program Requirements (Tac Nos MC6693 & MC6694) ML0208502842002-04-11011 April 2002 Relief, Approval of Inservice Inspection Relief Request - Use of ASME Code Case N-597 (TAC MB1693 and MB1694) 2020-10-22
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Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. O. Box 56 Avila Beach, CA 93424 805.545.4888 April 11 ,2013 Internal: 691 .4888 Fax: 805.545.6445 PG&E Letter DCL-13-038 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Docu ment Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Diablo Canyon Power Plant (DCPP) Unit 1 ASME Section XI Inservice Inspection Program Relief Request NDE-RCS-SE-1 R18 to Allow Use of Alternative Depth Sizing Criteria
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 Inservice Inspection (lSI) Relief Request (RR)
NDE-RCS-SE-1 R18 for the Diablo Canyon Power Plant Unit 1 third lSI interval. The details of the request are enclosed.
This request is associated with the use of alternatives to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Appendix VIII, Supplements 2 and 10 as modified by Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,"
and Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted From the Inside Surface."
To support the examinations to be performed in the upcoming Unit 1 eighteenth refueling outage scheduled for February 2014, PG&E requests approval of this RR by January 15, 2014.
This communication contains new commitments to be implemented following NRC approval of this RR. The commitments are contained in Attachment 1 of the Enclosure.
If you have any questions regarding the information enclosed, or other lSI program activities, please contact Mr. Tom Baldwin at (805) 545-4720.
Sincerely, A member of the STARS (StrategiC Teaming and Resource Sharing) Alliance Callaway
Document Control Desk PG&E Letter DCL-13-038 April 11 ,2013 Page 2 rnttJ4231/SAPN 50033145-62 Enclosure cc: Diablo Distribution celene!: Arthur T. Howell, III, NRC Region IV Thomas R. Hipschman, NRC Senior Resident Inspector James T. Polickoski, NRR Project Manager State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
Enclosure PG&E Letter DCL-13-038 10 CFR SO.SSa Relief Request NDE ..RCS ..SE ..1R18 Relief Request in Accordance with 10 CFR SO.SSa(g)(S)(iii)
--Inservice Inspection Impracticality--
- 1. ASME Code Components Affected The American Society of Mechanical Engineers (ASME)Section XI Class 1 Diablo Canyon Power Plant (DCPP) Unit 1 Reactor vessel nozzle to safe-end and safe-end to piping welds are listed below. See Figure 1 for safe-end general configuration and materials.
Code Cat.! Description Weld Number
Item No.
- Nominal Wall N-770-1, A-2 Loop 1 outlet nozzle to safe-end WIB-RC-1-1 (SE) 29 in.!2.5 in.
R-A, R1.20 Loop 1 outlet safe-end to pipe WIB-RC-1-1 29 in'!2.5 in.
R-A, R1.20 Loop 1 inlet elbow to safe-end WIB-RC-1-18 27.5 in/2.38 in.
N-770-1, B Loop 1 inlet safe-end to nozzle WIB-RC-1-18(SE) 27.5 in/2.38 in.
N-770-1, A-2 29 in.!2.5 in.
Loop 2 outlet nozzle to safe-end WIB-RC-2-1 (SE)
R-A, R1.20 29 in'!2.5 in.
Loop 2 outlet safe-end to pipe WIB-RC-2-1 R-A, R1.20 27.5 in'!2.38 in.
Loop 2 inlet elbow to safe-end WIB-RC-2-20 N-770-1, B 27.5 in'!2.38 in.
Loop 2 inlet safe-end to nozzle WIB-RC-2-20(SE)
N-770-1, A-2 29 in'!2.5 in.
Loop 3 outlet nozzle to safe-end WIB-RC-3-1 (SE)
R-A, R1.20 29 in.!2.5 in.
Loop 3 outlet safe-end to pipe WIB-RC-3-1 R-A, R1.20 27.5 in'!2.38 in.
Loop 3 inlet elbow to safe-end WIB-RC-3-18 N-770-1, B 27.5 in'!2.38 in.
Loop 3 inlet safe-end to nozzle WIB-RC-3-18(SE)
N-770-1 , A-2 29 in'!2.5 in.
Loop 4 outlet nozzle to safe-end WIB-RC-4-1 (SE)
R-A, R1.20 29 in.!2.5 in.
Loop 4 outlet safe-end to pipe WIB-RC-4-1 R-A, R1.20 27.5 in'!2.38 in.
Loop 4 inlet elbow to safe-end WIB-RC-4-18 N-770-1, B 27.5 in'!2.38 in.
Loop 4 inlet safe-end to nozzle WIB-RC-4.;18(SE)
1
Enclosure PG&E Letter DCL-13-038
2. Applicable Code Edition and Addenda
The DCPP third Inservice Inspection Interval Program Plan (lSIPP) is based on the ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition with Addenda through 2003. Appendix VIII ultrasonic piping examinations are performed in accordance with the Section XI, 2001 Edition without Addenda.
The ISIPP is augmented with the requirements of ASME Code Case N-770-1 as modified by 10 CFR 50.55a for the examination of dissimilar metal vessel nozzle butt welds containing Alloy 82/182 material. These requirements apply to the dissimilar metal welds connecting the reactor nozzles to the reactor coolant system (RCS) piping.
The DCPP lSI PP references Code Cases N-695 and N-696 which are unconditionally approved for use in NRC Regulatory Guide 1.147, Revision 16.
The DCPP Unit 1 third inspection interval commenced on January 1, 2006, and is nominally scheduled to end on May 6, 2015.
3. Applicable Code Requirement
ASME Code Case N-770-1 stipulates ultrasonic examination of dissimilar metal welds fabricated from Alloy 82/182 material. These requirements apply to the DCPP reactor nozzle to safe-end welds.
DCPP Risk Informed ISIPP examination Category R-A, Item R.120 (formerly Code Category B-F, B5.10 in the 2001 Edition through 2003 Addenda) specifies volumetric examination for the associated RCS safe-end to piping welds that are fabricated with 308 stainless steel weld material.
All of the specified ultrasonic examinations are to be conducted per Appendix VIII Supplements 2 (wrought austenitic welds) and 10 (dissimilar metal welds) with the exception of the cast side of the cold leg elbow to safe-end welds. The cast elbow sides of these welds are to be examined per Appendix III as modified by Supplement 1.
ASME Code Cases N-695 and N-696 provide alternatives to Appendix VIII Supplements 10 and 2 and include requirements for depth sizing accuracy; Code Case N-695 3.3( c) states, " ... are qualified for depth sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)," while Code Case N-696 3.3(d) states, " ... qualified for depth sizing when the flaw depths estimated by ultrasonics, as compared with the true depths, do not exceed 0.125 in. (3 mm) RMS, when they are combined with a successful Supplement 10 qualification."
2
Enclosure PG&E Letter OCL-13-038
- 4. Impracticality of Compliance Relief is requested from the 0.125 in. root mean square error (RMSE) depth sizing accuracy requirement of Code Cases N-695 and N-696.
To date, although examination vendors have qualified for detection and length sizing in accordance with the Appendix VIII requirements for examinations from the inside diameter (10), the vendors have not met the established RMSE of 0.125 in. for depth sizing despite numerous attempts.
Consequently, relief from the Code Case specified 0.125 in. RMSE depth sizing accuracy requirement is necessary to support evaluation of findings from examinations when conducted from the inside surface.
Additionally, relief is requested from implementing the requirements of ASME Appendix III as modified by Supplement 1 for the examination of the cast stainless steel side of the cold leg elbow to safe-end welds. Specific examination requirements for cast stainless steel materials are not yet addressed in Section XI. Examinations using the Appendix III, Supplement 1 process rather than the Appendix VIII Supplement 2 requirements used on the wrought side of the weld would result in additional burden without benefit. The Appendix III examinations would require separate procedures, calibration blocks, calibrations and possibly completely separate examinations.
- 5. Proposed Alternative and Basis for Use PG&E proposes to use a vendor qualified for 10 detection and length sizing per Appendix VIII as applicable to the welds included in this request. Indications requiring depth sizing will be treated as detailed in the following paragraphs.
Since PG&E is in the vendor selection process at the time of this submittal, references to vendor depth sizing RMSE values are generic.
The Appendix VIII examination process will also be used to examine the cast stainless steel side of the cold leg elbow to safe-end welds in lieu of Appendix III and Supplement 1 requirements. All 10 examinations will be augmented with inside surface profilometry and eddy current examination.
If a reportable flaw is detected and determined to be 10 surface connected during examination of the welds that are included in Relief Request NOE-RCS-SE-1 R18, PG&E will provide a flaw evaluation including the measured flaw size as determined by ultrasonic testing (UT) for review by NRC. Eddy current testing will be used to determine if flaws are surface connected.
Additional data including details of the surrounding 10 surface contour in the region of the flaw and percentage of the exam area where UT probe lift-off is evident, if any, will be included.
3
Enclosure PG&E Letter DCL-13-038 In the event that any flaw(s) requiring depth sizing is detected during examination of the welds that are included in Relief Request NDE-RCS-SE-1 R18, the following criteria shall be implemented:
- ID connected flaws detected and measured as less than 50 percent through-wall in depth shall be adjusted by adding a correction factor to the flaw depth such that the adjusted flaw depth is equal to the measured flaw depth plus (contractor RMSE minus 0.125 in.), prior to comparison to the applicable acceptance criteria;
- If ID connected flaw(s) are detected and measured as 50 percent through-wall depth or greater and to remain in service without mitigation or repair, PG&E shall submit flaw evaluation(s) for review and approval by NRC prior to reactor startup. The flaw evaluation shall include:
o Information concerning the mechanism that caused the flaw o Information concerning the inside surface roughness/profile of the region surrounding the flaw o Information concerning areas where UT probe lift-off is observed All welds included in this request have been previously examined from the ID with an Appendix VIII qualified detection process and length sizing in the thirteenth refueling outage. The ultrasonic examinations were supplemented by surface profilometry and eddy current testing. Greater than 90 percent coverage of the required exam areas was achieved in all cases. This inspection history confirms that the inside surface profiles of the welds included in this request are suitable for ultrasonic examination from the ID in accordance with the referenced requirements as modified by the proposed alternative sizing requirements.
In the event the proposed alternate examination process grossly mischaracterizes a significant planar flaw, the potential resulting failure of one of the RCS loop outlet/inlet welds could result in a loss of coolant accident (LOCA).
Depending on the size of the postulated break, the specific consequences will vary. At the smallest end of the break size spectrum, the charging system would be capable of maintaining RCS pressure through normal makeup. Larger break sizes would result in depressurization of the RCS, reactor trip and a safety injection. The worst case consequence would occur if one of the nozzle to pipe welds were to suffer 360 degree circumferential cracking. In this case, the break size is bounded by the line ID which is less than the break size used in the large break LOCA design basis analysis.
NRC review of vendor performance demonstration initiative qualification test results concluded that for flaw depths less than 50 percent wall thickness, reasonable assurance that a flaw will be appropriately depth sized may be 4
Enclosure PG&E Letter DCL-13-038 obtained by adding a correction factor to the flaw depth such that the adjusted flaw depth is equal to the measured flaw depth+(contractor RMSE-0.125 in.).
Therefore, the proposed alternative for correcting measured flaw depths for flaws less than 50 percent wall thickness and performing and submitting a flaw specific analysis for flaws equal to or greater than 50 percent wall thickness will have little effect on the likelihood of gross flaw mischaracterization and any potential for resulting component failure.
For the cast stainless steel side of cold leg elbow to safe-end welds, the proposed Appendix VIII 10 exam supplemented with surface profilometry and eddy current is better suited for detection of 10 connected flaws than the Code specified Appendix III, Supplement 1 process.
The DCPP Unit 1 reactor coolant nozzle to safe-end and safe-end to piping welds will be examined from the 10 using personnel, procedures and equipment qualified by demonstration in all aspects except depth sizing. The proposed method of addressing depth sizing RMSE values will compensate for the potential through-wall sizing variation and provides reasonable assurance of structural integrity in accordance with 10 CFR 50.55a(g)(5)(iii).
- 6. Duration of Proposed Alternative The duration of the proposed alternative is for the remainder of the DCPP Unit 1 third lSI interval, nominally scheduled to end on May 6, 2015.
- 7. Precedent The proposed alternative method of addressing the difference between the demonstrated and required RMSE values was approved for McGuire Nuclear Station Unit 2 by NRC letter dated September 24, 2012 (ADAMS Accession No. ML12258A363).
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Enclosure PG&E Letter DCL-13-038 Weld (Inconel182)
L Nozzle-to-Safe-End Safe-End-to-Piping Weld (308 SS)
__ l Cold Leg Elbow (A-3S1 CF8M)
/
Hot Leg SS Pipe (A-376 Type 316)
SS Clad DCPP Unit 1 Sketch of Hot Leg and Cold Leg Safe-End General Configuration Figure 1 6
Enclosure Attachment 1 PG&E Letter OCL-13-038 List of Regulatory Commitments Commitment 1 If a reportable flaw is detected and determined to be 10 surface connected during examination of the welds that are included in Relief Request NOE-RCS-SE-1 R18, PG&E will provide a flaw evaluation including the measured flaw size as determined by UT for review by NRC. Eddy current testing will be used to determine if flaws are surface connected. Additional data including details of the surrounding 10 surface contour in the region of the flaw and percentage of the exam area where UT probe lift-off is evident, if any, will be included.
Commitment 2 In the event that any flaw(s) requiring depth sizing is detected during examination of the welds that are included in Relief Request NOE-RCS-SE-1 R18, the following criteria shall be implemented:
- 10 connected flaws detected and measured as less than 50 percent through-wall in depth shall be adjusted by adding a correction factor to the flaw depth such that the adjusted flaw depth is equal to the measured flaw depth plus (contractor RMSE minus 0.125 in.), prior to comparison to the applicable acceptance criteria;
- If 10 connected flaw(s) are detected and measured as 50 percent through-wall depth or greater and to remain in service without mitigation or repair, PG&E shall submit flaw evaluation(s) for review and approval by NRC prior to reactor startup.
The flaw evaluation shall include:
o Information concerning the mechanism that caused the flaw o Information concerning the inside surface roughness/profile of the region surrounding the flaw o Information concerning areas where UT probe lift-off is observed 1