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| | number = ML14351A047 | | | number = ML14351A047 |
| | issue date = 12/23/2014 | | | issue date = 12/23/2014 |
| | title = Diablo Canyon, Units 1 and 2, Request for Additional Information, Relief Request REP-SI, Alternative for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments, Remaining Useful Life +20 Years (TAC Nos. MF4476 an | | | title = Request for Additional Information, Relief Request REP-SI, Alternative for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments, Remaining Useful Life +20 Years |
| | author name = Lingam S P | | | author name = Lingam S |
| | author affiliation = NRC/NRR/DORL/LPLIV-1 | | | author affiliation = NRC/NRR/DORL/LPLIV-1 |
| | addressee name = Halpin E D | | | addressee name = Halpin E |
| | addressee affiliation = Pacific Gas & Electric Co | | | addressee affiliation = Pacific Gas & Electric Co |
| | docket = 05000275, 05000323 | | | docket = 05000275, 05000323 |
| | license number = DPR-080, DPR-082 | | | license number = DPR-080, DPR-082 |
| | contact person = Lingam S P | | | contact person = Lingam S |
| | case reference number = TAC MF4476, TAC MF4477 | | | case reference number = TAC MF4476, TAC MF4477 |
| | document type = Letter, Request for Additional Information (RAI) | | | document type = Letter, Request for Additional Information (RAI) |
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| =Text= | | =Text= |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2014 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424 SUBJECT: DIABLO CANYON POWER PLANT, UNITS 1 AND 2 -INSERVICE INSPECTION PROGRAM REQUEST FOR ALTERNATIVE REP-SI: PROPOSED ALTERNATIVE TO REQUIREMENTS FOR REPAIR AND REPLACEMENT ACTIVITIES FOR CERTAIN SAFETY INJECTION PUMP WELDED ATTACHMENTS (TAG NOS. MF4476 AND MF4477) Dear Mr. Halpin: By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML 14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWA-4000, "Repair/Replacement Activities," including IWA-4130, "Alternative Requirements," and IWA-4131, "Small Items," as corrective action for the four affected Code Class 2, NPS [nominal pipe size] 3/4-inch socket welds on each safety injection (SI) pump at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection pumps. Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i), the licensee requested NRC approval to use an alternative from the requirements of the ASME Code, Section XI, for Repair/Replacement rules governing socket welded attachments to safety injection pumps. The NRC staff has reviewed the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the following enclosed request for additional information (RAI). A draft form of these RAis was provided to DCPP licensee staff. During a recent licensing status call on December 9, 2014, between the NRC project manager for DCPP and DCPP licensing staff, it was agreed that no additional licensee clarifications (telecoms) were necessary. Please respond to the enclosed final RAis no later than 45 days from the date of this letter. | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2014 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424 |
| E. Halpin -2-If you have any questions, please contact me at 301-415-1564 or via e-mail at Siva.Lingam@nrc.gov. Docket Nos. 50-275 and 50-323 Enclosure: Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION REQUEST FOR ALTERNATIVE REP-SI ACCEPTANCE OF VENT AND DRAIN SOCKET WELDS AT SAFETY INJECTION PUMPS PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML 14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWA-4000, at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection (SI) pumps. In order to to complete its review, the NRC staff requests the following additional information. RAI-EVIB 1.0 The relief request is based on the premise that existing Sl pump vent and drain socket welds may be determined acceptable as-is for continued service. However, the possibility exists that all vent and drain socket welds could fail on an Sl pump. To properly evaluate the relief request, the NRC staff requests that the licensee describe: (a) The safety significance if all vent and drain weld connections are severed on a single Sl pump. (b) In case of failure of all vent and drain weld connections on a single Sl pump, please explain if the safety function of nearby equipment and personnel will be affected. RAI-EVIB 2.0 The relief request partly bases the acceptability of existing Sl pump vent and drain socket welds on the performance of a welding procedure qualification test with representative type 410 stainless steel (P-6) and type 304 stainless steel (P-8) base materials using type 309 filler metal per the production welding procedure parameters without post-weld heat treatment, and that the qualification test assembly passed the required ASME Code destructive tests. However, for this Enclosure | | |
| -2-material combination, the possibility exists that untempered martensite may be formed in the P-6 heat affected zone of the qualification weldment. The NRC staff requests that the licensee discuss the following: (a) Whether hardness testing was performed in the P-6 heat affected zone of the qualification weldment to objectively measure the possibility and extent of untempered martensite formation. (b) Since the destructive test specimens were machined down from 0.375-inch in thickness to 0.300-inch in thickness, please justify that the machined destructive test specimens are representative of the qualification weldment in terms of microstructure. Discuss whether the possibility exists for detrimental untempered martensite to be formed during welding but removed during the preparation of the destructive test specimens. RAI-EPNB 1.0 The licensee submitted the relief request under paragraph 50.55a(a)(3)(i) of Title 1 0 of the Code of Federal Regulations (1 0 CFR). The NRC staff believes that the use of the proposed alternative (i.e., continued operation with the existing non-conforming welds) provides a lower level of quality and safety than with ASME Code compliant welds. The NRC staff suggests that it is more appropriate to request the relief pursuant to 10 CFR 50.55a(a)(3)(ii). Please propose the alternative to the ASME Code under 10 CFR 50.55a(a)(3)(ii) or justify why it is appropriate to submit this relief request under 10 CFR 50.55a(a)(3)(i). The licensee should also provide hardship justification for not complying with the ASME Code requirements. RAI-EPNB 2.0 The industry operating experience has shown that socket welds are susceptible to through-wall cracking. (a) Please discuss whether any leakage detection systems are available to detect any potential leak from the subject socket welds and whether the operators in the control room would be notified of the leakage. (b) Please discuss the consequence of a through-wall leak and a complete severance at any of the subject socket welds. RAI-EPNB 3.0 When the safety injection pumps are running; Please discuss whether vibration exists on the drain lines and vent lines. If yes, discuss the potential for the vibration that may cause cracking at the socket welds.
| | ==SUBJECT:== |
| -3-Structural Integrity Associates Stress Calculation RAI-EPNB 4.0 Section 2.4, page 2-3, of the stress calculation states that unit axial load of 1000 pounds (lbs). was used as an input. Please discuss where and how 1 000 lbs. was obtained and derived. RAI-EPNB 5.0 Section 3 of the stress calculation states that when analyzing the outside diameter flaw, the methods of the ASME Code, Section XI, Appendix C, C-7300 and American Petroleum Institute, API-579, were used to obtain stress intensity factors. However, when analyzing the inside diameter flaw, it does not appear that the API-579 method was used. Please clarify whether API-579 method was used to evaluate the inside diameter flaw. If not, discuss the reference of the flaw evaluation method. RAI-EPNB 6.0 Section 3.3.2, page 3-8, of the stress calculation states that " ... residual stresses would not contribute to fatigue crack growth ... " The NRC staff believes that although residual stresses are steady state in nature (i.e., constant), they affect the maximum tensile stress and R [stress ratio] and may, therefore, affect fatigue crack growth. Please clarify the above quoted statement. Discuss whether the above statement is specifically applied to the inside diameter flaw in the subject component. RAI-EPNB 7.0 Section 4 of the stress calculation discusses fatigue crack growth. Section 4 also discussed a postulated inside diameter initiated flaw. If an inside diameter flaw is postulated, please discuss the likelihood of stress-corrosion cracking in the subject welds. RAI-EPNB 8.0 Page 5-1 of the stress calculation states that the allowable flaw depth for an inside diameter flaw exceeds 80 percent of the wall thickness. IWB-3643 of the ASME Code, Section XI, 2003 addenda (Code of record) limits the maximum allowable flaw depth to 75 percent through-wall. Please explain why the allowable flaw depth for an inside diameter flaw exceeds (i.e., non-conservative) the ASME Code, Section XI limitation.
| | DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - INSERVICE INSPECTION PROGRAM REQUEST FOR ALTERNATIVE REP-SI: |
| E. Halpin -2 -If you have any questions, please contact me at 301-415-1564 or via e-mail at Siva. Lingam@nrc.gov. Docket Nos. 50-275 and 50-323 Enclosure: Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION: PUBLIC LPL4-1 Reading RidsAcrsAcnw_MaiiCTR Resource RidsNrrDeEvib Resource RidsNrrDeEpnb Resource ADAMS Accession No. ML 14351A047 Sincerely, /RAJ Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDorllpl4-1 Resource RidsNrrLAJBurkhardt Resource RidsNrrPMDiabloCanyon Resource RidsRgn4MaiiCenter Resource *email dated OFFICE NRR/DORL/LPL4-1/PM NRR/DORL/LPL4-1/PM NRR/DORLILPL4-1/LA NRR/DE/EPNB NAME RHaskell SLingam JBurkhardt DAiley* DATE 12/12/2014 12/23/14 12/18/14 10/8/14 OFFICE NRR/DE/EVIB/BC NRR/DORLILPL4-1/BC(A) NRR/DORL/LPL4-1/PM NAME SRosenberg* EOesterle SLingam DATE 11/12/14 12/23/14 12/23/14 OFFICIAL RECORD COPY
| | PROPOSED ALTERNATIVE TO REQUIREMENTS FOR REPAIR AND REPLACEMENT ACTIVITIES FOR CERTAIN SAFETY INJECTION PUMP WELDED ATTACHMENTS (TAG NOS. MF4476 AND MF4477) |
| }} | | |
| | ==Dear Mr. Halpin:== |
| | |
| | By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWA-4000, "Repair/Replacement Activities," |
| | including IWA-4130, "Alternative Requirements," and IWA-4131, "Small Items," as corrective action for the four affected Code Class 2, NPS [nominal pipe size] 3/4-inch socket welds on each safety injection (SI) pump at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection pumps. |
| | Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i), the licensee requested NRC approval to use an alternative from the requirements of the ASME Code, Section XI, for Repair/Replacement rules governing socket welded attachments to safety injection pumps. |
| | The NRC staff has reviewed the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the following enclosed request for additional information (RAI). A draft form of these RAis was provided to DCPP licensee staff. During a recent licensing status call on December 9, 2014, between the NRC project manager for DCPP and DCPP licensing staff, it was agreed that no additional licensee clarifications (telecoms) were necessary. Please respond to the enclosed final RAis no later than 45 days from the date of this letter. |
| | |
| | E. Halpin If you have any questions, please contact me at 301-415-1564 or via e-mail at Siva.Lingam@nrc.gov. |
| | Sincerely, |
| | ~~~*~ |
| | Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323 |
| | |
| | ==Enclosure:== |
| | |
| | Request for Additional Information cc w/encl: Distribution via Listserv |
| | |
| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION REQUEST FOR ALTERNATIVE REP-SI ACCEPTANCE OF VENT AND DRAIN SOCKET WELDS AT SAFETY INJECTION PUMPS PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWA-4000, at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection (SI) pumps. In order to to complete its review, the NRC staff requests the following additional information. |
| | RAI-EVIB 1.0 The relief request is based on the premise that existing Sl pump vent and drain socket welds may be determined acceptable as-is for continued service. However, the possibility exists that all vent and drain socket welds could fail on an Sl pump. To properly evaluate the relief request, the NRC staff requests that the licensee describe: |
| | (a) The safety significance if all vent and drain weld connections are severed on a single Sl pump. |
| | (b) In case of failure of all vent and drain weld connections on a single Sl pump, please explain if the safety function of nearby equipment and personnel will be affected. |
| | RAI-EVIB 2.0 The relief request partly bases the acceptability of existing Sl pump vent and drain socket welds on the performance of a welding procedure qualification test with representative type 410 stainless steel (P-6) and type 304 stainless steel (P-8) base materials using type 309 filler metal per the production welding procedure parameters without post-weld heat treatment, and that the qualification test assembly passed the required ASME Code destructive tests. However, for this Enclosure |
| | |
| | material combination, the possibility exists that untempered martensite may be formed in the P-6 heat affected zone of the qualification weldment. The NRC staff requests that the licensee discuss the following: |
| | (a) Whether hardness testing was performed in the P-6 heat affected zone of the qualification weldment to objectively measure the possibility and extent of untempered martensite formation. |
| | (b) Since the destructive test specimens were machined down from 0.375-inch in thickness to 0.300-inch in thickness, please justify that the machined destructive test specimens are representative of the qualification weldment in terms of microstructure. Discuss whether the possibility exists for detrimental untempered martensite to be formed during welding but removed during the preparation of the destructive test specimens. |
| | RAI-EPNB 1.0 The licensee submitted the relief request under paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR). The NRC staff believes that the use of the proposed alternative (i.e., continued operation with the existing non-conforming welds) provides a lower level of quality and safety than with ASME Code compliant welds. The NRC staff suggests that it is more appropriate to request the relief pursuant to 10 CFR 50.55a(a)(3)(ii). |
| | Please propose the alternative to the ASME Code under 10 CFR 50.55a(a)(3)(ii) or justify why it is appropriate to submit this relief request under 10 CFR 50.55a(a)(3)(i). |
| | The licensee should also provide hardship justification for not complying with the ASME Code requirements. |
| | RAI-EPNB 2.0 The industry operating experience has shown that socket welds are susceptible to through-wall cracking. |
| | (a) Please discuss whether any leakage detection systems are available to detect any potential leak from the subject socket welds and whether the operators in the control room would be notified of the leakage. |
| | (b) Please discuss the consequence of a through-wall leak and a complete severance at any of the subject socket welds. |
| | RAI-EPNB 3.0 When the safety injection pumps are running; Please discuss whether vibration exists on the drain lines and vent lines. If yes, discuss the potential for the vibration that may cause cracking at the socket welds. |
| | |
| | Structural Integrity Associates Stress Calculation RAI-EPNB 4.0 Section 2.4, page 2-3, of the stress calculation states that unit axial load of 1000 pounds (lbs). |
| | was used as an input. |
| | Please discuss where and how 1000 lbs. was obtained and derived. |
| | RAI-EPNB 5.0 Section 3 of the stress calculation states that when analyzing the outside diameter flaw, the methods of the ASME Code, Section XI, Appendix C, C-7300 and American Petroleum Institute, API-579, were used to obtain stress intensity factors. However, when analyzing the inside diameter flaw, it does not appear that the API-579 method was used. |
| | Please clarify whether API-579 method was used to evaluate the inside diameter flaw. If not, discuss the reference of the flaw evaluation method. |
| | RAI-EPNB 6.0 Section 3.3.2, page 3-8, of the stress calculation states that "... residual stresses would not contribute to fatigue crack growth ... " The NRC staff believes that although residual stresses are steady state in nature (i.e., constant), they affect the maximum tensile stress and R [stress ratio] |
| | and may, therefore, affect fatigue crack growth. |
| | Please clarify the above quoted statement. Discuss whether the above statement is specifically applied to the inside diameter flaw in the subject component. |
| | RAI-EPNB 7.0 Section 4 of the stress calculation discusses fatigue crack growth. Section 4 also discussed a postulated inside diameter initiated flaw. |
| | If an inside diameter flaw is postulated, please discuss the likelihood of stress-corrosion cracking in the subject welds. |
| | RAI-EPNB 8.0 Page 5-1 of the stress calculation states that the allowable flaw depth for an inside diameter flaw exceeds 80 percent of the wall thickness. IWB-3643 of the ASME Code, Section XI, 2003 addenda (Code of record) limits the maximum allowable flaw depth to 75 percent through-wall. |
| | Please explain why the allowable flaw depth for an inside diameter flaw exceeds (i.e., |
| | non-conservative) the ASME Code, Section XI limitation. |
| | |
| | ML14351A047 *email dated OFFICE NRR/DORL/LPL4-1/PM NRR/DORL/LPL4-1/PM NRR/DORLILPL4-1/LA NRR/DE/EPNB NAME RHaskell SLingam JBurkhardt DAiley* |
| | DATE 12/12/2014 12/23/14 12/18/14 10/8/14 OFFICE NRR/DE/EVIB/BC NRR/DORLILPL4-1/BC(A) NRR/DORL/LPL4-1/PM NAME SRosenberg* EOesterle SLingam DATE 11/12/14 12/23/14 12/23/14}} |
|
|---|
Category:Letter
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DCL-2024-507, 2023 Annual Report on Discharge Self-Monitoring2024-02-27027 February 2024
2023 Annual Report on Discharge Self-Monitoring
ML24056A0022024-02-27027 February 2024
– License Renewal Regulatory Audit Regarding the Environmental Review of the License Renewal Application
ML24058A2552024-02-22022 February 2024
Formal Acceptance from Sy Chumash Re Regarding Consultation on the DC NPP
ML24058A2562024-02-22022 February 2024
Santa Ynez Band or Chumash Tribe - Consultation Request Diablo Canyon NPP Units 1 and 2
2024-08-08
[Table view]
Category:Request for Additional Information (RAI)
MONTHYEARML24184C0422024-07-0202 July 2024
NRR E-mail Capture - Request for Additional Information Diablo Canyon 50.69 risk-informed Categorization
ML24065A1312024-03-20020 March 2024
ISFSI Renewal RAI Transmittal Letter
ML24065A1332024-03-20020 March 2024
Enclosure- Diablo Canyon ISFSI Renewal Request for Additional Information
ML24024A2072024-01-24024 January 2024
Inservice Inspection Request for Information
IR 05000275/20240152023-10-10010 October 2023
Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection (050002752024015 and 050003232024015)
ML23249A2782023-09-0606 September 2023
Inservice Inspection Request for Information
ML23159A2372023-07-25025 July 2023
ISFSI Renewal RAI Transmittal Letter Enclosure
ML23159A2382023-07-25025 July 2023
Request for Additional Information for the Technical Review of the Application for Renewal of the Diablo Canyon Independent Spent Fuel Storage Installation (Cac/Epid Nos. 001028/L-2022-RNW-0007)
ML23096A1792023-04-0606 April 2023
NRR E-mail Capture - Request for Additional Information Diablo Canyon Exemption Request Regarding Senior Reactor Operator License Application
ML23047A0062023-02-21021 February 2023
Request for Additional Information Alternative Security Measures for Early Warning System (EPID: L-2022-LLA-0029) (Public Version)
ML23041A1862023-02-17017 February 2023
Request for Information Regarding Diablo Canyon Power Plant, Units 1 and 2 - December 8, 2022, Public Meeting
ML22258A1112022-09-14014 September 2022
2022 Diablo Canyon PIR Request for Information
ML22200A2572022-07-19019 July 2022
Notification of In-service Inspection (Inspection Report 05000323/2022004) and Request for Information
ML22187A2652022-07-0606 July 2022
NRR E-mail Capture - Request for Additional Information: Diablo Canyon Exemption Request for Part 73 force-on-force Training Due to COVID-19
ML22152A1502022-06-21021 June 2022
Request for Additional Information Regarding License Amendment Request for Technical Specifications and Revised License Conditions for the Permanently Defueled Condition
ML22122A1412022-05-0505 May 2022
.02 Doc Request
ML22068A2312022-03-17017 March 2022
Notification of NRC Design Bases Assurance Inspection (Team) 05000275/2022011 and 05000323/2022011 and Initial Request for Information
ML22061A2192022-03-0202 March 2022
NRR E-mail Capture - Request for Additional Information - Diablo Canyon Proposed Changes to Emergency Plan for post-shutdown and Permanently Defueled Condition
ML21363A1692021-12-29029 December 2021
Inservice Inspection Request for Information
ML22019A0412021-12-29029 December 2021
RFI
ML21215A3432021-08-0303 August 2021
NRR E-mail Capture - Request for Additional Information - Diablo Canyon Request to Revise Technical Specification 3.8.1, AC Sources - Operating to Support Diesel Fuel Oil Transfer System Component Planned Maintenance
ML21130A3642021-06-21021 June 2021
DC 2021401 Information Request
ML21104A3642021-04-14014 April 2021
NRR E-mail Capture - Request for Additional Information - Diablo Canyon Proposed Technical Specifications and Revised License Conditions for the Permanently Defueled Condition
ML21062A0642021-03-0202 March 2021
NRR E-mail Capture - Request for Additional Information for Diablo Canyon Generic Letter 2004-02 Submittal (L-2017-LRC-0000)
ML20329A0692020-11-23023 November 2020
NRR E-mail Capture - Request for Additional Information: Diablo Canyon COVID-19 Exemption Request to Extend force-on-force Exercise
ML20323A4532020-11-18018 November 2020
NRR E-mail Capture - Request for Additional Information: Diablo Canyon COVID-19 Exemption Request to Extend Firearms Requalification
ML20301A2212020-10-27027 October 2020
Notification of an NRC Fire Protection Baseline Inspection (NRC Inspection Report 05000275/2021010 and 05000323/2021010) and Request for Information
ML20261H4232020-09-17017 September 2020
NRR E-mail Capture - Request for Additional Information: Diablo Canyon Unit 2 Fall 2019 Steam Generator Tube Inspection Report
ML20231A2372020-08-17017 August 2020
NRR E-mail Capture - Diablo Canyon Additional Request for Additional Information: Exigent License Amendment Request for Application to Provide a New Technical Specification 3.7.5, Auxiliary Feedwater System, Condition G EPID: L-2020-LLA-017
ML20230A0732020-08-14014 August 2020
NRR E-mail Capture - Diablo Canyon Request for Additional Information: Exigent License Amendment Request for Application to Provide a New Technical Specification 3.7.5, Auxiliary Feedwater System, Condition G
ML20280A5432020-06-0303 June 2020
DC 2020 PIR Request for Information
ML20041E6012020-02-10010 February 2020
NRR E-mail Capture - Diablo Canyon Nuclear Power Plant, Units 1 and 2 - Request for Additional Information for Post -Shutdown Decommissioning Activities Report (PSDAR)
ML19262G7482019-08-0909 August 2019
Request for Information
ML19149A6012019-05-28028 May 2019
NRR E-mail Capture - Request for Additional Information (Supplemental) - License Amendment Request (LAR) to Revise Emergency Plan Response Organization Staffing and Augmentation
ML19123A2162019-05-0202 May 2019
NRR E-mail Capture - Diablo Canyon Nuclear Power Plant -Request for Exemption from Operator Written Examination and Operating Test - Request for Additional Information
ML19084A2572019-03-21021 March 2019
NRR E-mail Capture - Request for Additional Information - License Amendment Request (LAR) to Revise Emergency Plan Response Organization Staffing and Augmentation
ML19043A9452019-02-20020 February 2019
Notification of Inspection (NRC Inspection Report 05000275/2019002, 05000323/2019002 and Request for Information
ML17306A9382017-11-0202 November 2017
NRR E-mail Capture - Request for Additional Information - Request for Approval for Application of Full Weld Overlay REP-RHR-SWOL, Diablo Canyon Power Plant, Units 1 and 2
ML17152A3192017-06-0101 June 2017
NRR E-mail Capture - Request for Additional Information (RAI)- Relief Requests NDE-SLH U2, NDE-LSL U2, NDE-LHC U2, NDE-LHM U2, and NDE-ONV U2 (CAC Nos. MF9386 Through MF9390)
ML17102B6072017-04-12012 April 2017
NRR E-mail Capture - Request for Additional Information (RAI) - Revised Emergency Action Level Schemes Pursuant to Nuclear Energy Institute (NEI) 99-01, Revision 6
ML16347A0032016-12-0909 December 2016
NRR E-mail Capture - Request for Additional Information (RAI) - Diablo Canyon Power Plant License Amendment Request for Adoption of NEI 94-01
ML16326A3562016-11-21021 November 2016
NRR E-mail Capture - Request for Additional Information - License Amendment Request to Adopt Nuclear Energy Institute (NEI) 94-01, Revision 2-A for Diablo Canyon Power Plant, Units 1 and 2 - CAC Nos. MF7731 and MF7732
ML16048A2322016-02-17017 February 2016
NRR E-mail Capture - Diablo Canyon 1 and 2 - Met Data Second Round of Requests for Additional Information for License Amendment Request 15-03 to Adopt the Alternative Source Term Per 10 CFR 50.67 (TAC Nos. MF6399 and MF640
ML16011A3652016-02-0202 February 2016
Requests for Additional Information for the Review of the Diablo Canyon Power Plant, Units 1 and 2, License Renewal Application - Set 39 (TAC Nos. ME2896 and ME2897)
ML16011A3172016-01-11011 January 2016
NRR E-mail Capture - Diablo Canyon 1 and 2 - Requests for Additional Information for License Amendment Request 15-03 to Adopt the Alternative Source Term Per 10 CFR 50.67
ML15358A0022015-12-23023 December 2015
Request for Additional Information Email (Follow-up Pra), Request to Adopt National Fire Protection Association NFPA 805, Performance-Based Standard for Fire Protection for LWR Generating Plants (2001 Edition)
ML15357A3822015-12-23023 December 2015
Request for Additional Information, Round 4 - Amendment Request to Replace Digital Process Protection System for Reactor Protection System and Engineered Safety Features Actuation System Functions
ML15295A3732015-11-0505 November 2015
Requests for Additional Information Related to the Diablo Canyon LRA Environmental Review
ML15287A1652015-10-23023 October 2015
Requests for Additional Information Related to the Diablo Canyon LRA Environmental Review - SAMA
ML15267A7742015-10-0101 October 2015
Request for Additional Information Associated with Near-Term Task Force Recommendation 2.1 Seismic Evaluations
2024-07-02
[Table view] |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2014 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - INSERVICE INSPECTION PROGRAM REQUEST FOR ALTERNATIVE REP-SI:
PROPOSED ALTERNATIVE TO REQUIREMENTS FOR REPAIR AND REPLACEMENT ACTIVITIES FOR CERTAIN SAFETY INJECTION PUMP WELDED ATTACHMENTS (TAG NOS. MF4476 AND MF4477)
Dear Mr. Halpin:
By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-4000, "Repair/Replacement Activities,"
including IWA-4130, "Alternative Requirements," and IWA-4131, "Small Items," as corrective action for the four affected Code Class 2, NPS [nominal pipe size] 3/4-inch socket welds on each safety injection (SI) pump at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection pumps.
Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i), the licensee requested NRC approval to use an alternative from the requirements of the ASME Code,Section XI, for Repair/Replacement rules governing socket welded attachments to safety injection pumps.
The NRC staff has reviewed the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the following enclosed request for additional information (RAI). A draft form of these RAis was provided to DCPP licensee staff. During a recent licensing status call on December 9, 2014, between the NRC project manager for DCPP and DCPP licensing staff, it was agreed that no additional licensee clarifications (telecoms) were necessary. Please respond to the enclosed final RAis no later than 45 days from the date of this letter.
E. Halpin If you have any questions, please contact me at 301-415-1564 or via e-mail at Siva.Lingam@nrc.gov.
Sincerely,
~~~*~
Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323
Enclosure:
Request for Additional Information cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION REQUEST FOR ALTERNATIVE REP-SI ACCEPTANCE OF VENT AND DRAIN SOCKET WELDS AT SAFETY INJECTION PUMPS PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 By letter dated July 21, 2014 (Agencywide Documents Access and Management System Accession No. ML14202A613), Pacific Gas and Electric Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-4000, at Diablo Canyon Power Plant (DCPP), Units 1 and 2. The licensee submitted lnservice Inspection Request for Alternative REP-SI for U.S. Nuclear Regulatory Commission (NRC) review and approval. The relief request is related to acceptance of socket welds at the vent and drain lines associated with the three safety injection (SI) pumps. In order to to complete its review, the NRC staff requests the following additional information.
RAI-EVIB 1.0 The relief request is based on the premise that existing Sl pump vent and drain socket welds may be determined acceptable as-is for continued service. However, the possibility exists that all vent and drain socket welds could fail on an Sl pump. To properly evaluate the relief request, the NRC staff requests that the licensee describe:
(a) The safety significance if all vent and drain weld connections are severed on a single Sl pump.
(b) In case of failure of all vent and drain weld connections on a single Sl pump, please explain if the safety function of nearby equipment and personnel will be affected.
RAI-EVIB 2.0 The relief request partly bases the acceptability of existing Sl pump vent and drain socket welds on the performance of a welding procedure qualification test with representative type 410 stainless steel (P-6) and type 304 stainless steel (P-8) base materials using type 309 filler metal per the production welding procedure parameters without post-weld heat treatment, and that the qualification test assembly passed the required ASME Code destructive tests. However, for this Enclosure
material combination, the possibility exists that untempered martensite may be formed in the P-6 heat affected zone of the qualification weldment. The NRC staff requests that the licensee discuss the following:
(a) Whether hardness testing was performed in the P-6 heat affected zone of the qualification weldment to objectively measure the possibility and extent of untempered martensite formation.
(b) Since the destructive test specimens were machined down from 0.375-inch in thickness to 0.300-inch in thickness, please justify that the machined destructive test specimens are representative of the qualification weldment in terms of microstructure. Discuss whether the possibility exists for detrimental untempered martensite to be formed during welding but removed during the preparation of the destructive test specimens.
RAI-EPNB 1.0 The licensee submitted the relief request under paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR). The NRC staff believes that the use of the proposed alternative (i.e., continued operation with the existing non-conforming welds) provides a lower level of quality and safety than with ASME Code compliant welds. The NRC staff suggests that it is more appropriate to request the relief pursuant to 10 CFR 50.55a(a)(3)(ii).
Please propose the alternative to the ASME Code under 10 CFR 50.55a(a)(3)(ii) or justify why it is appropriate to submit this relief request under 10 CFR 50.55a(a)(3)(i).
The licensee should also provide hardship justification for not complying with the ASME Code requirements.
RAI-EPNB 2.0 The industry operating experience has shown that socket welds are susceptible to through-wall cracking.
(a) Please discuss whether any leakage detection systems are available to detect any potential leak from the subject socket welds and whether the operators in the control room would be notified of the leakage.
(b) Please discuss the consequence of a through-wall leak and a complete severance at any of the subject socket welds.
RAI-EPNB 3.0 When the safety injection pumps are running; Please discuss whether vibration exists on the drain lines and vent lines. If yes, discuss the potential for the vibration that may cause cracking at the socket welds.
Structural Integrity Associates Stress Calculation RAI-EPNB 4.0 Section 2.4, page 2-3, of the stress calculation states that unit axial load of 1000 pounds (lbs).
was used as an input.
Please discuss where and how 1000 lbs. was obtained and derived.
RAI-EPNB 5.0 Section 3 of the stress calculation states that when analyzing the outside diameter flaw, the methods of the ASME Code,Section XI, Appendix C, C-7300 and American Petroleum Institute, API-579, were used to obtain stress intensity factors. However, when analyzing the inside diameter flaw, it does not appear that the API-579 method was used.
Please clarify whether API-579 method was used to evaluate the inside diameter flaw. If not, discuss the reference of the flaw evaluation method.
RAI-EPNB 6.0 Section 3.3.2, page 3-8, of the stress calculation states that "... residual stresses would not contribute to fatigue crack growth ... " The NRC staff believes that although residual stresses are steady state in nature (i.e., constant), they affect the maximum tensile stress and R [stress ratio]
and may, therefore, affect fatigue crack growth.
Please clarify the above quoted statement. Discuss whether the above statement is specifically applied to the inside diameter flaw in the subject component.
RAI-EPNB 7.0 Section 4 of the stress calculation discusses fatigue crack growth. Section 4 also discussed a postulated inside diameter initiated flaw.
If an inside diameter flaw is postulated, please discuss the likelihood of stress-corrosion cracking in the subject welds.
RAI-EPNB 8.0 Page 5-1 of the stress calculation states that the allowable flaw depth for an inside diameter flaw exceeds 80 percent of the wall thickness. IWB-3643 of the ASME Code,Section XI, 2003 addenda (Code of record) limits the maximum allowable flaw depth to 75 percent through-wall.
Please explain why the allowable flaw depth for an inside diameter flaw exceeds (i.e.,
non-conservative) the ASME Code,Section XI limitation.
ML14351A047 *email dated OFFICE NRR/DORL/LPL4-1/PM NRR/DORL/LPL4-1/PM NRR/DORLILPL4-1/LA NRR/DE/EPNB NAME RHaskell SLingam JBurkhardt DAiley*
DATE 12/12/2014 12/23/14 12/18/14 10/8/14 OFFICE NRR/DE/EVIB/BC NRR/DORLILPL4-1/BC(A) NRR/DORL/LPL4-1/PM NAME SRosenberg* EOesterle SLingam DATE 11/12/14 12/23/14 12/23/14
