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| | number = ML16308A401 | | | number = ML16308A401 |
| | issue date = 11/16/2016 | | | issue date = 11/16/2016 |
| | title = Unit 1 - Followup Request for Additional Information (CAC Nos. MF8061 and MF8062) | | | title = Followup Request for Additional Information |
| | author name = Martin R E | | | author name = Martin R |
| | author affiliation = NRC/NRR/DORL/LPLII-1 | | | author affiliation = NRC/NRR/DORL/LPLII-1 |
| | addressee name = Pierce C R | | | addressee name = Pierce C |
| | addressee affiliation = Southern Nuclear Operating Co, Inc | | | addressee affiliation = Southern Nuclear Operating Co, Inc |
| | docket = 05000424, 05000425 | | | docket = 05000424, 05000425 |
| | license number = NPF-068, NPF-081 | | | license number = NPF-068, NPF-081 |
| | contact person = Martin R E, NRR/DORL/LPL2-1 | | | contact person = Martin R, NRR/DORL/LPL2-1 |
| | case reference number = CAC MF8061, CAC MF8062 | | | case reference number = CAC MF8061, CAC MF8062 |
| | 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 November 16, 2016 Mr. Charles R. Pierce Regulatory Affairs Director Southern Nuclear Operating Co., Inc. P.O. Box 1295, Bin 038 Birmingham, AL 35201-1295 | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 16, 2016 Mr. Charles R. Pierce Regulatory Affairs Director Southern Nuclear Operating Co., Inc. |
| | P.O. Box 1295, Bin 038 Birmingham, AL 35201-1295 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| VOGTLE ELECTRIC GENERATING PLANT, UNITS 1AND2, AND JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 -FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. MF8061 AND MF8062) | | VOGTLE ELECTRIC GENERATING PLANT, UNITS 1AND2, AND JOSEPH M. |
| | FARLEY NUCLEAR PLANT, UNIT 1 - FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. MF8061 AND MF8062) |
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| ==Dear Mr. Pierce:== | | ==Dear Mr. Pierce:== |
| By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML 16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-AL T-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-AL T-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-ALT-11, Version 2.0, and FNP-ISl-ALT-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components. | | |
| The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML 16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML 16298A049), and has determined that the enclosed followup RAI is needed to complete its review. | | By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-ALT-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-ALT-11, Version 2.0, and FNP-ISl-ALT-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components. |
| Please provide responses to these questions within 30 days of the date of this letter. Docket Nos. 50-424, 50-425, and 50-348 | | The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the enclosed followup RAI is needed to complete its review. Please provide responses to these questions within 30 days of the date of this letter. |
| | Sincerely, |
| | (_.,/; ll/ 1VJJ::{-' |
| | b Mahin, Senior Project Manager nt Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424, 50-425, and 50-348 |
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| ==Enclosure:== | | ==Enclosure:== |
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| Followup Request for Additional Information Sincerely, (_.,/; ll/ 1VJJ::{-' | | Followup Request for Additional Information |
| b Mahin, Senior Project Manager nt Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION ALTERNATIVE REQUESTS VEGP-ISl-AL T-11, VERSION 2.0 AND FNP-ISl-ALT-19, VERSION 2.0 REACTOR PRESSURE VESSEL FLANGE BOLT HOLE THREAD EXAMINATION SOUTHERN NUCLEAR OPERATING COMPANY, INC. VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-424, 50-425, 50-348 By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML 16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-AL T-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-AL T-11, Version 2.0, and FNP-ISl-AL T-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components.
| | |
| The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML 16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML 16298A049), and has determined that the following questions are needed to complete its review. Followup RAl-1 (Related to NRC RAl-5, Question 2a) The licensee's response to the previous RAl-5, Question 2a, indicated that, "Thermal loads are applied as uniform surface convection on the inside surface only." This response did not clarify how the heatup transient was applied. Please provide: (a) the thermal boundary conditions for the top, bottom, and RPV flange outer surfaces to confirm that this part of modeling is appropriate, and (b) a revision of the response to this RAI regarding how the heatup transient was applied to the thermal model since the application of thermal loads was not answered clearly. Enclosure Followup RAl-2 (Related to Previous RAl-5, Question 3) Regarding selection of the heatup transient instead of the cooldown transient in the finite element method (FEM) analysis, the licensee's response to previous RAl-5, Question 3, states that, "Since heatup and cooldown have the same temperature change rate, in linear elastic analysis they will produce identical maximum and minimum stress range for crack growth calculation, despite an opposite time history." The above description of stresses does not appear to be consistent with the similar pressure-temperature (PT) limits application (ignoring the crack growth part because it does not apply to the PT limit application), wherein the cooldown transient will create tensile stresses in the RPV inner wall and compressive stresses in the outer wall, and vice versa for the heatup transient. | | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION ALTERNATIVE REQUESTS VEGP-ISl-ALT-11, VERSION 2.0 AND FNP-ISl-ALT-19, VERSION 2.0 REACTOR PRESSURE VESSEL FLANGE BOLT HOLE THREAD EXAMINATION SOUTHERN NUCLEAR OPERATING COMPANY, INC. |
| Please provide additional discussion on your response to previous RAl-5, Question 3, to justify that heatup and cooldown transients will produce identical maximum and minimum stress ranges. Followup RAl-3 (Related to Previous RAl-5, Question 4a) The licensee's response to previous RAl-5, Question 4a, indicated that, " ... the FEM model for the applied K determination is the same as the FEM model for the stress determination." Please clarify how loads are applied to both the FEM model for the stress determination and the FEM model for the applied K determination.
| | VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-424, 50-425, 50-348 By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-ALT-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-ALT-11, Version 2.0, and FNP-ISl-ALT-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components. |
| Followup RAl-4 (Related to Previous RAl-5, Question 6) The licensee's response to previous RAl-5, Question 6, indicated that, " ... the maximum calculated Kat any crack depth is about 20ksi"in.
| | The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the following questions are needed to complete its review. |
| This requires a Kie of 20*"10 = 3.2ksi"in." The Kie of 3.2 ksi"in may be a misprint of 63.2 ksi"in. Please provide the operating temperature at the time when K is 20 ksi"in to justify that "an RT NDT of up to 70°F will not affect the results."
| | Followup RAl-1 (Related to NRC RAl-5, Question 2a) |
| Mr. Charles R. Pierce Regulatory Affairs Director Southern Nuclear Operating Co., Inc. P.O. Box 1295, Bin 038 Birmingham, AL 35201-1295 November 16, 2016
| | The licensee's response to the previous RAl-5, Question 2a, indicated that, "Thermal loads are applied as uniform surface convection on the inside surface only." This response did not clarify how the heatup transient was applied. Please provide: (a) the thermal boundary conditions for the top, bottom, and RPV flange outer surfaces to confirm that this part of modeling is appropriate, and (b) a revision of the response to this RAI regarding how the heatup transient was applied to the thermal model since the application of thermal loads was not answered clearly. |
| | Enclosure |
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| ==SUBJECT:==
| | Followup RAl-2 (Related to Previous RAl-5, Question 3) |
| VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, AND JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 -FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. MF8061 AND MF8062)
| | Regarding selection of the heatup transient instead of the cooldown transient in the finite element method (FEM) analysis, the licensee's response to previous RAl-5, Question 3, states that, "Since heatup and cooldown have the same temperature change rate, in linear elastic analysis they will produce identical maximum and minimum stress range for crack growth calculation, despite an opposite time history." The above description of stresses does not appear to be consistent with the similar pressure-temperature (PT) limits application (ignoring the crack growth part because it does not apply to the PT limit application), wherein the cooldown transient will create tensile stresses in the RPV inner wall and compressive stresses in the outer wall, and vice versa for the heatup transient. Please provide additional discussion on your response to previous RAl-5, Question 3, to justify that heatup and cooldown transients will produce identical maximum and minimum stress ranges. |
| | Followup RAl-3 (Related to Previous RAl-5, Question 4a) |
| | The licensee's response to previous RAl-5, Question 4a, indicated that, " ... the FEM model for the applied K determination is the same as the FEM model for the stress determination." |
| | Please clarify how loads are applied to both the FEM model for the stress determination and the FEM model for the applied K determination. |
| | Followup RAl-4 (Related to Previous RAl-5, Question 6) |
| | The licensee's response to previous RAl-5, Question 6, indicated that, " ... the maximum calculated Kat any crack depth is about 20ksi"in. This requires a Kie of 20*"10 = 3.2ksi"in." |
| | The Kie of 3.2 ksi"in may be a misprint of 63.2 ksi"in. Please provide the operating temperature at the time when K is 20 ksi"in to justify that "an RT NDT of up to 70°F will not affect the results." |
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| ==Dear Mr. Pierce:==
| | ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the enclosed followup RAI is needed to complete its review. Please provide responses to these questions within 30 days of the date of this letter. |
| By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML 16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-AL T-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-AL T-11, Version 2.0, and FNP-ISl-AL T-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components.
| | Sincerely, IRA/ |
| The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML 16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML 16298A049), and has determined that the enclosed followup RAI is needed to complete its review. Please provide responses to these questions within 30 days of the date of this letter. Docket Nos. 50-424, 50-425, and 50-348
| | Bob Martin, Senior Project Manager Plant Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424, 50-425, and 50-348 |
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| ==Enclosure:== | | ==Enclosure:== |
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| Sincerely, IRA/ Bob Martin, Senior Project Manager Plant Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Followup Request for Additional Information DISTRIBUTION:
| | Followup Request for Additional Information DISTRIBUTION: |
| PUBLIC LPL2-1 R/F RidsNrrDeEvib Resource RidsNrrLALRonewicz Resource RidsNrrPMFarley Resource RidsNrrDorllpl2-1 Resource RidsRgn2MailCenter Resource RidsACRS_MailCTR Resource ADAMS A ccess1on N ML 16308A401 o.: RidsNrrPMVogtle Resource SSheng, NRR SWilliams, NRR *b *1 1y e-ma1 OFFICE DORL/LPL2-1 | | PUBLIC LPL2-1 R/F RidsNrrDorllpl2-1 Resource RidsNrrPMVogtle Resource RidsNrrDeEvib Resource RidsRgn2MailCenter Resource SSheng, NRR RidsNrrLALRonewicz Resource RidsACRS_MailCTR Resource SWilliams, NRR RidsNrrPMFarley Resource ADAMS A ccess1on No.: ML16308A401 *b1y e-ma1*1 OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA DE/EVIB/BC* DORL/LPL2-1 /BC DORL/LPL2-1 /PM NAME BMartin LRonewicz DRudland MMarkley BMartin DATE 11/16/2016 11/4/2016 11/02/2016 11/16/2016 11/16/2016}} |
| /PM DORL/LPL2-1 | |
| /LA DE/EVIB/BC* | |
| DORL/LPL2-1 | |
| /BC DORL/LPL2-1 | |
| /PM NAME BMartin LRonewicz DRudland MMarkley BMartin DATE 11/16/2016 11/4/2016 11/02/2016 11/16/2016 11/16/2016 OFFICIAL RECORD COPY}} | |
Letter Sequence RAI |
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MONTHYEARNL-16-0724, Proposed Lnservice Inspection Alternative VEGP-ISI-ALT-11, Version 1.02016-06-28028 June 2016
Proposed Lnservice Inspection Alternative VEGP-ISI-ALT-11, Version 1.0
Project stage: Request
NL-16-0723, Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 1.02016-06-30030 June 2016
Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 1.0
Project stage: Request
ML16204A0422016-07-26026 July 2016
Acceptance Letter, Unacceptable, with Opportunity to Supplement
Project stage: Acceptance Review
NL-16-1329, Proposed Inservice Inspection Alternative VEGP-ISI-ALT-11, Version 2.0 and Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 2.02016-08-0404 August 2016
Proposed Inservice Inspection Alternative VEGP-ISI-ALT-11, Version 2.0 and Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 2.0
Project stage: Request
ML16224A1432016-08-23023 August 2016
Acceptance Review Regarding Reactor Vessel Threads in Flange Examination
Project stage: Acceptance Review
ML16256A2212016-09-15015 September 2016
Request for Additional Information
Project stage: RAI
ML16308A4012016-11-16016 November 2016
Followup Request for Additional Information
Project stage: RAI
NL-16-2513, Response to Follow-up Request for Information Regarding Reactor Pressure Vessel Threads-in-Flange Examination Requirement2016-11-23023 November 2016
Response to Follow-up Request for Information Regarding Reactor Pressure Vessel Threads-in-Flange Examination Requirement
Project stage: Request
ML17006A1092017-01-26026 January 2017
Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads Inflange Inspection
Project stage: Other
2016-07-26
[Table View] |
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Category:Letter
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IR 05000424/20240022024-07-29029 July 2024
Integrated Inspection Report 05000424/2024002 and 05000425/2024002
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NL-24-0126, – Units 3 and 4, License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Action a and SR 3.7.6.62024-07-25025 July 2024
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ML24204A0722024-07-23023 July 2024
Issuance of Amendment No. 225, Regarding LAR to Revise TS 3.7.9 for a one-time Change to Support Nuclear Service Cooling Water Transfer Pump Repairs - Emergency Circumstances
NL-24-0286, Emergency Request to Revise Technical Specification 3.7.9 for a One-Time Change to Support a Unit 1 Nuclear Service Cooling Water Transfer Pump Repair2024-07-20020 July 2024
Emergency Request to Revise Technical Specification 3.7.9 for a One-Time Change to Support a Unit 1 Nuclear Service Cooling Water Transfer Pump Repair
NL-24-0261, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20232024-07-19019 July 2024
10 CFR 50.46 ECCS Evaluation Model Annual Report for 2023
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ML24194A0342024-07-12012 July 2024
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ML24191A3792024-07-10010 July 2024
– Initial Test Program and Operational Programs Inspection Report 05200026/2024011
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Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown
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NL-24-0087, License Amendment Request: Changes to Technical Specification 3.7.6, Main Control Room Emergency Habitability System (Ves) Air Storage Tanks2024-06-21021 June 2024
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Expedited Issuance of Amendment No. 198 Change to Technical Specification 5.5.13, Ventilation Filter Testing Program (VFTP)
NL-24-0191, Annual Radiological Environmental Operating Reports for 20232024-05-10010 May 2024
Annual Radiological Environmental Operating Reports for 2023
ML24127A2372024-05-0909 May 2024
Initial Test Program and Operational Programs Inspection Report 05200026/2024010
NL-24-0194, Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1)2024-05-0707 May 2024
Revised Request for Relief and Alternative Requirements for Squib Valves First Test Interval (V34-IST-ALT-03-R1)
ML24120A2832024-04-30030 April 2024
Project Manager Reassignment
2024-09-09
[Table view]
Category:Request for Additional Information (RAI)
MONTHYEARML24290A1572024-10-16016 October 2024
NRR E-mail Capture - for Your Action - RAI - Farley and Vogtle 1 and 2 - Alternative Request for Pressurizer Welds (L-2024-LLR-0047)
ML24290A1552024-10-16016 October 2024
NRR E-mail Capture - for Your Action - RAI - Farley, Hatch, and Vogtle 1 and 2 - Proposed Alternative Request for Code Case N-572
ML24289A2002024-10-15015 October 2024
NRR E-mail Capture - for Your Action - RAI - Proposed License Amendment Request (LAR) for Vogtle Electric Generating Plant, Units 3 and 4 TS 3.7.6 Action E and SR 3.7.6.5 (L-2024-LLA-0083)
ML24199A1592024-07-17017 July 2024
NRR E-mail Capture - Request for Additional Information - Vogtle 3 and 4 - Exemption for the Requirements in 10 CFR 50.71 Pertaining to the Submittal of Updated Final Safety Analysis Report
ML24100A7842024-04-0909 April 2024
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) - Vogtle, Units 3 and 4 - Alternative Request for Explosively Actuated Valves (L-2023-LLR-0016)
ML24072A3982024-03-12012 March 2024
NRR E-mail Capture - for Your Action - Second Round Request for Additional Information (RAI) - Vogtle, Unit 1 - SG Tube Inspection Report - 1R24 (L-2023-LRO-0067)
IR 05000424/20244032024-01-26026 January 2024
Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000424/2024403; 05000425/2024403
ML23341A2042024-01-12012 January 2024
Request for Additional Information Exemption Requests for Physical Barriers (EPID L-2023-LLE-0018 & L-2023-LLE-0021)
ML23342A0802023-12-0808 December 2023
NRR E-mail Capture - Request for Additional Information (RAI) - Vogtle, Unit 1 - Review of SG Tube Inspection Report for Refueling Outage 24 (L-2023-LRO-0067)
ML23279A2082023-10-0505 October 2023
Issuance of Formal RAIs - Vogtle, Units 1 and 2 - Proposed LAR and Proposed Alternative Request to Revise TS 3.4.14.1 and IST ALT-VR-02 (EPIDs L-2023-LLA-0061 and L-2023-LLR-0023)
ML23257A2092023-09-14014 September 2023
NRR E-mail Capture - Formal Issuance of 2nd Round RAIs for Surry Units 1&2 and North Anna Units 1&2 Emergency Plans LAR
ML23248A3482023-09-0505 September 2023
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) 6 - Vogtle - TSTF-339 LAR (L-2023-LLA-0053)
ML23243A9862023-08-31031 August 2023
NRR E-mail Capture - Draft RAIs for EP Staff Augmentation Times LAR (L-2022-LLA-0166)
ML23188A1512023-08-0909 August 2023
Round 2 RAIs for LAR 22-002 TS 3.8.3 Inverters-Operating, Completion Time Extension
ML23200A0112023-07-18018 July 2023
Document Request for Vogtle Unit 3 RP Inspection
ML23198A1552023-07-17017 July 2023
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) - Hatch, Farley, and Vogtle, Units 1 and 2 Quality Assurance Topical Report (QATR) Submittal Dated June 15, 2023
ML23193A7832023-07-12012 July 2023
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) - Vogtle - TSTF-339 LAR (L-2023-LLA-0053)
ML23065A0612023-03-0303 March 2023
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) 14 - Vogtle - AST, TSTF-51, TSTF-471, and TSTF-490 LAR (L-2022-LLA-0096)
ML23006A0882023-01-0606 January 2023
NRR E-mail Capture - for Your Action - Request for Additional Information (RAI) - Vogtle - Accident Source Term (Ast), TSTF-51, TSTF-471, and TSTF-490 LAR (L-2022-LLA-0096)
ML22348A0332022-12-13013 December 2022
NRR E-mail Capture - for Your Action - RAI - Vogtle, Unit 2 - Steam Generator Tube Inspection Report (L-2022-LRO-0120)
ML22192A1042022-08-0101 August 2022
Acceptance of Requested Licensing Action Amendment Request Application to Allow Use of Lead Test Assemblies for Accident Tolerant Fuel with Request for Additional Information
ML22157A0902022-06-0606 June 2022
NRR E-mail Capture - RAIs - Vogtle, Unit 1 - Refueling Outage (RFO) 1R23 Steam Generator (SG) Tube Inspection Report
ML22104A1312022-04-14014 April 2022
NRR E-mail Capture - Request for Additional Information - Farley and Vogtle - Relocate Piping Inspection License Amendment Request (L-2021-LLA-0235)
ML22026A3942022-01-26026 January 2022
NRR E-mail Capture - Request for Additional Information - Vogtle, Units 1 and 2, TS 3.7.2 LAR
ML21350A1032021-12-16016 December 2021
NRR E-mail Capture - Request for Additional Information - Vogtle, Units 1 and 2, TS 3.7.2 LAR
ML21321A3762021-11-15015 November 2021
NRR E-mail Capture - Request for Additional Information - Vogtle, Units 1 and 2, TS 3.7.2 LAR
ML21075A0032021-03-12012 March 2021
Emergency Preparedness Exercise Inspection Request for Information for - Brunswick, Catawba, North Anna, Oconee, Vogtle 1 & 2
ML21033B1072021-02-0202 February 2021
NRR E-mail Capture - RAI - Vogtle Unit 1 - SG Report (L-2020-LRO-0059)
ML20338A1512020-12-0303 December 2020
NRR E-mail Capture - RAIs for Vogtle GSI-191 LAR (L-2020-LLA-0182)
ML20325A0432020-11-20020 November 2020
NRR E-mail Capture - for Your Review - Draft RAIs for Vogtle GSI-191 LAR (L-2020-LLA-0182)
ML20297A3052020-10-22022 October 2020
NRR E-mail Capture - RAIs for Vogtle Relief Request - EPRI Report (L-2020-LLR-0109)
ML20293A0752020-10-14014 October 2020
NRR E-mail Capture - RAIs for SNC Fleet EP LAR (L-2020-LLA-0150 and L-2020-LLA-0151)
ML20149K6252020-05-27027 May 2020
NRR E-mail Capture - RAIs for SNC Fleet Fire Protection Exemption Requests
ML19263A6432019-09-19019 September 2019
NRR E-mail Capture - RAI - Vogtle End State License Amendment Request (LAR) to Revise the Technical Specifications (Tss) for Vogtle, Units 1 and 2 (L-2019-LLA-0148)
ML19156A1872019-06-0505 June 2019
NRR E-mail Capture - Request for Additional Information (RAI) for Vogtle Unit 2 Core Operating Limits Report, Cycle 21
ML19105B1632019-04-15015 April 2019
NRR E-mail Capture - Request for Additional Information (RAI) for Voglte TSTF-412 Adoption LAR (L-2018-LLA-0731)
ML18355A4772019-01-0404 January 2019
Request for Additional Information Revise TS 5.2.2.g and Update Emergency Plan Minimum On-Shift Staff Tables
ML18337A4032018-12-0606 December 2018
Request for Additional Information Revise Technical Specification 5.2.2.G and Updating Emergency Plan Minimum On-Shift Staff Tables
ML18236A4452018-08-30030 August 2018
Review of Response to RAI License Amendment Request for Approval to Utilize the Tornado Missile Risk Evaluator to Analyze Tornado Missile Protection Non-Conformances
ML18227A0222018-08-15015 August 2018
Notification of Inspection and Request for Information
ML18225A3362018-08-13013 August 2018
NRR E-mail Capture - Request for Additional Information Regarding Relief Request VEGP-ISI-RR-03
ML18197A0602018-07-13013 July 2018
Us NRC Final RAI No. 1 for Vogtle 3 and 4 LAR-18-015, Fire Protection System Non-Safety Cable Spray Removal
ML18197A1882018-07-13013 July 2018
Us NRC Final Request for Additional Information (Final RAI) No. 4 on Vogtle LAR-17-024, Technical Specification Updates for Reactivity Controls and Other Miscellaneous Changes
ML18197A1052018-07-13013 July 2018
Us NRC Final Request for Additional Information (Final RAI) No. 1 on Vogtle 3 and 4 LAR-17-043, Containment Pressure Analysis
ML18192C0782018-07-11011 July 2018
RAI (LAR 18-009) - Vogtle Electric Generating Plant, Unit 3 and 4
ML18192B7692018-07-11011 July 2018
Us NRC Draft Request for Additional Information (Draft RAI) No. 4 on Vogtle LAR-17-024, Technical Specification Updates for Reactivity Controls and Other Miscellaneous Changes
ML18136A4972018-06-0101 June 2018
Request for Additional Information, License Amendment Request to Incorporate Tornado Missile Risk Evaluator Methodology Into Licensing Basis
ML18109A1152018-05-0101 May 2018
Request for Additional Information
ML18079A9572018-03-28028 March 2018
Request for Additional Information Incorporate Seismic Probabilistic Risk Assessment Into the 10 CFR 50.69 Categorization Process (CAC Nos. MF9861 and MF9862; EPID L-2017-LLA-0248)
ML18058A0812018-02-27027 February 2018
Enclosurequest for Additional Information(Rai for Review of Southern Nuclear Operating Company'S Decommissioning Funding Plan Updates for Joseph M. Farley,Unit 1 and 2;Edwin I. Hatch,Units 1 and 2; and Vogtle Electric Generating Plant, Unit
2024-07-17
[Table view] |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 16, 2016 Mr. Charles R. Pierce Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
P.O. Box 1295, Bin 038 Birmingham, AL 35201-1295
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1AND2, AND JOSEPH M.
FARLEY NUCLEAR PLANT, UNIT 1 - FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. MF8061 AND MF8062)
Dear Mr. Pierce:
By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-ALT-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-ALT-11, Version 2.0, and FNP-ISl-ALT-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the enclosed followup RAI is needed to complete its review. Please provide responses to these questions within 30 days of the date of this letter.
Sincerely,
(_.,/; ll/ 1VJJ::{-'
b Mahin, Senior Project Manager nt Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424, 50-425, and 50-348
Enclosure:
Followup Request for Additional Information
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION ALTERNATIVE REQUESTS VEGP-ISl-ALT-11, VERSION 2.0 AND FNP-ISl-ALT-19, VERSION 2.0 REACTOR PRESSURE VESSEL FLANGE BOLT HOLE THREAD EXAMINATION SOUTHERN NUCLEAR OPERATING COMPANY, INC.
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-424, 50-425, 50-348 By application dated August 4, 2016 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML16221A072), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted Request for Alternative VEGP-ISl-ALT-11, Version 2.0, for the Vogtle Electric Generating Plant, Units 1 and 2, and Alternative FNP-ISl-ALT-19, Version 2.0, for the Joseph M. Farley Nuclear Plant, Unit 1. Alternatives VEGP-ISl-ALT-11, Version 2.0, and FNP-ISl-ALT-19, Version 2.0, propose to eliminate the reactor pressure vessel (RPV) threads-in-flange examination requirement as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for inservice inspection of RPV components.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's previous responses, dated October 24, 2016 (ADAMS Accession No. ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the following questions are needed to complete its review.
Followup RAl-1 (Related to NRC RAl-5, Question 2a)
The licensee's response to the previous RAl-5, Question 2a, indicated that, "Thermal loads are applied as uniform surface convection on the inside surface only." This response did not clarify how the heatup transient was applied. Please provide: (a) the thermal boundary conditions for the top, bottom, and RPV flange outer surfaces to confirm that this part of modeling is appropriate, and (b) a revision of the response to this RAI regarding how the heatup transient was applied to the thermal model since the application of thermal loads was not answered clearly.
Enclosure
Followup RAl-2 (Related to Previous RAl-5, Question 3)
Regarding selection of the heatup transient instead of the cooldown transient in the finite element method (FEM) analysis, the licensee's response to previous RAl-5, Question 3, states that, "Since heatup and cooldown have the same temperature change rate, in linear elastic analysis they will produce identical maximum and minimum stress range for crack growth calculation, despite an opposite time history." The above description of stresses does not appear to be consistent with the similar pressure-temperature (PT) limits application (ignoring the crack growth part because it does not apply to the PT limit application), wherein the cooldown transient will create tensile stresses in the RPV inner wall and compressive stresses in the outer wall, and vice versa for the heatup transient. Please provide additional discussion on your response to previous RAl-5, Question 3, to justify that heatup and cooldown transients will produce identical maximum and minimum stress ranges.
Followup RAl-3 (Related to Previous RAl-5, Question 4a)
The licensee's response to previous RAl-5, Question 4a, indicated that, " ... the FEM model for the applied K determination is the same as the FEM model for the stress determination."
Please clarify how loads are applied to both the FEM model for the stress determination and the FEM model for the applied K determination.
Followup RAl-4 (Related to Previous RAl-5, Question 6)
The licensee's response to previous RAl-5, Question 6, indicated that, " ... the maximum calculated Kat any crack depth is about 20ksi"in. This requires a Kie of 20*"10 = 3.2ksi"in."
The Kie of 3.2 ksi"in may be a misprint of 63.2 ksi"in. Please provide the operating temperature at the time when K is 20 ksi"in to justify that "an RT NDT of up to 70°F will not affect the results."
ML16298A049), to the NRC staff's Request for Additional Information (RAI) (ADAMS Accession No. ML16298A049), and has determined that the enclosed followup RAI is needed to complete its review. Please provide responses to these questions within 30 days of the date of this letter.
Sincerely, IRA/
Bob Martin, Senior Project Manager Plant Licensing Branch, 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424, 50-425, and 50-348
Enclosure:
Followup Request for Additional Information DISTRIBUTION:
PUBLIC LPL2-1 R/F RidsNrrDorllpl2-1 Resource RidsNrrPMVogtle Resource RidsNrrDeEvib Resource RidsRgn2MailCenter Resource SSheng, NRR RidsNrrLALRonewicz Resource RidsACRS_MailCTR Resource SWilliams, NRR RidsNrrPMFarley Resource ADAMS A ccess1on No.: ML16308A401 *b1y e-ma1*1 OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA DE/EVIB/BC* DORL/LPL2-1 /BC DORL/LPL2-1 /PM NAME BMartin LRonewicz DRudland MMarkley BMartin DATE 11/16/2016 11/4/2016 11/02/2016 11/16/2016 11/16/2016
