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MONTHYEARML23013A2242023-01-13013 January 2023 Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure Temperature Limitation Figures Project stage: Request ML23038A0152023-02-0707 February 2023 Acceptance Review Determination LAR to Revise the Applicability Term for RCS Heatup and Cooldown Pressure-Temperature Limitations Figures Project stage: Acceptance Review RA-18-0007, License Amendment Application to Revise Control Room Cooling Technical Specifications2023-06-19019 June 2023 License Amendment Application to Revise Control Room Cooling Technical Specifications Project stage: Request ML23199A2832023-07-18018 July 2023 Request for Additional Information (E-mail Dated 7/18/2023) LAR to Revise the Pressure-Temperature Limits Project stage: RAI ML23248A2132023-08-30030 August 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature. Project stage: Response to RAI ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures Project stage: Approval ML24024A2452024-01-24024 January 2024 1/24/2024 E-mail from R.Guzman to S.Sinha - Acknowledgement of Error in SE for Amendment No. 288 - Revision to Applicability Term for RCS Heatup and Cooldown Pressure-Temperature Limitations Figures Project stage: Other ML24024A0202024-01-24024 January 2024 Correction to Safety Evaluation Associated Wit Amendment No. 288 - Revision to Applicability Term for RCS Heat-Up and Cooldown Pressure-Temperature Limitations Figures Project stage: Approval ML24017A0652024-03-0808 March 2024 Issuance of Amendment Nos. 319 and 315 to Technical Specification 3.7.11, Control Room Area Chill Water System (Cracws) Project stage: Approval 2023-07-18
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Category:Letter
MONTHYEARIR 05000336/20240032024-11-0707 November 2024 Integrated Inspection Report 05000336/2024003 and 05000423/2024003 and Apparent Violation and Independent Spent Fuel Storage Installation Inspection Report 07200008/2024001 ML24289A0152024-10-21021 October 2024 Review of the Fall 2023 Steam Generator Tube Inspection Report 05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024 Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary IR 05000336/20244022024-10-0808 October 2024 Security Baseline Inspection Report 05000336/2024402 and 05000423/2024402 (Cover Letter Only) ML24281A1102024-10-0707 October 2024 Requalification Program Inspection 05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status ML24260A2192024-09-16016 September 2024 Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification ML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24248A2272024-09-0404 September 2024 Operator Licensing Examination Approval ML24240A1532024-09-0303 September 2024 Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel IR 05000336/20240052024-08-29029 August 2024 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005 IR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 ML24221A2872024-08-0808 August 2024 Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage IR 05000336/20244412024-08-0606 August 2024 Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only) ML24212A0742024-08-0505 August 2024 Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification ML24211A1712024-07-25025 July 2024 Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change IR 05000336/20244032024-07-22022 July 2024 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403 IR 05000336/20245012024-07-0101 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501 ML24180A0932024-06-28028 June 2024 Readiness for Additional Inspection: EA-23-144 IR 05000336/20240102024-06-26026 June 2024 Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010 ML24178A2422024-06-25025 June 2024 2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46 IR 05000336/20244402024-06-24024 June 2024 Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only) ML24176A1782024-06-20020 June 2024 Update to the Final Safety Analysis Report ML24176A2622024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations ML24280A0012024-06-20020 June 2024 Update to the Final Safety Analysis Report (Redacted Version) ML24281A2072024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 (Redacted Version) 05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications ML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML24165A1292024-06-0505 June 2024 ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023 ML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities ML24110A0562024-05-21021 May 2024 Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML24142A0952024-05-20020 May 2024 End of Cycle 22 Steam Generator Tube Inspection Report 05000423/LER-2023-006-01, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24141A2432024-05-20020 May 2024 Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15 IR 05000336/20240012024-05-14014 May 2024 Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation ML24123A2042024-05-0202 May 2024 Pre-Decisional Replay to EA-23-144 05000423/LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications IR 05000336/20244012024-04-30030 April 2024 Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only) ML24123A1222024-04-30030 April 2024 Inservice Inspection Program - Owners Activity Report, Refueling Outage 22 ML24116A0452024-04-25025 April 2024 Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only) ML24116A1742024-04-24024 April 2024 Annual Radiological Environmental Operating Report ML24114A2662024-04-24024 April 2024 Submittal of 2023 Annual Radioactive Effluent Release Report ML24103A0202024-04-22022 April 2024 Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits ML24106A2032024-04-15015 April 2024 2023 Annual Environmental Operating Report ML24088A3302024-04-0404 April 2024 Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel ML24093A2162024-04-0101 April 2024 Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits 2024-09-04
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24141A2432024-05-20020 May 2024 Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15 ML24093A2162024-04-0101 April 2024 Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits ML23361A0942023-12-21021 December 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies and Core Operating Limits Report . ML23248A2132023-08-30030 August 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature. ML23208A0922023-07-26026 July 2023 Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of Framatome ORFEO-GAIA and OORFE-NMGRID CHF Correlations in the Dominion Energy Vipre-D Computer Code Response ML23124A3642023-04-20020 April 2023 Response to Request for Additional Information for Spring 2022 Steam Generator Tube Inspection Report ML23096A2982023-04-0606 April 2023 Units 1 and 2 and Millstone Power Station, Units 2 and 3 - Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion ML22312A4432022-11-0707 November 2022 NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Fleet Response to RAI ML21259A0852021-09-15015 September 2021 North Ann, and Surry, Units 1 and 2, Millstone, Units 2 and 3, Request for Approval of Appendix E of Fleet Report DOM-NAF-2-A Qualification of the Framatome BWU-I CHF Correlation in the Vipre-D Computer Code Response to Request for Addition ML21209A7622021-07-26026 July 2021 Response to Request for Additional Information for Alternative Request V-01 - Proposed Request for Alternative Frequency to Supplemental Valve Positionn Verification Testing Requirements ML21153A4132021-06-0202 June 2021 Response to Request for Additional Information Regarding License Amendment Request for Measurement Uncertainty Recapture Power Uprate ML21147A4772021-05-27027 May 2021 NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors - Final Supplemental Response ML21140A2992021-05-20020 May 2021 Response to Request for Additional Information for Proposed License Amendment Request to Add an Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident ML21133A2852021-05-13013 May 2021 Stations, Units 1 & 2 and Millstone Power Station, Units 2 and 3 - Request for Approval of Appendix E Fleet Report DOM-NAF-2-A Qualification of the Framatome Bwui CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML21105A4332021-04-15015 April 2021 Final Supplemental Response to NRC Genetic Letter 2004-02 on Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accident at Pressurized-Water Reactors ML21105A4822021-04-15015 April 2021 Response to Request for Additional Information for Proposed License Amendment Request to Revise the Millstone, Unit 2 Technical Specification for Steam Generator Inspection Frequency ML21081A1362021-03-19019 March 2021 Response to Request for Additional Information for Alternative Request RR-05-06 - Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles ML20274A3462020-09-30030 September 2020 Response to Request for Additional Information for License Amendment Request to Revise Battery Survillance Requirements ML20261H5982020-09-17017 September 2020 Response to Request for Additional Information for License Amendment Request to Revise Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20252A1912020-09-0404 September 2020 Response to Request for Additional Information Regarding License Amendment Request for a One-Time Deferral of the Millstone Unit 3 Steam Generator Inspections ML20209A5362020-07-27027 July 2020 Response to Request for Additional Information Regarding Relief Request IR-3-33 for Limited Coverage Examinations Performed in the Second Period of the Third 10-Year Inspection Interval ML20079K4242020-03-19019 March 2020 Response to Request for Additional Information for License and Request to Revise TS 3.8.1.1, A.C Sources - Operating, to Support Maintenance and Replacement of the Millstone Unit 3 'A' Reserve Station Service Transformer and 345 Kv South Bu ML20076C8332020-03-16016 March 2020 Response to Request for Additional Information (E-mail Dated 3/16/2020) Alternative Request IR-4-03 for Use of Alternative Non-Code Methodology ML20048A0192020-02-11011 February 2020 Response to Request for Additional Information for License Amendment Request to Revise TS 3.8.1.1, A.C. Sources - Operating, to Support Maintenance and Replacement of a Reserve Station Service Transformer and 345 Kv South Bus. ML20042D9962020-02-10010 February 2020 Response to March 12, Request for Information Enclosure 2, Recommendation 2.1, Flooding Focused Evaluation/Integrated Assessment Submittal ML19284A3972019-10-0303 October 2019 Response to NRC Request for Additional Information on License Amendment Request to Adopt 10 CFR 50.69 ML19249B7672019-08-29029 August 2019 Enclosure 1 - Millstone, Units 2 and 3 and ISFSI; North Anna, Units 1 and 2 and ISFSI; and Surry, Units 1 and 2 and ISFSI - Response to EAL Scheme Change RAIs ML19092A3322019-03-27027 March 2019 Response to Request for Additional Information for Proposed Technical Specification Changes for Spent Fuel Pool Storage and New Fuel Storage ML19011A1112018-12-18018 December 2018 Supplement to the Flooding Hazard Reevaluation Report in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding ML18340A0282018-11-29029 November 2018 Response to Request for Additional Information for Proposed Technical Specifications Changes for Spent Fuel Pool Storage and New Fuel Storage ML18302A1202018-10-22022 October 2018 Response to Request for Additional Information for License Amendment Request to Revise the Technical Specification for Control Building Ventilation Inlet Instrumentation ML18235A3212018-08-17017 August 2018 Response to Request for Additional Information for Proposed Alternative Request P-06 for 'C' Charging Pump ML18225A0662018-08-0606 August 2018 Response to Request for Additional Information for Alternative Requests Associated with the In-Service Testing Program for Pumps, Valves, and Snubbers Fifth and Fourth 10-Year Interval Updates ML18205A1762018-07-19019 July 2018 Response to Request for Additional Information for Alternative Requests Associated with the In-Service Testing Program for Pumps, Valves, and Snubbers Fifth and Fourth 10-Year Interval Updates for Units 2 and 3 ML18170A0932018-06-14014 June 2018 Response to Request for Additional Information Regarding License Amendment Request to Revise Integrated Leak Rate Test (Type a) and Type C Test Intervals ML18151A4672018-05-24024 May 2018 Response to Request for Additional Information Regarding License Amendment Request to Revise Integrated Leak Rate Test (Typed a) and Type C Test Intervals ML17338A0572017-11-22022 November 2017 Response to Request for Supplemental Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools ML17300A2202017-10-24024 October 2017 Response to Information Need Request Regarding Mitigating Strategies Assessment (MSA) Report for Flooding ML17053A1062017-02-16016 February 2017 Response to Request for Additional Information Regarding Proposed Alternative Requests RR-04-24 and IR-3-30 for Elimination of the Reactor Pressure Vessel Threads in Flange Examination ML17038A0052017-01-31031 January 2017 Response to RAI Regarding End of Cycle 23 and End of Cycle 17 Steam Generator Tube Inspection Reports, CAC MF8507 & MF8506 ML16365A0362016-12-22022 December 2016 Response to March 12, 2012 Information Request High Frequency Sensitive Equipment Functional Confirmation for Recommendation 2.1 ML16365A0322016-12-21021 December 2016 Response to March 12, 2012 Information Request, Spent Fuel Pool Seismic Evaluation for Recommendation 2.1 ML16321A4542016-11-10010 November 2016 Connecticut and Virginia Electric & Power Company Response to Request for Additional Information Revision 22 of Quality Assurance Program Description Topical Report ML16312A0642016-11-0101 November 2016 Units 1 & 2, Surry, Units 1 & 2, Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools ML16294A2702016-10-18018 October 2016 Response to Request for Additional Information for License Amendment Request Regarding Realistic Large Break Loss of Coolant Accident Analysis - RAI Questions 1 Through 3 ML16291A5082016-10-12012 October 2016 Response to Follow Up Request to Revise ECCS TS 3/4.5.2 and FSAR Chapter 14 to Remove Charging ML16202A0402016-07-14014 July 2016 Response to Request for Additional Information Regarding Spent Fuel Pool Heat Load Analysis License Amendment Request ML16188A1962016-06-30030 June 2016 NRC Regulatory Issue Summary 2016-09 Preparation and Scheduling of Operator Licensing Examinations ML16182A0372016-06-27027 June 2016 Response to Request for Additional Information for License Amendment Request to Revise ECCS TS 3/4.5.2 and FSAR Chapter 14 to Remove Charging 2024-09-16
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Text
Dominion En e rgy Nu c lea r Connect ic ut, Inc.
5000 Dom in ion Bo ul evar d, Glen Alle n. VA 23 060 Dominion Energy.com August 30, 2023
U. S. Nuclear Regulatory Commission Serial No.22-361 B Attention: Document Control Desk NRA/SS: RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49
DOMINION ENERGY NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT REQUEST TO REVISE THE APPLICABILITY TERM FOR REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN PRESSURE TEMPERATURE LIMITATIONS FIGURES
By letter dated January 13, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23013A224), Dominion Energy Nuclear Connecticut, Inc. (DENC) submitted a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change would revise MPS3 TS 3.4.9.1, "Reactor Coolant System Pressure/Temperature Limits,"
to reflect that Figures 3.4-2 and 3.4-3 (Heatup and Cooldown Limitations, respectively) would be applicable up to 54 effective full power years (EFPY).
In an email dated June 29, 2023, the NRC issued a draft request for additional information (RAI) related to the proposed LAR. On July 18, 2023, the NRC staff conducted a conference call with DENC staff to clarify the request. In an email dated July 18, 2023, the NRC transmitted the final version of the RAI (ADAMS Accession No. ML23199A283).
DENC agreed to respond to the RAI within 45 days of issuance, or no later than September 1, 2023.
provides DENC's response to the RAI. Attachment 2 provides a Figure and Tables which support DENC's responses to the RAI questions.
Serial No.22-361 B Docket No. 50-423 Page 2 of 3
If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.
Sincerely,
f Douglas rence Senior V resident - Nuclear Operations and Fleet Performance
COMMONWEAL TH OF VIRGINIA
COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Douglas C. Lawrence who is Senior Vice President - Nuclear Operations and Fleet Performance of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this 3D4"'day of ~YeY~-\\- I 2023.
My Commission Expires: °JO\\'\\\\ACAY\\.l '3\\ 1 2C2LJ
Kathryn Hill Barret Notary Public ~~- ]J.:-1f3P7\\U[
Commonwealth of Virginia Notary Public Reg.No. 7905256 My Commission Expires January 31, 2024 Attachments:
- 1. Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures
- 2. Figure and Tables to Support Responses to Request for Additional Information
Commitments made in this letter: None Serial No.22-361 B Docket No. 50-423 Page 3 of 3
cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road, Suite 105 King of Prussia, PA 19406-1415
Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 9 E3 11555 Rockville Pike Rockville, MD 20852-2738
NRC Senior Resident Inspector Millstone Power Station
Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Serial No. 22-361B Docket No. 50-423
Attachment 1
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT REQUEST TO REVISE THE APPLICABILITY TERM FOR REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMITATIONS FIGURES
Dominion Energy Nuclear Connecticut, Inc.
Millstone Power Station Unit 3 Serial No. 22-361B Docket No. 50-423 Attachment 1, Page 1 of 4
By letter dated January 13, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23013A224), Dominion Energy Nuclear Connecticut, Inc. (DENC) submitted a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change would revise MPS3 TS 3.4.9.1, Reactor Coolant System Pressure/Temperature Limits, to reflect that Figures 3.4-2 and 3.4-3 (Heatup and Cooldown Limitations, respectively) would be applicable up to 54 effective full power years (EFPY).
In an email dated June 29, 2023, the NRC issued a draft request for additional information (RAI) related to the proposed LAR. On July 18, 2023, the NRC staff conducted a conference call with DENC staff to clarify the request. In an email dated July 18, 2023, the NRC transmitted the final version of the RAI (ADAMS Accession No. ML23199A283).
DENC agreed to respond to the RAI within 45 days of issuance, or no later than September 1, 2023.
This attachment provides DENCs response to the RAI.
RAI 1
Table 2 in Attachment 3, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, of the LAR presents Docketed Beltline Chemistries and Initial RTNDT Values from Millstone LRA, and identifies the initial RTNDT value for intermediate shell plate B9805-1 as 80.0°F. However, in Tables 1, 6, 7 and 8 of Attachment 3, as well as Table 4.2-2 of the Millstone Unit 3 License Renewal Application (ML040260103), the initial RTNDT value for intermediate shell plate B9805-1 is shown as 60.0°F. Identify the correct initial RTNDT value for intermediate shell plate B9805-1. Explain the basis for this discrepancy.
DENC Response to RAI-1
DENC has reviewed Table 2 in Attachment 3 to the LAR submittal and has confirmed that the correct initial RTNDT value for intermediate shell plate B9805-1 is 60.0°F. The basis for the discrepancy was a resolution/clarity issue in the electronic file for the LAR submittal. A clarified version of this table from the LA R is provided as Table 2 in to this RAI response letter.
RAI 2
Figure 5, Millstone Unit 3 RPV Rollout Map, in Attachment 3 of the LAR identifies th e MPS3 reactor pressure vessel (RPV) shell plates. The nozzle shell plates are identified as Plates B9404-1, B9404-2, and B9404-3. By letter dated November 19, 2020 (ML20324A703), DENC requested a Measurement Uncertainty Recapture Power Uprate.
Table IV-2, Calculated Neutron Fluence Projections of RV Beltline and Extended Beltline Materials for MPS3 at 54 and 60 EFPY, in that document identifies the Nozzle Shell Plates as B9804-1, B9804-2, and B9804-3. Provide confirmation of the plate identifications for RPV Nozzle Shell Plates in MPS3.
Serial No. 22-361B Docket No. 50-423 Attachment 1, Page 2 of 4
DENC Response to RAI-2
DENC has reviewed Figure 5 in Attachment 3 to the LAR submittal and has confirmed that the correct identifiers for the MPS3 RPV Nozzle Shell Plates are B9804-1, B9804-2, and B9804-3. A corrected version of this figure from the LAR Attachment 3, Figure 5 is provided as Figure 5 in Attachment 2 to this RAI response letter.
RAI 3
On October 14, 2014, the NRC issued Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, (ML14149A165) whic h clarified that the beltline definition in 10 CFR 50, Appendix G is applicable to all RPV ferritic materials with projected neutron fluence values greater than 1x1017 neutrons/centimeter-squared (n/cm2) with energy greater than one million electron volts (E > 1 MeV), and this neutron fluence threshold remains applicable for the licensed operating period.
As noted in section 3.4 of Attachment 3, RPV nozzles with projected neutron fluence values less than the 4.28 x 1017 n/cm2 threshold defined in PWROG-15109-NP-A do not require further evaluation. Table 4 in Attachment 3 presents projected neutron fluence values at 54 and 60 EFPY for all beltline RPV materials. However, all beltline materials have not been included in Tables 1, 2, 5, 6, 7 and 8 in Attachment 3.
Describe how the P-T limit curves for MPS3 consider all ferritic pressure boundary components of the reactor vessel that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period. If the current P-T limit curves do not consider all applicable ferritic pressure boundary components of the RPV that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period, provide appropriately revised P-T limit curves for review. Provide all inputs and projected values for each beltline material. Similarly, for RTPTS, provide all inputs and projected values for each beltline material.
DENC Response to RAI-3
The P-T limit curves for MPS3 consider all ferritic pressure boundary components of the RPV that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period of 54 EFPY (also referred to as in-scope RPV ferritic pressure boundary materials). Specifically, the current TS P-T limits were developed using bounding 32 EFPY 1/4T Adjusted Reference Temperature (ART) and 3/4T ART values of 124.8 °F and 107.0 °F, respectively, and were associated with RPV intermediate shell plate B9805-1. The bounding 54 EFPY 1/4T and 3/4T ART values are 118.7 °F and 107.0 °F and are associated with RPV lower shell plate B9820-
- 2. On this basis, the existing TS P-T limit curves are concluded to be valid to 54 EFPY.
The inputs and projected values for each beltline material predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed Serial No. 22-361B Docket No. 50-423 Attachment 1, Page 3 of 4
operating period (54 EFPY) are presented in Tables 7, 8, 10, and 11 in Attachment 2 to this RAI response letter.
Similar to the ART data, the inputs and projected values for the RTPTS values for all the in-scope RPV ferritic pressure boundary materials are included in the Tables 9 and 12 in to this RAI response letter.
RAI 4
As required by 10 CFR 50.61, To verify that RTNDT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results.
The licensee stated the MPS3 surveillance weld material (heat 4P6052) is additionally represented in surveillance capsules at Seabrook and at a non-U.S. plant. The test results and information for the Seabrook reactor vessel material surveillance program are publicly available, including:
Yankee Atomic Electric Company, Analysis of Seabrook Station Unit 1 Reactor Vessel Surveillance Capsule U, June 26, 1992 (ML20099J203)
Duke Engineering and Services, Report No. DES-NFOA-98-01 Analysis of Seabrook Stations Unit 1, Reactor Vessel Surveillance Capsules U & Y, May 5, 1998 (ML20248L409)
WCAP-16526-NP, Rev. 0, Analysis of Capsule V from FPL Energy - Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program, March 31, 2006 (ML061030088)
WCAP-18607-NP, Analysis of Capsule X from the NextEra Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program, September 30, 2021 (ML21277A388)
Provide the results of an evaluation of the applicability, including credibility, of the surveillance program results pertaining to weld heat 4P6052. Describe how these results were taken into consideration in evaluating the RTNDT of weld heat 4P6052 at MPS3.
DENC Response to RAI-4
The Seabrook Unit 1 surveillance capsule program results for the shared weld material heat 4P6052 have been evaluated and determined to be applicable to the evaluation of MPS3 weld heat 4P6052. Availability of the surveillance capsule data for weld material heat 4P6052 at the non-U.S. plant (Ascó Unit 2) was also investigated, but the data was found to be proprietary and unavailable for use.
The combined surveillance data from the Seabrook Unit 1 surveillance capsule program documented in WCAP-18607-NP and the MPS3 surveillance capsule program Serial No. 22-361B Docket No. 50-423 Attachment 1, Page 4 of 4
documented in WCAP-16629-NP were determined to be credible. Referring to Table 4 in to this RAI response letter, the unadjusted values for RTNDT (shown in paratheses) for MPS3 and for Seabrook 1 were adjusted based on considerations from Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity. Specifically, the values were adjusted based on an Irradiation Temperature difference from MPS3 (558°F) and Seabrook 1 (559°F) and by the ratio of the chemistry factor of the vessel weld (30.7°F) to the chemistry factor of the surveillance weld (28.9°F for MPS3 and 30.7°F for Seabrook 1). Negative values of RTNDT were set to zero. The MPS3 and Seabrook Unit 1 surveillance weld data were used to develop a fitted Chemistry Factor (CF) value and RTNDT trend curve for weld heat number 4P6052, and the RTNDT data points of weld heat 4P6052 fell within the 28°F scatter band specified in Regulatory Guide (RG) 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2 for surveillance weld materials. Therefore, the surveillance data for weld heat number 4P6052 are deemed credible.
DENC concludes that the limiting plate material RTNDT and RTPTS for MPS3 (B9820-2) bound the RTNDT and RTPTS of weld material 4P6052 considering either position 1.1 or position 2.1 of RG 1.99, Revision 2. The results which support this conclusion are shown in Tables 7, 8, and 9 of Attachment 2 to this RAI response letter.
Serial No. 22-361B Docket No. 50-423
Attachment 2
FIGURE AND TABLES TO SUPPORT RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION
Dominion Energy Nuclear Connecticut, Inc.
Millstone Power Station Unit 3 Serial No. 22-361B Docket No. 50-423 Attachment 2, Page 1 of 8
Note: The Reference numbers in table/figure footnotes correlate to the Reference numbers in LAR Attachment 3, with the exception of new Reference [22].
Table 2: Docketed Beltline Chemistries and Initial RTNDT Values from Millstone LRA Ta ble 4.2-2 Millstone Unit 3
Chemi c al Ma teria l De sc ription Compo sition In itial C h emis try Inner Mar gin t.RTPTs R Tprs
Re a c to r RT NDT Fac to r Su rfac e *F *F *F Ve sse l *F *F F luence m 2 Beltline Ma tl. Hea t Ty p e Cu Ni E19 n / c R egio n ldent. Number Wt. % W t.%
L o ca tio n
Intermed iate B9805 -1 C4039-2 SA-533 B 0.0 5 0.64 60.0 1 31.02 3.31 34.0 40.7 134. 7 She ll C l. 1
Intermed iate B9805 -2 C4068-1 SA-533B 0.05 0.64 6.2 1 31.02 3.31 34.0 40.7 80.9 She ll C l. 1
Intermed ia te B9805-3 C4028-1 SA-533B 0.05 0.65 -3.3 1 31.0 2 3.3 1 34.0 40.7 71.4 She ll C l. 1
Lo w er She ll B9 820 - 1 B8961 - 1 SA-533B 0.08 0.63 7.0 1 51.02 3.31 34.0 67.0 108. 0 C l. 1
Lo w er She ll B9820 -2 D1242-2 SA-533B 0.07 0.60 38.8 44.02 3.31 34.0 57.8 130.6 C l.1
Lower Sh ell B9820 -3 D 1242 - 1 SA-533 B 0.06 0.6 1 18.6 37.02 3.31 34.0 48.6 101.2 C l.1
A ll We lds 4P6052 Linde 0 09 1 0.05 0.05 -50.0 1 3 1.72 3.31 56.0 47.7 47.7
- 1. Measu re V alue
- 2. Regu la tory Guide 1.99, Re v ision 2, Pos itio n 1 Figure 5: Millstone Unit 3 RPV Rollout Map
0 Outlet hie! rile! Outlet OIAJel hie! Inlet Outlet
Plate 89804 -1 P la te 898 04 -2 Plate 89804 -3 40 103-121 (Heat No. 4P60521 80...... 'i 'i 'i Plate !!! Pla te !!! Plate !!! z 99805 -1 ! z 89805 - 2 ~ z 89805-3 ! ? ? ?
(Heat No. C4039-2)...... (Heat No. C4068*1)...... (Heat No. C4028 -1)..,,..,..
(/).. "D... "D.... "D.... n O>
w 120 0 0 0 J: _g; "' "' !::! !::!
(.) ------ ------------ ------------- *
~ 101-171 (Heat No. 4P6052)
....J 160 J: x %.... ;
i Plate ~ !!! !!! ~ ate ~ z Plate ! z B9820-3 ? B9820-1.!. ? B9820-2 ? ~ z Pl 200 (Heat No. B8961-1).... ~...... r;: "D (Heat No. 0 1242-2) "D ( Heat No. 01 242-1).... "D O>.,.. n..
0 ------- 0 - - - - - -* 0
- - - - *- _;-..,. _ - u,-- -!::! !::! - * -
240 257 0 90 180 270 360 AZIMUTHAL DEGREES Note :
( 1 ) P late iden tifiers we re co rrected to al ign with C MT R data [9]
Serial No. 22-361B Docket No. 50-423 Attachment 2, Page 2 of 8
Table 4: Surveillance Data for Surveillance Weld Heat #4P6052 for Calculation of Chemistry Factor
Plant Capsule Capsu le Fluence <1> (x 10 19 FF <2>.6.RT NDT (3) FF*.6.RT NDT FF2 n/cm 2, E > 1.0 MeV) ( OF) ( OF) u 0.400 0.746 32.5 24.2 0.557 (30.6 )
Millstone X 1.98 1. 186 0 0.0 1.407 3 (-6.3) w 3. 16 1.303 0 0.0 1.698 (-1.3) u 0.311 0.680 26.4 18.(25.4 ) 0 0.462 3 1.073 (24.25.3 Se abrook y 1.3 ) 27. 1 1. 151 1 V 2.66 1.262 42.7 53.9 1.593 (41. 7)
X 6.03 1.437 34.8 50.(33.8 ) 0 2.065 SUM : 173.2 8.932 CF 4P6052 = L ( FF *b.R T NDT ) + L ( F F 2) = (1 73.2 ) + (8.932 ) = 19.4°F Not es:
(1) M illsto n e 3 flu e nc es ta ke n fro m [12]. Sea brook 1 fluence s t a ke n from [22).
(2) FF= f 02&--0I O!oef.J)
(3) U n a djus ted v a lu es fo r LI.R T No r (shown in pa re nth e ses) w e re ta ke n fro m [12] for M illston e 3 a n d (22] fo r Sea broo k 1. V a lues w ere a djusted b ased on a n Irra dia tion T em pe ra ture d iffe re nce from M illstone 3 (558 °F ) a nd Sea brook 1 (55 9 °F) a nd the n w e re a djuste d b y the ratio o f the c he mistry facto r o f the vessel we ld (30. 7°F) to t he c he mistry fa c to r o f the su rve illa n ce w eld ( 28.9°F for M illsto n e 3 a n d 30.7° F for Sea brook 1). Nega tive v a lues o f LI. RT No r w ere set to zero.
New Reference [22]:
Westinghouse Report, WCAP-18607-NP, Revision 0, Analysis of Capsule X from the NextEra Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program, March 2021, SI File Number 1901402.204.