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Category:Letter
MONTHYEARML24030A7522024-01-30030 January 2024 Technical Specification Bases Pages IR 05000336/20234022024-01-30030 January 2024 Security Baseline Inspection Report 05000336/2023402 and 05000423/2023402 (Cover Letter Only) 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 IR 05000336/20234402024-01-11011 January 2024 Special Inspection Report 05000336/2023440 and 05000423/2023440 (Cover Letter Only) ML24004A1052024-01-0404 January 2024 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000336/2024010 & 05000423/2024010 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 . ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23361A0312023-12-20020 December 2023 Intent to Pursue Subsequent License Renewal ML23352A0202023-12-18018 December 2023 Senior Reactor and Reactor Operator Initial License Examinations ML23334A2242023-11-30030 November 2023 Request for Exemption from Enhanced Weapons Firearms Background Checks, and Security Event Notifications Implementation ML23324A4222023-11-20020 November 2023 Reactor Vessel Internals Inspections Aging Management Program Submittal Related to License Renewal Commitment 13 ML23324A4302023-11-20020 November 2023 Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment 15 ML23317A2702023-11-13013 November 2023 Core Operating Limits Report, Cycle 23 IR 05000336/20230032023-11-0606 November 2023 Integrated Inspection Report 05000336/2023003 and 05000423/2023003 ML23298A1652023-10-26026 October 2023 Requalification Program Inspection IR 05000336/20234202023-10-0404 October 2023 Security Inspection Report 05000336/2023420 and 05000423/2023420 ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23226A0052023-09-26026 September 2023 Issuance of Amendment No. 287 Supplement to Spent Fuel Pool Criticality Safety Analysis IR 05000245/20230012023-09-19019 September 2023 Safstor Inspection Report 05000245/2023001 IR 05000336/20230102023-09-0808 September 2023 Commercial Grade Dedication Report 05000336/2023010 and 05000423/2023010 IR 05000336/20230052023-08-31031 August 2023 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Report 05000336/2023005 and 05000423/2023005) 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. ML23242A0142023-08-30030 August 2023 Operator Licensing Examination Approval ML23223A0552023-08-18018 August 2023 Request for Withholding Information from Public Disclosure for License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies, and COLR Related to Framatome Gaia Fuel ML23223A0482023-08-18018 August 2023 Request for Withholding Information from Public Disclosure for License Amendment Request to Use Framatome Small Break and Realistic Large Break LOCA Evaluation Methodologies for Establishing COLR Limits IR 05000336/20230022023-08-0909 August 2023 Integrated Inspection Report 05000336/2023002 and 05000423/2023002 ML23188A0432023-07-31031 July 2023 Authorization and Safety Evaluation for Alternative Request No. IR-04-11 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 ML23188A0202023-07-26026 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23207A1102023-07-26026 July 2023 NRC Regulatory Issues Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations IR 05000336/20234012023-07-17017 July 2023 Material Control and Accounting Program Inspection Report 05000336/2023401 and 05000423/2023401 - (Cover Letter Only) ML23175A0052023-07-12012 July 2023 Alternative Request P-07 for Pump Periodic Verification Testing Program for Containment Recirculation Spray System Pumps ML23193A8562023-06-28028 June 2023 Submittal of Updates to the Final Safety Analysis Reports ML23178A1682023-06-26026 June 2023 2022 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML23153A1732023-06-16016 June 2023 Correction to Amendment Nos. 346 & 286 Millstone, 294 & 277 North Anna, 311 & 311 Surry, and 225 Summer to Revise Technical Specifications to Adopt TSTF-554,Rev Reactor Coolant Leakage Requirement ML23151A0742023-06-12012 June 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML23159A2202023-06-0808 June 2023 Associated Independent Spent Fuel Storage Installation Revision to Emergence Plan - Report of Changes IR 07200047/20234012023-06-0808 June 2023 NRC Independent Spent Fuel Storage Installation Security Inspection Report No. 07200047/2023401 2024-01-04
[Table view] Category:Licensee 30-Day Written Event Report
MONTHYEARML18075A0252018-03-0808 March 2018 Revision to 30-Day Special Report for an RCS Pressure Transient ML17319A0892017-11-0707 November 2017 30-Day Special Report for an RCS Pressure Transient ML15345A2432015-12-0404 December 2015 Connecticut, Inc. Millstone Power Station Unit 1 Decommissioning Trust Fund Disbursement Thirty-Day Written Notification ML14139A0122014-05-0808 May 2014 30-Day Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML13260A2512013-09-0909 September 2013 Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML12340A0102012-11-29029 November 2012 30-Day Report of Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 Cfr 50.46 ML0925803822009-09-0404 September 2009 Day Special Report for Area Temperature Monitoring - Containment Area CS-03 (Pressurizer Cubicle) ML0632605692006-11-20020 November 2006 Day Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 2018-03-08
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Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, Virginla 2.3060 Wet, Address: www.dom.com November 20, 2006 U.S. Nuclear Regulatory Commission Serial No.06-981 Attention: Document Control Desk MPS LicANDB RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION UNIT 3 30-DAY REPORT OF EMERGENCY CORE COOLING SYSTEM (ECCS) MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 In accordance with 10 CFR 50.46(a)(3)(ii), Dominion Nuclear Connecticut, Inc. (DNC) hereby submits information regarding changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Model for the BASH Large Break Loss of Coolant Accident (LBLOCA) analysis for Millstone Power Station Unit 3 (MPS3) and its application in existing analyses. provides a report describing changes associated with the Westing house BASH LBLOCA ECCS Evaluation Model for MPS3.
Information regarding the effect of the ECCS evaluation model changes upon the reported LBLOCA analysis of record (AOR) result for MPS3 is provided in Attachment 2.
To summarize the information in Attachment 2, the calculated peak cladding temperature (PCT) for the LBLOCA analysis is increased by 74°F to a new value of 2048°F. This result represents a significant change in PCT, as defined in 10 CFR 50.46(a)(3)(i).
10 CFR 50.46(a)(3)(ii) requires the licensee to provide a report within 30 days, which includes a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46. Dominion has reviewed the information provided by Westinghouse and determined that the adjusted LBLOCA PCT value and the manner in which it was derived continue to conform to the requirements of 10 CFR 50.46. As such, Dominion considers the schedular requirements of 10 CFR 50.46(a)(3)(ii) to be satisfied with the submission of this notification. Dominion routinely tracks adjustments to the Small Break Loss of Coolant Accident (SBLOCA) and LBLOCA calculated PCT values to ensure that reasonable margins to the acceptance value set by 10 CFR 50.46 are maintained.
This information satisfies the 30 day reporting requirements of 10 CFR 50.46(a)(3)(ii).
Serial Number 06-981 Docket No. 50-423 bc Page 2 of 2 If you have any further questions regarding this submittal, please contact Mr. David W.
Dodson at (860) 447-1791 , Extension 2346.
Very truly yours,
~ e r a l T.
d Bischof w Vice President - Nuclear Engineering Commitments made in this letter: None Attachments: (2)
- 1) Report of Changes in Westinghouse BASH Large Break LOCA ECCS Evaluation Model - Millstone Power Station Unit 3.
- 2) Reporting of 10 CFR 50.46 Margin Utilization - Westinghouse BASH Large Break LOCA ECCS Evaluation Model - Millstone Power Station Unit 3.
Serial Number 06-981 Docket No. 50-423 ATTACHMENT 1 REPORT OF CHANGES IN WESTINGHOUSE BASH LARGE BREAK LOCA ECCS EVALUATION MODEL DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
Serial Number 06-981 Docket No. 50-423 Attachment 1, Page 1 of 1 REPORT OF CHANGES IN WESTINGHOUSE BASH LARGE BREAK LOCA ECCS EVALUATION MODEL MILLSTONE POWER STATION UNIT 3 Identification of ECCS Evaluation Model Chancres The current large break loss of coolant accident (LBLOCA) analysis for Millstone Power Station Unit 3 (MPS3) was performed using the Westinghouse BASH LBLOCA Evaluation Model. Westinghouse identified the change described below and provided the results of an assessment to determine the impact on peak cladding temperature (PCT).
Change: BASH Minimum and Maximum Time Step Sizes Westinghouse reviewed recent BASH LBLOCA Evaluation Model sensitivity calculations. This review led to a recommendation to reduce the minimum and maximum time step sizes in the BASH LBLOCA Evaluation Model during the reflood phase of the accident. Westinghouse performed sensitivity calculations using the BASH LBLOCA Evaluation Model with reduced minimum and maximum time step sizes. The results showed a decrease in the integral flooding rate late in the reflood phase of the accident that resulted in a 44°F increase in PCT (APCT = 44°F).
Conclusion Dominion has performed an evaluation of PCT for comparison to 10 CFR 50.46 requirements. The Analysis of Record (AOR) PCT is 1974°F. Considering the current PCT change as well as all previously reported changes, the corrected LBLOCA PCT is 2048°F. The MPS3 LBLOCA results have sufficient margin to the 2200°F limit specified in 10 CFR 50.46(b)(1). The PCT assessments for 10 CFR 50.46(a)(3)(i) accumulation include the current assessment (APCT = 44°F) and a previous assessment resulting from a rebaseline of the AOR (APCT = 30°F). The 10 CFR 50.46(a)(3)(i) accumulation of APCT = 74°F is greater than the 50°F limit for reporting; hence, the changes are significant and submittal of this 30 day report to the NRC is required.
Serial Number 06-981 Docket No. 50-423 ATTACHMENT 2 REPORTING OF 10 CFR 50.46 MARGIN UTILIZATION WESTINGHOUSE BASH LARGE BREAK LOCA ECCS EVALUATION MODEL DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
Serial Number 06-981 Docket No. 50-423 Attachment 2, Page 1 of 1
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10 CFR 50.46 Margin Utilization Large Break LOCA Plant Name: Millstone Power Station Unit 3 Utilitv Name:
I Dominion Nuclear Connecticut, Inc.
Analvsis Information EM: BASH Limiting Break Size: Cd=0.6 Analysis Date: 08/90 Vendor: Westinghouse FQ: 2.6 FAH: 1.7 Fuel: RFNR FA-2 SGTP (%): 10 Notes: All V5H assemblies have been removed from the core; the current fuel type is RFNRFA-2.
Clad T e m (~O F )
LICENSING BASIS Analysis of Record PCT 1974 PCT ASSESSMENTS (Delta PCT)
A. Prior ECCS Model Assessments
- 1. None B. Planned Plant Modification Evaluations
- 1. CHG/SI Alternate MiniFlow C. 2006 ECCS Model Assessments
- 1. BASH Minimum and Maximum Time Step Sizes 44 D. Other
- 1. Rebaseline of AOR LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 2048