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Category:Letter
MONTHYEARML24032A1112024-02-0101 February 2024 Owner'S Activity Report for North Anna, Unit 2, Refueling Outage N2R29 - First Period of the Fifth ISI Interval ML24032A4662024-01-16016 January 2024 Response to Comments on Draft Vpdes Permit No. VA0052451 ML24017A0802024-01-11011 January 2024 Notification of Licensed Operator Initial Examination 05000338/2024301 and 05000339/2024301 ML24003A8372024-01-0909 January 2024 Letter to Elizabeth Toombs, Thpo, Cherokee Nation, Notice of Availability of the Draft Site-Specific EIS for North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8412024-01-0909 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8502024-01-0909 January 2024 Letter to Wenonah Haire, Thpo, Catawba Indian Nation, Notice of Availability of the Draft Site-Specific EIS for North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML23332A7272024-01-0909 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8382024-01-0909 January 2024 Letter to G. Anne Richardson, Chief, Rappahannock Tribe, Inc., Notice of Availability of Draft Site-Specific Environmental Impact Statement for North Anna Power Station Units 1 and 2, Subsequent License Renewal ML24003A8482024-01-0909 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8392024-01-0909 January 2024 Letter to Gerald Stewart, Chief, Chickahominy Indian Tribe, Eastern Division Re Re Notice of Availability of the Draft Site-Specific Environmental Impact Statement for North Anna Power Station Subsequent License Renewal ML24003A8472024-01-0909 January 2024 Letter to Tom Jonathan, Chief, Tuscarora Nation, Notice of Availability of the Draft Site-Specific EIS for North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8492024-01-0909 January 2024 Letter to W. Frank Adams, Chief, Upper Mattaponi Tribe Re Re Notice of Availability of the Draft Site-Specific Environmental Impact Statement for North Anna Power Station Subsequent License Renewal ML24003A8462024-01-0909 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8352024-01-0909 January 2024 Letter to Acee Watt, Thpo, United Keetoowah Band of Cherokee Indians in Oklahoma Re Notice of Availability of the Draft Site-Specific Environmental Impact Statement for North Anna Power Station Subsequent License Renewal ML24003A8422024-01-0909 January 2024 Letter to Paul Barton, Thpo, Eastern Shawnee Tribe of Oklahoma, Notice of Availability of Draft Site-Specific Environmental Impact Statement for North Anna Power Station Units 1 and 2, Subsequent License Renewal ML24003A8432024-01-0909 January 2024 Letter to Robert Gray, Chief, Pamunkey Indian Tribe, Notice of Availability of the Draft Site-Specific EIS for North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8362024-01-0909 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML24003A8442024-01-0909 January 2024 Letter to Russell Townsend, Thpo, Eastern Band of Cherokee Indians ML24003A8402024-01-0909 January 2024 Letter to Katelyn Lucas, Thpo, Delaware Nation, Oklahoma, Notice of Availability of Draft Site-Specific Environmental Impact Statement for North Anna Power Station Units 1 and 2, Subsequent License Renewal ML24003A8452024-01-0909 January 2024 Letter to Stephen Adkins, Chief, Chickahominy Indian Tribe, Notice of Availability of Draft Site-Specific Environmental Impact Statement for North Anna Power Station Units 1 and 2, Subsequent License Renewal ML23332A1852024-01-0404 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML23332A0872024-01-0404 January 2024 Notice of Availability of the Draft Site-Specific Environmental Impact Statement for the North Anna Power Station, Units 1 and 2, Subsequent License Renewal ML23360A6352024-01-0404 January 2024 Letter to Dominion Availability for North Anna Units 1 and 2 Draft Site Specific Environmental Impact Statement ML23332A1472023-12-21021 December 2023 Issuance of Environmental Scoping Summary Report Associated with the Staff'S Review of the North Anna Power Station, Units 1 and 2, Subsequent License Renewal Application 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) ML23346A0972023-12-11011 December 2023 License Amendment Request to Revise the Emergency Plan Relocation of the Technical Support Center - Response to NRC Request for Additional Information ML23346A1022023-12-11011 December 2023 2024 North Anna Power Station Brq Inspection Notification Letter ML24012A0942023-12-0707 December 2023 GL Report from Dominion Virginia Power Regarding Transfer of Devices ML24012A0892023-12-0707 December 2023 GL Registration from Dominion Virginia Power ML23334A2432023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23362A0602023-11-29029 November 2023 ISFSI, Submittal of Cask Registration for Spent Fuel Storage IR 05000339/20234022023-11-27027 November 2023 Security Baseline Inspection Report 05000338 and 05000339/2023402 IR 05000338/20230032023-11-13013 November 2023 Integrated Inspection Report 05000338/2023003 and 05000339/2023003 ML23311A2042023-10-0909 October 2023 Annual Submittal of Technical Specification Bases Changes Pursuant to Technical Specification 5.5.13.d IR 05000338/20234032023-10-0505 October 2023 Reissue - North Anna Power Station - Cyber Security Inspection Report 05000338/2023403 and 05000339/2023403 (Cover Letter) ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23275A0992023-09-28028 September 2023 Subsequent License Renewal Application Third 10 CFR 54.21(b) Annual Amendment IR 05000338/20230112023-09-26026 September 2023 Quadrennial Focused Engineering Inspection Commercial Grade Dedication Report 05000338/2023011 and 05000339/2023011 ML23242A3482023-09-11011 September 2023 Cyber Security Inspection Report 05000338/2023403 and 05000339/2023403 - Cover ML23254A0242023-09-0606 September 2023 Paragon Energy Solutions LLC, Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23249A2592023-09-0606 September 2023 Core Operating Limits Report Cycle 30, Pattern Ukr, Revision 0 ML23214A2022023-09-0505 September 2023 Staff Assessment of Updated Seismic Hazards Following the NRC Process for the Ongoing Assessment of Natural Hazards Information 2024-02-01
[Table view] Category:Report
MONTHYEARML24032A1112024-02-0101 February 2024 Owner'S Activity Report for North Anna, Unit 2, Refueling Outage N2R29 - First Period of the Fifth ISI Interval ML24032A4662024-01-16016 January 2024 Response to Comments on Draft Vpdes Permit No. VA0052451 ML23214A1942023-09-0505 September 2023 Staff Assessment of Updated Seismic Hazards Following the NRC Process for the Ongoing Assessment of Natural Hazards Information - Report ML23103A2282023-04-12012 April 2023 Stations Units 1 and 2; Millstone Power Station Units 2 and 3, DOM-NAF-2-P/NP-A, Revision 0.4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML22353A6202022-12-19019 December 2022 Request for Approval of Appendix F Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code ML22255A0102022-06-28028 June 2022 Owner'S Activity Report ML22263A2852022-05-23023 May 2022 Alert and Notification System Evaluation Report_Ans Evaluation_Redacted Page 1-65 ML22119A1722022-04-13013 April 2022 Post-Accident Monitoring (PAM) Report ML21333A2842021-11-29029 November 2021 Requal Notification Letter ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML21036A0772021-02-23023 February 2021 Review of Steam Generator Tube Inspection Report for the Cycle 26 Spring 2019 Refueling Outage ML21042B3212021-02-11011 February 2021 Stations, Units 1 & 2; Millstone Power Station, Units 2 & 3 - Request for Approval of Fleet Report DOM-NAF-2 Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML20254A3472020-09-0808 September 2020 Supplement to Operator License Examination Comments ML20247J6162020-09-0303 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20246G7062020-08-24024 August 2020 Enclosure 4: Attachment 1 - PWROG-18005-NP, Revision 2, Determination of Unirradiated Rt and Upper-Shelf Values of the North Anna Units 1 and 2 Reactor Vessel Materials ML20203M1602020-07-20020 July 2020 VA Elec. & Power Co., Dominion Energy Nuclear Co. Inc., Dominion Energy Sc Inc., Millstone Power Station 2, N. Anna & Surry Power Stations 1 & 2, Virgil C. Summer Station 1, Updated Anchor Darling Double Disc Gate Valve Information & Status ML20149K6712020-05-31031 May 2020 PWROG-19047-NP, Revision 0, North Anna, Units 1 and 2, Reactor Vessels Low Upper-Shelf Fracture Toughness Equivalent Margin Analysis ML20140A2392020-05-18018 May 2020 ASME Section XI Inservicee Inspection Program Proposed Inservice Inspection Alternative N1-15-NDE-002 ML20246G7012020-03-31031 March 2020 Enclosure 4: Attachment 3 - WCAP-18364-NP, Rev. 1, North Anna Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal (SLR) ML20090B3972020-03-26026 March 2020 Revised License Renewal Commitment Pressurizer Surge Line Weld Inspection Frequency ML20246G7072020-01-31031 January 2020 Enclosure 4: Attachment 4 - WCAP-11164-NP, Rev. 2, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal ML19347A4212019-11-26026 November 2019 Owner'S Activity Report ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19106A3562019-04-23023 April 2019 Review of Steam Generator Tube Inspection Report for the Cycle 26 Spring 2018 Refueling Outage ML19011A1722019-01-0404 January 2019 Enclosure 3, Attachments 2C-3C - MPS3 EAL Technical Bases Document (Marked-Up) ML19011A1732019-01-0404 January 2019 Enclosure 4 - North Anna Power Station Units 1 & 2, EAL Scheme Revisions-Supporting Documents ML19011A1742019-01-0404 January 2019 Enclosure 5 - Surry Power Station, EAL Scheme Revisions-Supporting Documents ML20246G7092018-10-31031 October 2018 Enclosure 4: Attachment 2 - WCAP-18353-NP, Rev. 0, Reactor Internals Fluence Evaluation for a Westinghouse 3-Loop Plant with Two Units - Subsequent License Renewal ML18198A1192018-05-31031 May 2018 Attachment 5 to 18-233, ANP-3467NP, Rev. 0, North Anna Fuel-Vendor Independent Small Break LOCA Analysis Licensing Report ML17186A0842017-06-29029 June 2017 Flooding Focused Evaluation Summary ML16187A3232016-06-24024 June 2016 Submittal of Owner'S Activity Report (Form OAR-1), for Refueling Outage N2R24 ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report ML15238A8442015-09-25025 September 2015 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation ML15259A3422015-08-24024 August 2015 10 CFR 71.95 Report Evaluation Form; Submitted by Fenok, Erwin Resin Solutions, North Anna Power Stations, Et Al ML15232A8112015-08-24024 August 2015 Evaluation of Information Related to Commitments 6 and 8 from Confirmatory Action Letter No. NRR-2011-002 ML15238B5922015-08-17017 August 2015 Review of Commitment Action Completion Confirmation Action Letter Regarding Earthquake in 2011 ML15175A1902015-06-17017 June 2015 Owner'S Activity Report ML15057A2492015-04-20020 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 ML15058A0342015-02-23023 February 2015 Summary of Facility Changes, Tests and Experiments ML14133A0112014-05-0707 May 2014 March 12, 2012 Information Request Phase 2 Staffing Assessment Report ML14092A4162014-03-31031 March 2014 Response to March 12, 2012 Information Request Seismic Hazard and Screening Report (CEUS Sites)For Recommendation 2.1 ML14080A0022014-03-31031 March 2014 PNNL-22553, Final Assessment of Manual Ultrasonic Examinations Applied to Detect Flaws in Primary System Dissimilar Metal Welds at North Anna Power Station. ML14084A3272014-03-27027 March 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14084A2122014-03-27027 March 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14283A0462014-02-28028 February 2014 MRP-375, Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (EPRI 3002002441), Attachment 1 ML13338A4482014-01-29029 January 2014 Interim Staff Evaluation and Audit Report Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14006A1712014-01-23023 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for North Anna Power Station, Units 1 and 2, TAC Nos.: MF0998 and MF0999 2024-02-01
[Table view] Category:Miscellaneous
MONTHYEARML24032A1112024-02-0101 February 2024 Owner'S Activity Report for North Anna, Unit 2, Refueling Outage N2R29 - First Period of the Fifth ISI Interval ML22353A6202022-12-19019 December 2022 Request for Approval of Appendix F Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code ML22255A0102022-06-28028 June 2022 Owner'S Activity Report ML21333A2842021-11-29029 November 2021 Requal Notification Letter ML21036A0772021-02-23023 February 2021 Review of Steam Generator Tube Inspection Report for the Cycle 26 Spring 2019 Refueling Outage ML20254A3472020-09-0808 September 2020 Supplement to Operator License Examination Comments ML20247J6162020-09-0303 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML19347A4212019-11-26026 November 2019 Owner'S Activity Report ML17186A0842017-06-29029 June 2017 Flooding Focused Evaluation Summary ML16187A3232016-06-24024 June 2016 Submittal of Owner'S Activity Report (Form OAR-1), for Refueling Outage N2R24 ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report ML15238A8442015-09-25025 September 2015 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation ML15259A3422015-08-24024 August 2015 10 CFR 71.95 Report Evaluation Form; Submitted by Fenok, Erwin Resin Solutions, North Anna Power Stations, Et Al ML15232A8112015-08-24024 August 2015 Evaluation of Information Related to Commitments 6 and 8 from Confirmatory Action Letter No. NRR-2011-002 ML15238B5922015-08-17017 August 2015 Review of Commitment Action Completion Confirmation Action Letter Regarding Earthquake in 2011 ML15175A1902015-06-17017 June 2015 Owner'S Activity Report ML15057A2492015-04-20020 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 ML15058A0342015-02-23023 February 2015 Summary of Facility Changes, Tests and Experiments ML14084A3272014-03-27027 March 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14084A2122014-03-27027 March 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14024A6022014-01-10010 January 2014 Post Accident Monitoring (PAM) Report ML14015A3142014-01-0808 January 2014 Submittal of Owner'S Activity Report Refueling Outage N1R23 (Form OAR-1) ML12160A2682012-06-11011 June 2012 Review of 60-Day Response to Request for Information Regarding Recommendation 9.3, of the Near-Term Task Force Related to the Fukushima Daiichi Nuclear Power Plant Accident ML12060A3492012-02-16016 February 2012 Owner'S Activity Report for Refueling Outage N2R21 ML12039A1602012-01-25025 January 2012 Steam Generator Tube Inspection Report ML12039A1612012-01-25025 January 2012 Steam Generator Tube Inspection Report ML12005A0112011-12-0909 December 2011 Independent Spent Fuel Storage Installation Response to Earthquake ML1103107402011-01-25025 January 2011 Owner'S Activity Report ML1019304172010-05-0606 May 2010 Tritium Database Report ML0921908942009-08-0404 August 2009 Units 1 & 2, Millstone, Units 2 and 3 and Kewaunee - Approved Topical Report DOM-NAF-2, Revision 0.1-A ML0913206392009-05-12012 May 2009 Reactor Coolant Pressure Boundary Visual Inspections Proposed Alternative - N1-I3-NDE-024 and N2-I3-NDE-025 ML0912504812009-05-0404 May 2009 30-Day Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML0901304802009-01-0505 January 2009 Owner'S Activity Report for Refueling Outage ML0812702032008-04-25025 April 2008 Summary of Facility Changes, Tests and Experiments ML0807903432008-03-18018 March 2008 Response to Request for Additional Information Steam Generator Tube Inspection Report ML0802800722008-01-0808 January 2008 Update on North Anna Zirlo Characterization and LOCA Embrittlement Testing ML0732002342007-11-15015 November 2007 Steam Generator Tube Inspection Report ML0728205762007-10-0909 October 2007 Steam Generator Tube Inspection Report ML0719704722007-07-13013 July 2007 Owner'S Activity Reports, for Refueling Outage (N2R18) ML0705903272007-02-27027 February 2007 Fitness-For-Duty Program Semi-Annual Performance Data Report ML0623300712006-08-16016 August 2006 Fitness-for-Duty Program Semi-Annual Performance Data Report ML0609404202006-03-27027 March 2006 Summary of Facility Changes, Tests and Experiments ML0634101622006-02-27027 February 2006 Fitness-For-Duty Program Semi-Annual Performance Data Report ML0602600412006-01-25025 January 2006 Virginia Electric and Power Company North Anna Power Station Unit 2 - Owner'S Activity Reports ML0523501142005-08-22022 August 2005 Fitness-For-Duty Program Semi-Annual Performance Data Report ML0514603252005-05-25025 May 2005 .2 & 3, North Anna Power Station Units 1 & 2, Surry Power Station Units 1 & 2 - Nuclear Liability Insurance Endorsement ML0505403042005-02-22022 February 2005 Fitness-for-Duty Program Semi-Annual Performance Data Report Correction of Information ML0505300392005-02-21021 February 2005 Fitness-for-duty Program Semi-Annual Performance Data Report ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement ML0427101862004-09-24024 September 2004 Emergency Response Data System (ERDS) Database Revisions 2024-02-01
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 2, 2002 U. S. Nuclear Regulatory Commission Serial No. 02-167C Attention: Document Control Desk Docket Nos. 50-338 Washington, D.C. 20555 50-339 License Nos. NPF-4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 SMALL BREAK LOCA EVALUATION IN SUPPORT OF PROPOSED TECHNICAL SPECIFICATIONS CHANGES AND EXEMPTION REQUEST USE OF FRAMATOME ANP ADVANCED MARK-BW FUEL In a March 28, 2002 letter (Serial No.02-167), Virginia Electric and Power Company (Dominion) requested: 1) an amendment to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, and 2) associated exemptions from 10 CFR 50.44 and 10 CFR 50.46. The amendments and exemptions will permit North Anna Units 1 and 2 to use Framatome ANP Advanced Mark-BW fuel. This fuel design has been evaluated by Framatome and Dominion for compatibility with the resident Westinghouse fuel and for compliance with fuel design limits. The attachment to this letter documents the assessment of small break LOCA phenomena for the Advanced Mark-BW fuel. This information is provided in accordance with the proposed documentation for the transition effort as stated in our June 19, 2002 letter (Serial No.
02-305A). The remainder of the documentation required to establish compliance with the emergency core cooling system requirements of 10 CFR 50.46 for the transition to Advanced Mark-BW fuel will be submitted in separate correspondence as soon as possible.
As indicated in our June 19, 2002 letter, the approach taken relies upon application of the existing UFSAR analysis performed for the Westinghouse fuel. The attachment to this letter describes the assessment performed to determine the impact of fuel design effects upon small break LOCA phenomena. It is concluded that the existing small break LOCA analysis contained in the North Anna UFSAR is valid and provides a conservative representation of Advanced Mark-BW fuel.
As noted in previous correspondence, the initial reload batch of Advanced Mark-BW fuel is currently planned for North Anna Unit 1 Cycle 17, which is scheduled to begin operation in April 2003. We continue to request your assistance to achieve this reload schedule.
If you have any questions or require additional information on this, please contact us.
Very truly yours, L. N. Hartz Vice President - Nuclear Engineering Attachment Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060 Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station
COMMONWEALTH OF VIRGINIA)
)
COUNTY OF HENRICO )
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.
Acknowledged before me this --
___day of ,,ý , 2Z.
My Commission Expires: b - , 200.
Notary Public (SEAL)
ATTACHMENT Small Break LOCA Evaluation of Advanced Mark-BW Fuel Framatome Fuel Transition Program Virginia Electric and Power Company (Dominion)
North Anna Power Station Units I and 2
A SMALL BREAK LOCA EVALUATION OF ADVANCED MARK-BW FUEL FOR APPLICATION TO NORTH ANNA POWER STATION UNITS 1 AND 2 Framatome ANP will be delivering Advanced Mark-BW reload fuel to the North Anna Power Station (NAPS) Units 1 and 2 starting in the first quarter of 2003. The units are Westinghouse-designed, three-loop plants operating at a rated thermal power of 2,893 MWt. The plants have conventional ECCS systems and dry, sub-atmospheric containment buildings. In accordance with 10 CFR 50.46 and 10 CFR 50, Appendix K, an evaluation of ECCS performance is being performed for the Framatome ANP fuel.
One component of the overall LOCA evaluation is a small break LOCA (SBLOCA) assessment. The North Anna evaluation for SBLOCA follows a course employed (and approved by the NRC) in two prior transitions to Framatome ANP fuel for plants with recirculating steam generators (References 1 and 2). The approach involves demonstrating that the current Westinghouse SBLOCA licensing basis is equally applicable to Framatome ANP Advanced Mark-BW reload fuel. The potential impact of design feature differences between North Anna Improved Fuel (NAIF) and Advanced Mark-BW fuel on SBLOCA transient behavior was assessed. It was concluded that the existing SBLOCA analysis provides a conservative representation of Advanced Mark BW fuel.
SBLOCA Transient Description SBLOCA transients can be characterized as developing in five distinct phases: (1) subcooled depressurization, (2) pump/loop flow coastdown and natural circulation, (3) loop draining, (4) vessel/core boil-off, and (5) long-term cooling. These five transient phases are examined in the following paragraphs. This transient characterization provides a framework in which to evaluate the effects of fuel assembly design differences (between NAIF and Advanced Mark-BW fuel) on SBLOCA for the NAPS units.
The limiting SBLOCA event begins with a subcooled reactor coolant system (RCS) depressurization. Depressurization continues until the primary system pressure reaches the saturation pressure associated with the initial hot leg temperature. During this depressurization phase, the low-pressure reactor trip and ECCS injection trip signals are generated. Reactor coolant pumps trip (either manually or in response to loss-of-offsite power). This initiates the pump and loop flow coastdown period.
Following reactor trip, the core power drops sharply. The initial forced flow and subsequent coastdown flow provide continuous heat removal via the steam generators.
Thus, the initial stored energy, and the core power and decay heat during this transient phase are rejected directly to the steam generators. The pump coastdown and natural circulation flows during the second transient phase are sufficient to prevent critical heat flux (CHF) from occurring in the core. Consequently, the fuel pins are cooled toward the quasi-steady temperature distribution required to simply conduct and convect the decay heat energy from the pins. The pin temperatures approach the RCS saturation Page 1 of 5
temperature. Loss of continuous loop flow marks the end of this second transient phase.
The third transient phase is characterized as a period of loop draining. During this period, the system reaches a quiescent state in which core decay heat, break flows, pumped ECCS injection, and steam generator heat transfer combine to control the development of steam-water mixture levels within the RCS. The system inventory distribution is a strong function of the system geometry and break location. RCS liquid inventory will continue to decrease until component mixture levels provide a continuous path through which to vent steam produced in the core. Relief of core-produced steam allows the RCS to further depressurize and enter the boil-off mode.
The development and timing of events that mark the end of loop draining and the onset of core boil-off are governed by break location. For hot leg breaks, a continuous core steam-venting path is readily established. For cold leg breaks, a significant system inventory loss is required to establish a steam vent path. The limiting SBLOCA occurs in the cold leg pump discharge piping. In these breaks, liquid inventory is lost until primary levels descend to the pump suction piping spill-under elevation, which is the low point in the cold leg pump suction piping. This trap (loop seal) must be cleared of liquid to establish a steam-venting path to the break. Since the loop seal elevation is located slightly above the core mid-plane, the core collapsed liquid level is depressed by the manometer pressure balance imposed by RCS geometry. Once the loop seals clear, a steam-venting path is established and the residual liquid inventory in the pump discharge and downcomer regions drains into the core region.
The onset of the boil-off period typically coincides with the beginning of a final saturated depressurization. Voiding at the break increases the leak volumetric flow rate. This ultimately depressurizes the system until the accumulator fill pressure is reached or the pumped ECCS injection matches core steaming. During this period, the reactor vessel mixture level may drop into the heated core region. The fuel pin clad temperature excursions calculated for the upper core elevations are maximized by the assumption of a bounding, core outlet-skewed axial power distribution.
The clad temperature excursion is arrested as the combined ECCS flows exceed the core decay heat level and the final core refill begins. The suppression of core steam production further depressurizes the RCS. This increases ECCS injection flow and hastens core refill. Eventually the RCS will depressurize to the containment pressure and the core will refill. At this point, the start of long-term cooling is established and the transient mitigated.
Fuel Design Effects SBLOCA transients are controlled primarily by system design and core decay heat levels. Fuel assembly design influences calculated events only to the extent that it affects overall system behavior. In that regard, differences between the Advanced Page 2 of 5
Mark-BW and resident NAIF assemblies should not materially affect the bounding SBLOCA transients set forth in the North Anna UFSAR (Reference 3). The Framatome ANP and Westinghouse assemblies have important commonalties: clad OD and ID, and pellet OD. They also differ in several areas: mid-span mixing grids (MSMGs),
unrecoverable pressure drops across the assemblies, initial fuel temperature, clad material, and initial pin backfill pressure. The impact of these variations, with respect to the controlling aspects of the SBLOCA transient, is evaluated below.
The incorporation of MSMGs into the Advanced Mark-BW design creates three effects regarding application of the UFSAR SBLOCA analysis: 1) core CHF performance is improved, 2) convective heat transfer is improved, and 3) the core pressure drop is increased. With MSMGs present, the fuel critical heat flux is higher in the MSMG region than it is at the same location in NAIF. Thus, the Advanced Mark-BW fuel is less likely to experience a departure from CHF during the flow coastdown phase than NAIF. This makes the application of the NAIF calculations to the Advanced Mark-BW fuel during this transient phase conservative. The effect of the MSMGs on convective heat transfer is important in the upper steam-cooled regions of the core during the core-uncovering phase. In this phase, steam generated in the core below the mixture cools the upper core by convection. During this period, two parameters control clad temperature: vapor temperature and the differential temperature between the vapor and the cladding. The vapor temperature is controlled by the decay heat rate and is not influenced by fuel assembly design differences. The heat transfer coefficient near the MSMGs, however, increases with the result that the differential temperature between the clad and the vapor is decreased. Thus, the Advanced Mark-BW fuel will be somewhat lower in temperature near the MSMGs than NAIF and the application of the NAIF calculations to the Advanced Mark-BW fuel is conservative.
The effect of core pressure drop on the SBLOCA calculation is encompassed by the existing UFSAR analysis. Realistically, the use of MSMGs will decrease the core flow by one or two percent and cause a small increase in the core outlet temperature.
Analytically, however, SBLOCA evaluations are performed using a design flow assumption that is substantially reduced from the actual plant flow. Because the analytical assumptions for the initial system flow encompass the expected system flow, the existing SBLOCA calculation remains applicable to the Advanced Mark-BW fuel.
Changes in the initial fuel temperature (stored energy) add or subtract overall RCS energy. The initial fuel energy is removed from the fuel pin during the reactor coolant pump coastdown phase and rejected from the system via the steam generators.
Therefore, the initial fuel stored energy has virtually no impact beyond the loop coastdown period. The core energy release during the loop draining and boil-off mode will be identical to that in the current licensing basis.
Both the Advanced Mark-BW and the NAIF assemblies use advanced cladding material M for NAIF). The materials are T
(i.e., M5TM for the Framatome ANP fuel and ZIRLO comparable to each other, exhibiting analogous physical properties. The required Baker-Just oxidation model is conservative for both materials, and both materials can Page 3 of 5
be simulated with NUREG-0630 type swelling and rupture models. The rupture temperature curves are similar and the rupture strain correlations exhibit similar trends.
Thus, due to material commonalties, no SBLOCA analysis impact would be anticipated.
The Advanced Mark-BW fuel pin backfill pressure is similar, but somewhat less than NAIF. The internal gas pressure can affect fuel/cladding gap dimensions and rupture time. During the initial phase of the accident, the fuel temperature decreases rapidly (within a fraction of a minute following reactor trip) to a level consistent with the rejection of decay heat. The cladding temperature also decreases and approaches the system saturation temperature. Hence, the impact of small gap differences is negligible.
The fuel pin internal gas pressure effect on rupture favors rupture in NAIF over the Advanced Mark-BW fuel. NAIF fill pressure is somewhat higher than that of the Advanced Mark-BW fuel and it would remain so during a SBLOCA transient. Hence, the Framatome ANP fuel would not cause an occurrence of clad rupture where none was predicted for the higher pressure NAIF assembly. Moreover, the limiting UFSAR SBLOCA transients do not predict clad rupture. Thus, the NAIF SBLOCA licensing base is conservative for application to the Framatome ANP fuel.
Finally, SBLOCA-imposed plant operating limits, including maximum allowable total peaking, will not be altered due to the use of Framatome ANP fuel. Thus, the axial power profile used in the existing SBLOCA analysis remains bounding. This assures that the thermal load imposed on the fuel during a temperature excursion remains conservatively modeled. The thermal results, in terms of cladding temperatures, for the current UFSAR evaluations are, therefore, conservative for Advanced Mark-BW fuel.
In summary, core resistance variations will not affect loop flows such that the controlling hot leg temperature or CHF points are altered. The steam generator heat removal rate during the flow coastdown period will compensate for any initial fuel stored energy fluctuations. All controlling parameters in the phases following the pump coastdown and natural circulation phase will be unchanged. Therefore, since the overall RCS geometry, initial operating conditions, licensed power, and governing phenomena are effectively unchanged, the existing UFSAR calculations should remain bounding for operation of the NAPS units with Framatome ANP-supplied Advanced Mark-BW fuel.
Current UFSAR Results The UFSAR small break accident calculations for the NAPS units are not the limiting LOCA transients, according to the predictions of the NOTRUMP and LOCTA-IV computer codes. The calculated results documented in the current North Anna UFSAR 0
predict peak SBLOCA cladding temperatures of about 1,700 F. All parameters are well within the acceptance criteria limits of 10 CFR 50.46. Even wide variations in SBLOCA results would not cause the SBLOCA to be limiting. Thus, considerable margins exist such that variations in the SBLOCA results would not alter either the plant technical specifications or operating procedures.
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Compliance with Acceptance Criteria The existing SBLOCA calculations contained in the North Anna UFSAR are valid and bounding for the Framatome ANP Advanced Mark-BW fuel. The reactor coolant system, decay heat levels, and other system controlling parameters remain unchanged by the reload fuel. A significant safety margin exists between the calculated results and 10CFR50.46 limits. Design differences between the Westinghouse and Framatome ANP fuel do not substantially alter the results of SBLOCA evaluations. Adequate core cooling has been demonstrated by the existing analysis and does not need to be repeated due to the change in fuel vendors. The current SBLOCA calculations remain valid for the NAPS fuel reloads supplied by Framatome ANP. The current UFSAR assessment remains as the SBLOCA analysis of record for demonstrating compliance with the criteria of 10CFR50.46.
References
- 1. BAW-10174A, Revision 1, "Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units," September 1992.
- 2. Letter: Roby Bevan (NRC) to James E. Cross (Portland General Electric Company), "NRC Staff Evaluation of Topical Report BAW-1 0177 (TAC No.
80468)," September 24,1991.
- 3. NAPS UFSAR, Revision 37.
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