Similar Documents at Surry |
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Category:Legal-Affidavit
MONTHYEARML23096A2982023-04-0606 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 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 ML22322A1822022-11-18018 November 2022 Proposed Emergency Plan Revision - Relocation of the Technical Support Center ML21294A3382021-10-21021 October 2021 Proposed Amendment to Relocate Unit Staff Qualification Requirements from Technical Specifications to Nuclear Facility Quality . 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 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 ML21084A2732021-03-24024 March 2021 License Amendment Request - Application of Leak-Before-Break Methodology to Reactor Coolant System Branch Piping - Response to Request for Additional Information and Clarification 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 ML20296A6232020-10-22022 October 2020 Request for NRC Approval to Apply Leak-Before-Break Methodology to Reactor Coolant System Branch Piping ML20149K6942020-05-28028 May 2020 Response to Request for Additional Information, to Proposed License Amendment Requests Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident ML19343A0192019-12-0606 December 2019 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors ML19267A0422019-09-16016 September 2019 Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information Set 2 ML19253B3302019-09-0303 September 2019 Response to Requests for Additional Information - Sets 3 and 4 - Subsequent License Renewal Application ML19204A3572019-07-17017 July 2019 Subsequent License Renewal Application Response to Request for Additional Information - Set 2 ML19196A1092019-07-0909 July 2019 Stations, Units 1 and 2 - Proposed License Amendment Requests Re Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident (SBLOCA) Response to Request for Additional. ML19183A3862019-06-27027 June 2019 Subsequent License Renewal Application - Response to Requests for Additional Information - Set 1 ML19095A5972019-04-0202 April 2019 Submittal of Supplement to Subsequent License Renewal Application Change Notice 2 ML18291A8352018-10-15015 October 2018 Submittal of Application for Subsequent Renewed Operating Licenses ML18218A1702018-07-31031 July 2018 Proposed License Amendment Request Revision of Analytical Methodology Reference in Core Operating Limits Report for Small Break Loss of Coolant Accident ML17251A8972017-08-24024 August 2017 Enclosure 3 to 17-296, Affidavit from Structural Group ML17251A8982017-08-16016 August 2017 Affidavit from Simpson Gumpertz & Heger, Inc ML17055C3602017-02-23023 February 2017 Submittal of Supplement to Proposed Alternative to ASME Section XI, Requirements for Repair/Replacement of Buried Circulating and Service Water Class 3 Piping with Carbon Fiber Reinforced Polymer ML15324A0442015-11-12012 November 2015 Request for Withholding Proprietary Information from Public Disclosure ML0501903172004-12-17017 December 2004 Virginia Electric and Power Company Surry Power Station Units 1 and 2 - Proposed Technical Specifications Change Request for Reactor Coolant System Pressure/Temperature Limits. Ltops Setpoint, and Ltops Enable Temperature with Exemption. ML19093B5491980-09-18018 September 1980 Motion for a Stay Pending Appeal ML19093B3211979-04-0202 April 1979 Licensee'S Answer to Show Cause Orders ML19093B0171978-04-27027 April 1978 Attached Herewith Are 40 Copies of Additional Information Concerning the Steam Generator Replacement at the Surry Power Station ML19093B0321977-12-0202 December 1977 Provide Additional Information Concerning Steam Generator Replacement, Submitting as Revision 1 to Steam Generator Program, Surry Power Station, Unit Nos. 1 & 2. ML19093B0121977-08-17017 August 1977 Attached Herewith Certain Information Which Identified & Marked as Proprietary, Being Submitted in Conjunction with Document Entitled Steam Generator Program, Surry Power Station, Unit Nos. 1 & 2. ML19098B4491977-07-0505 July 1977 Response to Letter of 5/20/1977, Furnishing Information Re Postulated Fuel Handling Accident in Containment & Submitting Revised Analysis Including Proposed Tech Spec ML19094B4611975-07-23023 July 1975 Request for Amendment to Operating Licenses, Technical Specification Change No. 32. Changes Update Organizational Chart & Maximize Effectiveness of System Nuclear Safety & Operating Committee ML19093B3341975-07-17017 July 1975 Request for Amendment to Operating Licenses DPR-32 & DPR-37 for Technical Specification Change No. 31 2023-04-06
[Table view] Category:Report
MONTHYEARML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary 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 ML23054A1922023-02-22022 February 2023 ASME OM Code Inservice Testing Program Alternative Request V-01, Use of Mechanical Agitation Process for Pressure Isolation Valve 1-SI-241 Seat Leakage Testing Supplemental Information 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 ML22263A3642022-05-23023 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 1 - 95 ML22263A3692022-05-0101 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 96-190 ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML20091L9852021-05-0404 May 2021 Record of Decision ML20091L9832021-05-0404 May 2021 SLR License 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 ML20296A6802020-10-22022 October 2020 ASME Section XI Inservice Inspection Program Request for NRC Approval of Proposed Alternative S1-I5-ISL-05 and S2-15-ISL-06 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 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 ML20076A5762020-05-28028 May 2020 Staff Assessment of Flood Hazard Focused Evaluation and Integrated Assessment ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19235A0702019-08-19019 August 2019 Cera Blanket Website ML19235A0802019-08-19019 August 2019 Thermo-Lag Website ML19235A0742019-08-19019 August 2019 Kbs Sealbags Website ML19235A0682019-08-19019 August 2019 3M Interam Website ML19235A0722019-08-19019 August 2019 Duxseal Website ML19235A0782019-08-19019 August 2019 Mineral Wool Website ML19235A0692019-08-19019 August 2019 Bio K-10 Website ML19235A0732019-08-19019 August 2019 Gypsum Website ML19235A0712019-08-19019 August 2019 Cerafiber Website ML19235A0752019-08-19019 August 2019 Marinite Slide Plates Products - Piping Tech ML19235A0762019-08-19019 August 2019 Marinite Website ML19235A0772019-08-19019 August 2019 Micarta Website ML19235A0792019-08-19019 August 2019 Pyrocrete Website ML19204A3582019-06-30030 June 2019 WCAP-18258-NP, Revision 1, Flaw Tolerance Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel Elbow Components for Surry Units 1 and 2. ML19163A1542019-06-0606 June 2019 SLRA - Attachment 4D Letter - Draft to Dominion ML19140A2762019-05-20020 May 2019 Essential Fish Habitat Mapper Results for Surry Power Station ML19095A6022019-04-0202 April 2019 Enclosure 2 - SLRA Mark-Ups Change Notice 2 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 ML18218A1712018-07-31031 July 2018 Attachment 5, ANP-3676NP, Revision 0, Surry Fuel-Vendor Independent Small Break LOCA Analysis, Licensing Report ML18108A1232018-04-11011 April 2018 Proposed Alternative Requests S1-15-ISl-04 and S2-15-ISl-04 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55a(z)(1) ML17325A7792017-11-22022 November 2017 Review of Steam Generator Tube Report for the Fall 2016 Inspections During Refueling Outage 27 (CAC No. MF9683; EPID L-2017-LRO-0009) ML17339A1752017-09-0707 September 2017 Structural Integrity Associates Report No. 1601007.401, September 2017, Flaw Tolerance Evaluation of the Hot Leg Surge Line Nozzles Using ASME Code Section XI, Appendix L. ML17241A0462017-08-24024 August 2017 Submittal of Cycle 28 Startup Physics Tests Report ML17251A9002017-06-0505 June 2017 Enclosure 2, Attachment E to 17-296, WEKO-SEAL - Recommended Installation Procedures for Water and Waste-Water Piping, Miller Pipeline, Revised 03/05/17 ML16355A1722016-12-20020 December 2016 Storm Surge Technical Evaluation Report Surry Power Station Hazard Reevaluation Review ML15335A0932015-12-15015 December 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 for Recommendation2.1 of the Near-Term Task Force Review ML15244B2052015-08-24024 August 2015 Virginia Electric and Power Company (Dominion) Surry Power Station Unit 1, Cycle 27 Startup Physics Tests Report ML15097A0782015-04-0707 April 2015 Soarca Surry 2005 Site File ML15072A2992015-03-16016 March 2015 Review of the Fall 2013 Steam Generator Tube Inservice Inspection Report Regarding Surry Power Station, Unit 1 Refueling Outage 25 ML15078A2882015-03-12012 March 2015 Engineering Evaluation 14-E15, Dominion Flooding Hazard Reevaluation Report, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Revision 1, Pp. 2-115 Through 2-218 2023-05-31
[Table view] Category:Technical
MONTHYEARML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary 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 ML22263A3642022-05-23023 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 1 - 95 ML22263A3692022-05-0101 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 96-190 ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- 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 ML20296A6802020-10-22022 October 2020 ASME Section XI Inservice Inspection Program Request for NRC Approval of Proposed Alternative S1-I5-ISL-05 and S2-15-ISL-06 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 ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19235A0802019-08-19019 August 2019 Thermo-Lag Website ML19235A0682019-08-19019 August 2019 3M Interam Website ML19235A0692019-08-19019 August 2019 Bio K-10 Website ML19235A0702019-08-19019 August 2019 Cera Blanket Website ML19235A0792019-08-19019 August 2019 Pyrocrete Website ML19235A0712019-08-19019 August 2019 Cerafiber Website ML19235A0722019-08-19019 August 2019 Duxseal Website ML19235A0732019-08-19019 August 2019 Gypsum Website ML19235A0742019-08-19019 August 2019 Kbs Sealbags Website ML19235A0752019-08-19019 August 2019 Marinite Slide Plates Products - Piping Tech ML19235A0762019-08-19019 August 2019 Marinite Website ML19235A0772019-08-19019 August 2019 Micarta Website ML19235A0782019-08-19019 August 2019 Mineral Wool Website ML19204A3582019-06-30030 June 2019 WCAP-18258-NP, Revision 1, Flaw Tolerance Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel Elbow Components for Surry Units 1 and 2. ML19095A6022019-04-0202 April 2019 Enclosure 2 - SLRA Mark-Ups Change Notice 2 ML19011A1742019-01-0404 January 2019 Enclosure 5 - Surry Power Station, EAL Scheme Revisions-Supporting Documents ML19011A1732019-01-0404 January 2019 Enclosure 4 - North Anna Power Station Units 1 & 2, EAL Scheme Revisions-Supporting Documents ML19011A1722019-01-0404 January 2019 Enclosure 3, Attachments 2C-3C - MPS3 EAL Technical Bases Document (Marked-Up) ML18218A1712018-07-31031 July 2018 Attachment 5, ANP-3676NP, Revision 0, Surry Fuel-Vendor Independent Small Break LOCA Analysis, Licensing Report ML17339A1752017-09-0707 September 2017 Structural Integrity Associates Report No. 1601007.401, September 2017, Flaw Tolerance Evaluation of the Hot Leg Surge Line Nozzles Using ASME Code Section XI, Appendix L. ML17241A0462017-08-24024 August 2017 Submittal of Cycle 28 Startup Physics Tests Report ML17251A9002017-06-0505 June 2017 Enclosure 2, Attachment E to 17-296, WEKO-SEAL - Recommended Installation Procedures for Water and Waste-Water Piping, Miller Pipeline, Revised 03/05/17 ML16355A1722016-12-20020 December 2016 Storm Surge Technical Evaluation Report Surry Power Station Hazard Reevaluation Review ML15244B2052015-08-24024 August 2015 Virginia Electric and Power Company (Dominion) Surry Power Station Unit 1, Cycle 27 Startup Physics Tests Report ML14086A1732014-04-0909 April 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 ML14079A0182014-02-27027 February 2014 30-Day Report of Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML14002A1452014-02-19019 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14043A4442014-02-12012 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Surry Power Station, Units 1 and 2, TAC Nos.: MF1002 and MF1003 (Revision 2) ML13179A0152013-05-31031 May 2013 Units I and 2, Qualification of the ABB-NV and Wlop CHF Correlations in the Dominion VIPRE-D Computer Code ML11215A0592011-06-30030 June 2011 WCAP-17345-NP, Rev 2, H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (3-Loop Model 44F/ Model 51F), Attachment 5 to 11-403 ML1023904192010-08-31031 August 2010 DOM-NAF-2, Rev. 0.2-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code. ML1026002162010-05-13013 May 2010 Engineering Transmittal, ET-CEP-10-0006, Revision 0, Evaluation of Aluminum Conduit Seal Penetration Fire Tests, Dated May 13, 2010, North Anna and Surry Power Stations Units 1 and 2 (Redacted Version), Enclosure 2 ML0930701642009-10-15015 October 2009 Technical Report No. EE-0116, Revision 5 ML0926606162009-08-27027 August 2009 LTR-SGMP-09-108 NP-Attachment, Response to NRC Request for, Additional Information on H*; Model 44Fand Model 51F Steam Generators. ML0921504622009-06-30030 June 2009 WCAP-17092-NP, Revision 0, H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model 51F), Attachment 6 ML0921202022009-05-0606 May 2009 LTR-SGMP-09-64 Rev. 2 Np, Surry, Unit 1, Support for Emergency Technical Specification Amendment to Address Permeability Variation Signals at the Tube Ends. ML0826107152008-09-17017 September 2008 Final Precursor Analysis - Surry 2 (LER 281/06-002) ML0823200152008-08-18018 August 2008 Cycle 22 Startup Physics Tests Report 2023-05-31
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARML23220A1482023-08-0707 August 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Response to Request for Additional Information 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 L-04-002, Stations - NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Fleet Response to Request for Additional Information2023-05-0808 May 2023 Stations - NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Fleet Response to Request for Additional Information ML23096A1972023-04-0606 April 2023 License Amendment Request - Application of Risk-Informed Approach for Tornado Classification of the Fuel Handling Trolley Support Structure - Response to Request for Additional Information 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 ML23089A3542023-03-30030 March 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Emergency Plan Staff Augmentation Times ML23044A0392023-02-11011 February 2023 Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles Response to Request for Additional Information ML23025A1252023-01-25025 January 2023 Response to Request for Additional Information License Amendment Request, NRC Approval of Methodology Change and Reclassification of the Turbine Building as a Tornado Resistant Structure ML22343A0002022-12-0909 December 2022 Response to Request for RAI on ASME Operation & Maintenance Code Inservice Testing Program - Request for Approval of Alternative Request V-01 Use of Mechanical Agitation Process for Pressure Isolation Valve 1-SI-241 Seat Leakage Testing ML22342A4242022-12-0707 December 2022 Response to 3/12/2012 Request for Information Enclosure 2, Recommendation 2.0, Flooding Focused Evaluation Integrated Assessment Submittal Revised Commitment Completion Date 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 ML22300A1882022-10-27027 October 2022 License Amendment Request 10-Day Allowed Outage Time for Opposite Unit Auxiliary Feedwater Cross-Connect Capability Response to Request for Additional Information ML22220A2162022-08-0808 August 2022 License Amendment Request 10-Day Allowed Outage Time for Opposite Unit Auxiliary Feedwater (AFW) Cross-Connect Capability Supplemental Information ML22164A8532022-06-13013 June 2022 Proposed License Amendment Request Removal of Refueling Water Chemical Addition Tank and Replacement of Containment Sump Buffer - Response to Request for Confirmation ML22153A1352022-06-0202 June 2022 Proposed License Amendment Request (LAR) - Removal of Refueling Water Chemical Addition Tank and Replacement of Containment Sump Buffer - Response to Request for Additional Information ML21292A3022021-10-19019 October 2021 Response to Question on License Amendment Request to Revise Reactor Core Safety Limit 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 ML21133A4212021-05-13013 May 2021 Response to NRC Request for Additional Information Regarding Proposed License Amendment Request Update of LOCA Alternative Source Term Dose Analysis 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 ML21098A2612021-04-0808 April 2021 Subsequent License Renewal Application, Response to Request for Additional Information Re Environmental Review ML21084A2732021-03-24024 March 2021 License Amendment Request - Application of Leak-Before-Break Methodology to Reactor Coolant System Branch Piping - Response to Request for Additional Information and Clarification ML21056A5412021-02-25025 February 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 ML21042B9262021-02-11011 February 2021 ASME Section XI Inservice Inspection Program Request for NRC Approval of Proposed Alternative Requests S1-I5-ISl-05 and S2-I5-ISl-06 - Response to Request for Additional Information ML22040A1012021-02-0909 February 2021 Response to Request for Additional Information 2021 Steam Generator Inservice Report ML20296A7272020-10-22022 October 2020 Stations Units 1 & 2 - Proposed License Amendment Requests Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break LOCA Supplement to the Response to Request for Additional Information ML20247J6352020-09-0303 September 2020 Virginia Electric & Power Co - Subsequent License Renewal Application (SLRA) Environmental Review, Response to Request for Additional Information Concerning Aquatic Resources ML20181A4222020-06-29029 June 2020 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Response to Request for Additional Information ML20181A3082020-06-23023 June 2020 Response to Request for Additional Information on 2019 Steam Generator Inservice Inspection Report ML20269A4072020-06-0202 June 2020 Enclosure 2 - Dominion Energy Surry Power Station Vpdes Permit No. VA0004090: Application for Permit Reissuance - 316(b) 40 CFR 122.21(r)(2)-(13), Part 2 of 3 ML20269A4062020-06-0202 June 2020 Enclosure 2 - Dominion Energy Surry Power Station Vpdes Permit No. VA0004090: Application for Permit Reissuance - 316(b) 40 CFR 122.21(r)(2)-(13), Part 3 of 3 ML20269A4082020-06-0202 June 2020 Enclosure 2 - Dominion Energy Surry Power Station Vpdes Permit No. VA0004090: Application for Permit Reissuance - 316(b) 40 CFR 122.21(r)(2)-(13), Part 1 of 3 ML20149K6942020-05-28028 May 2020 Response to Request for Additional Information, to Proposed License Amendment Requests Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident ML20034F0532020-01-31031 January 2020 Proposed License Amendment Request: Allowed Outage Time Extension for Replacement of Reserve Station Service Transformer C 5KV Cables to Transfer Bus F Response to Request for Clarification and Additional Information ML19350A0042019-12-12012 December 2019 License Amendment Request Addition of 24-Hour Completion Time for an Inoperable Reactor Trip Breaker Response to NRC Request for Additional Information ML19283B6222019-10-10010 October 2019 Subsequent License Renewal Application Safety Review - Documentation of Draft Third Round Requests for Additional Information ML19291B0342019-10-0101 October 2019 Response to March 12, 2012 Request for Information Enclosure 2, Recommendation 2.1, Flooding Focused Evaluation/Integrated Assessment Submittal ML19269B7342019-09-19019 September 2019 Subsequent License Renewal Application - Response to Requests for Additional Information - Sets 3 and 4 Revised SLRA Mark-Ups ML19267A0422019-09-16016 September 2019 Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information Set 2 ML19253B3302019-09-0303 September 2019 Response to Requests for Additional Information - Sets 3 and 4 - Subsequent License Renewal Application 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 ML19204A3572019-07-17017 July 2019 Subsequent License Renewal Application Response to Request for Additional Information - Set 2 ML19198A0592019-07-11011 July 2019 Subsequent License Renewal Application - Response to NRC Requests for Confirmation of Information ML19196A1102019-07-0909 July 2019 Attachments 2, 3 & 4 - Response to NRC Request for Additional Information and Correction of SBLOCA Analysis Error and Revised Marked-Up/Typed NAPS and SPS Technical Specifications Pages ML19183A3862019-06-27027 June 2019 Subsequent License Renewal Application - Response to Requests for Additional Information - Set 1 ML19148A4212019-05-10010 May 2019 All Enclosures with Coversheets ML18304A3302018-10-25025 October 2018 Proposed License Amendment Request TSTF-490 & Update of Alternative Source Term Analyses Response to Request for Additional Information & Revised Proposed Technical Specifications ML18302A1212018-10-18018 October 2018 Response to Request for Additional Information, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules ML18303A2882018-10-18018 October 2018 Response to Request for Additional Information Regarding Relief Request S2-I5-LMT-P01 Examination Category R-A, Preservice Weld Examinations ML18267A0682018-09-18018 September 2018 Attachment - Response to Request for Clarification on Proposed Consolidated Emergency Operations Facility License Amendment Requests ML18212A2192018-07-25025 July 2018 ASME Section XI Inservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the First Period of the Fifth 10-Year Inservice Inspection Interval Response to Request for Additional Information 2023-08-07
[Table view] Category:Technical Specifications
MONTHYEARML23136B1392023-06-29029 June 2023 Issuance of Amendment No. 313 Administrative Changes to Subsequent Renewed Operating Licenses and Technical Specifications ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML22300A1882022-10-27027 October 2022 License Amendment Request 10-Day Allowed Outage Time for Opposite Unit Auxiliary Feedwater Cross-Connect Capability Response to Request for Additional Information ML21253A0632021-11-19019 November 2021 Issuance of Amendment Nos. 306 and 306 to Update of the Loss of Coolant Accident Alternate Source Term Dose Analysis ML21277A0652021-09-30030 September 2021 Proposed License Amendment Request Removal of Refueling Water Chemical Addition Tank and Replacement of Containment Sump Buffer ML21175A1852021-08-20020 August 2021 Issuance of Amendment Nos. 304 and 304 Leak-Before-Break for Pressurizer Surge, Residual Heat Removal, Safety Injection Accumulator, Reactor Coolant System Bypass and Safety Injection Lines ML21106A2792021-05-0404 May 2021 Technical Specifications ML20301A4522020-10-28028 October 2020 Non-Proprietary, Issuance of Amendment Nos. 300 and 300 to Revise Technical Specifications 6.2.C, Core Operating Limits Report ML20115E2372020-05-0707 May 2020 Issuance of Exigent Amendments Nos. 299 and 299, to Revise TS 6.4.Q, Steam Generator (SG) Program, to Allow a One-Time Deferral of the Inspection of Surry Unit 2 Sg B Spring 2020 Refueling Outage ML19196A1102019-07-0909 July 2019 Attachments 2, 3 & 4 - Response to NRC Request for Additional Information and Correction of SBLOCA Analysis Error and Revised Marked-Up/Typed NAPS and SPS Technical Specifications Pages ML17326A2252018-02-0909 February 2018 Issuance of Amendments Regarding License Amendment Request to Revise Residual Heat Removal and Component Cooling System Technical Specifications Requirements (CAC Nos. MF9124 and MF9125; EPID L-2017-LLA-0166) ML15245A4822015-10-0707 October 2015 Millstone Power Station, Surry Power Station, and North Anna Power Station - Issuance of Amendments to Revise the Cyber Security Milestone 8 Completion Date in the Renewed Facility Operating Licenses ML14169A3592014-08-12012 August 2014 Stations, Units 1 and 2, Issuance of Amendments Regarding Addition of an Analytical Methodology to the Core Operating Limits Reports and an Increase to the Surry Minimum Temperature for Criticality ML14148A2352014-07-0303 July 2014 Issuance of Amendment to Extend the Containment Type a and Type C Leak Rate Test Frequency ML13260A2502013-09-0909 September 2013 Proposed License Amendment Request Regarding Relocation of Snubber Surveillance Requirements from Technical Specifications 4.17 to Snubber Program - Supplement ML12279A2822012-09-26026 September 2012 Proposed License Amendment Request Regarding Temporary Service Water Jumper to the Component Cooling Heat Exchangers ML12222A3402012-07-31031 July 2012 Proposed Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process ML12109A2702012-04-17017 April 2012 Issuance of Amendments Regarding Virginia Electric and Power Company License Amendment Request for Permanent Alternate Repair Criteria for Steam Generator Tube Inspection and Repair ML12081A0162012-03-14014 March 2012 Supplemental Response to Request for Additional Information Related to License Amendment Request for Permanent Alternate Repair Criteria for Steam Generator Tube Inspection and Repair ML1106602302011-03-0404 March 2011 Supplemental Information &Typed TS Pages for Proposed License Amendment Request Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirement to Licensee Controlled Program (Adoption of TSTF-425, Rev 3) ML1035502062010-12-16016 December 2010 Proposed Technical Specification Change Temporary Alternate Repair Criteria for Steam Generator Tube Repair for Unit 2 ML1030502732010-11-0202 November 2010 Correction to Amendment Regarding Technical Specification Revisions Related to the Measurement Uncertainty Recapture (Mur) Power Uprate ML1027203442010-09-28028 September 2010 Attachment 2 to Email Regarding Typed Page 2.3-2 for Surry Dnbcolr LAR ML1013106042010-05-0606 May 2010 License Amendment Request Revised Cumulative Core Burnup Applicability Limit for Heatup and Cooldown Curves, Low Temperature Overpressure Protection System Setpoint, and Ltops Enable Temperature ML1001405592010-02-26026 February 2010 License Amendment, Correction of Technical Specification Table 3.7.1 Operator Action 3.b ML0929606162009-10-16016 October 2009 Proposed License Amendment Request Relocation of Core Operating Limits to the Core Operating Limits Report (COLR) and Addition of COLR References ML0910401472009-04-13013 April 2009 Proposed License Amendment Request Correction to Technical Specification Table 3.7-1 Operator Action 3.b ML0822500262008-09-17017 September 2008 Units, 1 and 2, Tech Spec Pages for Amendments 261 and 261 Regarding the Revision to Various Setting Limits and the Overtemperature T/Overpower T Time Constants ML0817506992008-07-0707 July 2008 Tech Spec Pages for Amendments 260 & 260 Regarding Control Room Habitability ML0814200442008-06-17017 June 2008 Tech Spec Pages for Amendments 259 and 259, Regarding the Increase of Maximum Service Water Temperature Limit from 95 F to 100 F ML0813401152008-05-16016 May 2008 Technical Specifications - 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MD3168 and MD3169) ML0724906142007-09-0606 September 2007 Technical Specifications, Issuance of Amendments Regarding Use of Westinghouse Best-Estimate Large Break Loss-of-Coolant (TAC Nos. MD3609 & MD3610) ML0723300982007-08-20020 August 2007 Proposed License Amendment Request Control Room Habitability Supplemental Response ML0719703012007-07-13013 July 2007 Proposed License Amendment Request Consolidated Line Item Improvement Process Technical Specification Improvement Control Room Habitability ML0712402632007-05-0303 May 2007 Technical Specifications, Issuance of Amendments Regarding Missed Surveillance Requirements TSTF-358 (TAC Nos. 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[Table view] |
Text
July 5, 'l.977'. :
Mr. Edson G. Case, Acting Director Serial No. 217 A/052077 Nuclear Reactor Regulation PO&M/TAP/ALH:das U. S. Nuclear Regulatory Commission Docket Nos. 50-280 Washington, D. C. 20555 50-281 License Nos. DPR-32 Attn: Mr. Robert W. Reid, Chief DPR-37 Operating Reactors Branch 4
Dear Mr. Case:
This is in.response to your letter of May 20, 1977, in which you address-ed the postulated Fuel Handling Accident in Containment. The bases for our conclusion that the potential consequences of a fuel handling accident in con-tainment are well within the guidelines of 10CFRlOO are as follows:
- 1. The presently installed ventilation systems at Surry are such that only one of two diverse means need function to assure that 10CFRlOO guidelines are met during a postulated fuel handling accident in con-tainment.
- a. The purge isolation system, which is actuated by any one of three seismic qualified radiation monitors, is capable of terminating a radioactive release before 10CFRlOO guidelines are exceeded. This is shown in attached cases 1, 2 and 4.
A loss of off-site power will render the purge isolation sys-tem inoperable, but would also cause all ventilation flow in and out of containment to cease. Since the containment is at atmospheric pressure during fuel handling, such a situation would not result in a radioactive release from containment.
- b. The ventilation filter system which is analyzed in attached case 3 also serves to prevent exceeding 10CFRlOO guidelines during a fuel handling accident. A seismic event might affect the filter performance since the charcoal beds and the duct-ing between the outboard isolation valve and the filter inlet are not seismic qualified.
- 2. The probability of a seismic event at the Surry Station as stated in the Reactor Safety Study (WASH 1400 - NUREG-74/014). is between 10-4 and 10:-6 events per year. The largest period of time that fuel is
.being handled in containment is ten days per 18 months or.1.9 percent of *the time. This results in a conservative probability of a seismic event at the same time as fuel handling accident (given it has a 100%
probability) of 2 x 10-6 events per year.
- e e VIRGINIA ELECTRIC AND POWER COMPANY TO ~r. Edson.G. Case Page 2 We therefore believe that a containment fuel handling accident which results in exceeding 10CFR100 guidelines is not credible. We believe that increased sur-veillance to determine equipment operability.is. sufficient to assure that 10CFR100 guidelines will not be exceeded during the postulated accident. We therefore sub-mit for your study.the attached Technical Specification changes. (.These changes will be pursued through the normal channels if this concept is determined to be ap-propriate.) The changes address the following requirements:
- 1. The purge isolation system will be proven operable, with the closing of the valves in less than 30 seconds, .before and .weekly during fuel handling operations in containment.
- 2. The filters will be proven operable before fuel handling operations in containment.
The attached .revised analysis includes the proposed Technical Specification parameters and should justify all of our assumptions regarding the radioactivity released to the environment.
Very truly yours,
'77
Cr2, )1),1. ':ffd&!c-.c,c')/.1/
C. M. Stallings Vice President-Power Supply and Production Operations Attaclnnents cc: Mr. Norman C. Moseley
Page 1 ~f 11 Revision 1 6-23-77 ATTACHMENT ANALYSIS OF REFUELING ACCIDENT IN CONTAINMENT Assumptions During refueling the contairnaent purge system is set up to exhaust through the ventilation filters in the Auxiliary Building. Both filter banks will be on 'line because the Fuel Building is also exhausting through the filters.
The total flow rate for both systems is 66,000 cfm while the design flow rate for each filter is 36,000 cfm.
While the purge system is in operation, the air flow in the containment is as follows. Air enters the containment through two 15000 cfm fans (FS-F-4A & B), through two 36 inch butterfly supply valves (MOV-VS-100 A &
B) and is dispersed through the ring header outside the crane wall at EL-3 1 6". The air is continuously recirculated inside the containment by three 75,000 cfm recirc fans (VS-F-1 A,B;C). The air is purged from the contain1nent through the ring header at EL-20' outside the crane wall.
The air discharges through two 36 inch butterfly valves in series (MOV-VS-100-C,D). The air then passes through the Auxiliary Building Filter Banks (VS-FL-3 A & B) and the two 15000 cfm Purge Exhaust Fans (VS-P-5 A,B) o The-worst* single failure would be loss of the valve closing circuit which closes the valves and secures the purge fans en an alarm from either the crane monitor or the containment gas and particulate monitors. The two output relays are sufficiently redundant to secure purge flow; however, a loss of power to this circuit would cause them not to function. This would cause a total release through the filters with a boundary dose as calculated for case 3 of the dose calculations. This dose is still within the allowable limits, but our worst case.
Locations of containment gas and particulate radiation monitors The sample line ties . int~ 30" recirc, duct in containment at EL-9' 10".
The monitors are located in the Auxiliary Building at EL 45'10" along the west wall.
Sample line 1 11 ARC-2-21B which connects the containment particulate and gaseous monitor to the recirculation duct Length from 11448-FK-10 A & B 214 ft.
System flo~ from 11448-SN-207 10 cfm 3
Volume 0.870 ID Tube-= 1526.6in Response Time of sample to be detected in the containment articulate and aseous monitor 3 3 3 3 Flow == 10 Ft /min X miri/60 sec X l 728in /Ft = 288. in /sec Time= 1526.6 in 3
~ 288 in 3 /sec = 5.3 secondsfuse 30 seconds for conservatism
Page 2 of 11 e
. *--*t/ . *
- Air !Flow Rate 3QOOO cfm for purge system alone to calculate transit time of release in ducting Duct Area 2
= 9o62 Ft Duct Lengths I
Inlet to exhaust valve 130' *I I
Exhaust valve to filters 130' Velocity 3 3 V ~ve = 30,000 Ft /min X min/60 sec X Ft/9.62 Ft V = 51.9 Ft/sec Peak centerline velocity Umax ir 6 Umax
= 0.8 (Re=lc096Xl0 = Turb~lant Flow)
TT.--.~-,.
u1uu..n.. == 5le'): 0.3 = 64.9 FL/sec Transit Times
= 130 T 64.9 = 2 seconds Use le4 seconds to be conservative for transit time in exhaust ducting__
This analysis assumes that the .air flow path in the containment is continuous from the surface of the reactor cavity to EL-27. This is very conservative *as there is no velocity profile across the reactor cavity. Also,it is assumed that 50% of the area outside the crane wall is blocked with structures and equipment. This is a conservative assumption based on figures used in the LOCA analysis. The only mixing vol~me used is the volQme of air the release contacts in the transit from the cavity.to the exnaust inlet. .
Determine the open area outside the crane wall .. * - -* * - .z:. .. _:.. :-; * .
Containment diameter 126 1 Diameter inside crane wall 106 1 2 2 A= [ ~ (63) - 1r (53) ] 0.5 = 1822 Ft 2
Air Velocity Recirc air flow 225,000 cfm Purge air flow 3QOOO cfm Total 25~000 cfm for least possible transit time
\
' Page 3 of 11 3 2 V = 255000 Ft /min X 1/1822 Ft X 1 min/60 sec = 2.33 Ft/sec which is also used for velocity across the cavity Air Transit Time Shortest distance from cavity surface to purge duct suction 70 Ft.
Time= 70 Ft~ 2.33 Ft/sec= 30 sec This is one-half the time assumed in our initial analysis.
Mixing As the transit time of the release is 30 seconds from the cavity to the inlet to the purge and recirculation system, the volume of air contacted during the transit.is considered for mixing.
Total flow 255,000 cfm Transit time 30 sec.
127,500 cubic feet for mixing
(
Page 4 of 11 CASE ANALYSIS PROBLEM BASIS of Supposed Refueling Accident inside of Containment Case 1: The manipulator crane monitor is functional and actuates the containment isolation valves and secures _the purge fans.
Case 2: The manipulator crane area monitor is not functional requiring the containment gas or the containment particulate wonitor to actuate the isolation valves and secure the purge fans.
Case 3: Neither the manipulator crane area monitor or the containment gas and particulate monitors are operational resulting in a total release.
Case 4: Assume worst case isolation (i.e. Case 2) and the filters are not available or the exhaust duct outside of containment and downstream of the isolation valves fails.
Physical Characteristics of Systems and Assumptions Item 1) As~ume a puff release of radioactivity from a ruptured fuel assembly in the reactcir fuel cavity. This puff release is as close to the nearest purge exhaust grill as is physically possible.
Item 2) The response of the area monitor is gamma radiation sensitive so that ir is not necessary for it to be immersed in a radioactive cloud to detect radioactiyity. It's position (i.e. approx. 10 feet) above Lhe fuel cavity, unshielded from direct /-rays from the pool reaffirms its capability to detect an accident release immediately.
Item 3) The travel time of a radioactive cloud from a puff release at a P,Oint on the pool surface to the purge duct suction and recirculation suction is 30 seconds which is twice as conservative as the-initial analysis.
Item 4) The closure time of the isolation valves (including all electrical impulses and p~ocesses) is 20 seconds. (Worst Case at Site 19.4 sec.
for vaives alone) for Tech. Spec. Case use 30 sec. for valves and circuitry.
Item 5) The response time from the recirculation suction to the gas and particulate monitors detection is assumed at 54 seconds., per calculation above. f Item 6) The purge rate is 30000 cubic feet per minute .. per system design and not considering reduced performance due to exhaust through the filter banks. f Item 7) Filtration of containment purge exhaust during refueling is assumed to be conservatively 70% efficient.
Item 8) The assumed volume of containment air with which the radioactive release is mixed, during the 30 second transit 1:ime from the reactor cavity to the purge duct equals 127,500 cubic feet.per calculations above. j*
Item 9) The time for air to travel through the purge duct to the inboard isolation valves is assumed to be 1.4 seconds.
Item 10) The delay time from reactor shutdown to initiation of fuel assembly transfer operations is at least 100 hrs (Tech. Spec.)
I
Page 5 of 11 e e Item 11) The radial peaking factor is assumed to be 1. 65, Item 12) The number of fuel assemblies in the CORE is 157.
Item 13) Ten percent of the Iodine fuel assembly activities are assumed to be released into the reactor cavity water with 99.75% being inorganic and .25% in the organic form. The fuel pool D.F. for inorganic iodine is 133 while the DF for organic iodine is 1.
Item 14) Ten percent of the Noble Gases present in the fuel assembly are released to the reactor cavity* pool with the exception of Kr 85; 30% of Kr 85 is released. The DF of the Fuel Pool water for Noble Gases is 1.
Item 15) Dose calculational method from R.P. 13, section "Dose Conversion Factors and Equations" (includes average garrnna and beta energies for each isotope and thyroid dose conversion factors) 3 3 Item 16) ~Q = 2.1 X 10- sec/m (Latest meteorology)
Item 17) Breathing rate= 3.47 X 10- 4 m: 3 /sec Item 18) Site boundary distance= 1650 ft Item 19) Surry 1 & 2 CORE activities Calculation Method for activity for hottest fuel element Total core activity (µCi)) /. 1 - - ) Pea](ing factor )
( after 100 hr. decay x ~57fuel assy X
( 1. 65
'***~
Page 6 of 11 CASE 1 When a puff release results in airborne and direct gamma irradiation of the Manipulator Crane area monitor so that a prefixed dose rate setpoint is surpassed, activation of the isolation valves occurs instantaneously. It takes 20 seconds to isolate the contairunent. Thirty seconds is required for gaseous transit time to the purge duct. An additional 1.4 seconds is required from the purge duct grill to the 1solation valve. Therefore, if the Manipulator Crane Area Monitor is operational following a Fuel Handli~g Accident, zero release of radioactivity from the containment is expected9 Using the 30 seconds for valve closure (proposed for a Technical Specification Limit) to isolate the containment versus 31.4 seconds for the release to reach the inboard valve results in zero release.
e :pagf 7 __pf 11 CASE 2
(
If the manipulator crane area monitor is not functional, following a fuel handling accident, the containment gas and particulate monitors will be used to actuate the containment isolation valves. The value* and site doses are as follows for a release of 74 seconds.which was ultra-conservative:
IsotQI>_!. Activity for hottest Activity Whole Bo21.(f) Whole Body(.8) Thyroid fuel element (},<~) Released Inorganic I 131 4.46 + 11 2.90 + 01 3.13 + 01 132 133 6.33 5.38
+
02 10 4.02 3o50 +
- 12 00 1.57 -
- 1. 02. +
13 00 134 6.93 - 23 4.51 - 33 8022 - 35 135 4.35 + 07 2o 83 - 03 2. 56 .., 04*
32.32 + 00 Organic I 131 4. 46 + 11 9. 71 + 00 1.05 + 01 132 6.33 - 02 1.38 - 12 5.38 - 14 133 5.38 + 10 1.17 + 00 3.41 - 01 134 6. 93 - 23 1.51 - 33 2.75 - 35 135 4.35 + 07 9.47 - 04 8.56 - 05 1.08 + 01 I 131 7.62 - 03 3.91 - 03 132 6.49 - 15* 1.10 - 15 133 1.56 - 03 9.09 - 04 134 7.93 - 36 1.62 - 36 135 2.90 - 06 8.67 - 07 9.18 - 03 rem 4.82 - 03 rem Kr 83m 85m 1.85 4.07
+
05 04 5.37 - 13 1.18 - 03 1.41 9.67
- 18 08 8.80
- 5. 77
- 18 08 85 6.06 + 09 5.27 + 02 5.80 - 04 5.80 - 02 87 4.03 - 12 1.17 - 19 5.27 - 23 5.70 - 23 88 89 L25 + 01 3.63 - 07 3. 80 - 10 5.40 - 11 Xe 131m 4.51 + 08 1.31 + 01 1.51 - 04 8.53 - 04 133m 9.99 + 09 2.90 + 02 3.63 - 03 2.68 - 02 133 8.54 + 11 2.48 + 04 __
9.60.,., ______
- 01 1.51 + 00 135m 2.60 + 00 + ---------
135 ______+.,,,_07 8.96 3.40 04 3 80 - 04 0
137 138
--------- 9.63 01 rem 1.59 + 00 rem Total 9. 72 - 01 rem 1.59 + 00 rem 43 .12 rem
Page 8 of 11 Case 2 (Continued)
(
Case 2: Total Whole Body dose is 2.04 rem for Tech. Spec. valve closure Total Thyroid dose is 34. 38 rem and a total release of 59
- seconds r Case 2 assumes a 30 second transit time from the reactor cavity fuel pool to the purge duct during which time a certain amount of mixing occurs, as specified by suction off of the recirculation and purge fanso An additional 1.4 seconds is consumed in the purge duct prior to the isolation valve. The containment will be isolated ~O seconds after the radioactivity reaches the purge duct, resulting in a 5ff second release of the concentrations present at the purge duct grill 30 seco~ds following the fuel handling accident.
Calculation Method-for Activity Released Time Release) (l~Filter efficiency)
X ( 5.°l/60 X 0.3( '>1.0 Calculation Method for Dose f Activity Released \ xi .23/J\x/E '
\ ) \ .25/"} \ mev/dis /
Calculation with Proposed Technical Specification valve closure time Release Detection 30 seconds to purge inlet 30 seconds to,Purge inlet 1.4 seconds to Isolation valve 30 seconds for release to transit sample line 31.4 seconds 30 seconds for valve closure (by T. S.)
90 seconds Release continues for (90 - 31.4 seconds) = 58.6 seconds prior to isolation Case 2 with actual valve closure time of 20 seconds Release Detection 31.4 seconds 80 seconds Release continues for 48.6 Sec.
Total Whole Body Dose 1. 68 rem Total Thyroid Dose 28.31 rem
\
e Page 9 of 11
( CASE 3 If none of the radiation monitors capable of the containment isolation is functional or if a selective power failure occurs at the semi-vital bus, a.11 of the radioactivity released from the reactor cavity fuel pool is assumed to be released to the environment through the filters. This assumes no manual securing of the purge systemo The releases and site doses are as follows:
Isotope Activity Whole Body(f) Whole Bodyy.s) Thyroid Released Inorganic I 131 1.00 + 02 lo08 + 02 132 1.42 - 11 5.54 - 13 133 lo21 + 01 3.53 + 00 134 lo56 - 32 2.84 - 34 135 9. 76 ~ 03 8.82 - 04 111.5 r~;
Organic I 131 3.35 + 01 3.61 + 01 132 :+.75 -12 1. 85 - 13 133 4.04+00 1.18 + 00 134 5.20 33 9.47 - 35 135 3.26 - 03 2.95 - 04 37~28 rem
, nc- 1" 'l nl, , ,
Ki. G.Ju1 l.
- OJ - J..~ 4. 87 - le '-'* V'""t' .- .I.. I 85m 4.07 - 03 3o33 - 07 lo98 - 06 85 1.82 + 03 2o00 - 03 2.00 - 01 87 4.03 - 19 1.81 - 22 1.97 - 22 88 1.25 - 06 1.31 - 09 1.85 - 10 89 Xe 131m 4.50 + 01 5.20 - 04 2. 93 - 03 133m 1.,00 + 03 1.25 - 02 9.23 - 02 133 8.53 + 04 3.31 + 00 5.20 + 00 135m 135 8.97 + 00 1.17 - 03 1.31 - 03 137 138 3.33 rem 5.50 rem I 131 2.64 - 02 1.35 - 02 132 2.28 - 14 3 86 - 15 0
133 5. 38 - 03 3.13 - 03 134 2.74 - 35 5.61 - 36 135 9.96 - 06 2.98 - 06 3.18 - 02 rem 1.66 - 02 rem Total 8.88 rem 14-8.,8 rem Calculation Method for Activity Released Activity of Assy ) ( 1 ) * (Fraction Re.leased ) 1-Filter eff. )
( Curies X \ DF X to Cavity )C.. ( 0.3 Iodine
. . .0075<>1.0 0.1..: > 0.3 1. O All orhers
e Page 10 of 11 Case 4 It is assumed that the worst case purge isolation exists (i.e., Case 2) and the filters have zero percent efficiency or the exhaust duct down stream of the isolation valves fails and the purge rate remains constant. The releases would be Case 2 modified by:
The contribution by Kr and Xe would remain unchanged with a whole body 'f9 of 9.63 - 01 rem and a& 1.59 + 00 rem. The Iodine contribution to whole body would be: __ 9.18 - 03 rem = -r 3.06 - 02 rem q 0.3
- 4. 82 - 03 rem
= 1.61 - 02 rem 1-A 0 3 0
Totals "f = 9.936 01 Whole body= 2.60 rem Total Thyroid Dose 43.12 :cem 143.7 rem 0.3
.,,\.,
Page 11 of 11 Conclusion Dose at Site Boundary Rem Whole Bod-y: Thyroid Case 1 (Manipulator Crane monitor triggers isolation) 0 0 Case 2 (Containment gas or containment particulate triggers isolation) Tech. Spec. Valve 2.04 34.38 Actual Valve Closure 1. 68 28.31 Case 3 (No containment isolation but 70% filter) 8,88 148.8 Case 4 (Worst case containment isolation and 0% filter) 2.60 14307 10CFR100 Guidelines 25 300
( \
\
/ .
3.10 REFUELING Applicability I Applies to _operating limitations during refueling operations.
Objective To assure that no accident could occur during refueling operatio~s that would affect public health and safety.
( Specification A. During refueling operations the folloking conditions are satisf.iecl:
I
- 1. The equipment door and at least one door in the personnel air
- 1ock- shall be properly closed;. For those systems. wh:i.ch provide a direct path from containment atmosphere to the -outside atmosphere, all automatic containment isolation valves,in the unit shall be operable or at least one . valve shall be. closed in each.line*
penetrating the containment.
- 2. The Containment Vent and Purge System;and the area and airborne radiation monitors which initiate isolation of this system, .shall
(
be tested and verified to be operable innnediately prior to refueling operati~ns1 CVrief(
~- a
~~ ~i.
.Lh,;,,
k7 F' -~ ~~ J,,_
r -L &P<<4< M ,,.,:t~
I to~.
9-13-76 b.
e .e /
The process vent gas monitor and particulate monitor shall be
. { operating.
- 6. All effluents to be discharged to the atmosphere from the waste gas decay tanks of the gaseous waste disposal system shall be sampled and analy~ed to demonstrate compliance with specification B-1 and B-3 138 above prior to releases via the process vent.
- 7. During periods* of primary to secondary leakage, the alar,Jl setpoint of< 1.3 µCi/cc will be based on actual isotopic content of samples obtained and analyzed on the multichannel analyzer.
- 8. Whenever the air ejector discharge monitor is inoperable and the steam generator blowdown monitors indicate a primary to secondary leak, the automatic divert feature shall be defeated and samples shall be taken from the air ejector discharge and analyzed from gross activity on a daily basis. If the gross activity reaches
.(
flow shall be manually diverted to the containment.
- 9. The maximum activity to be contained in one gas decay tank shall not exceed 95,400 curies equivalent of Xenon 133.
- 10. Purging of the containment shall be governed by the following conditions:
- a. Containment purge shall be fil.tered through the high efficiency particulate air filters and charcoal absorbers whenever the co~centration of iodine and particulate isotopes exceed the occupational MPG inside the containment.
- b. Containment purge shall be filtered through the high efficiency particulate air filters and charcoal absorbers whenever irradiated I. fuel is being handled or any object is being handled over irradiated Amendment No , 2L,
(*
' . e fuel in the containment. a.~~
TS 3.11-5 3-17-72 Basis*,
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The releases of radioactive materials will be kept as low as practicable as required by 10 CFR 50 and will not exceed the concentration limits specified in 10 CFR 20. At the same time, the licensee js permitted the flexibility of operation, compatible with considerations bf health and safety, ta assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases in excess of four percent of the conce*n tration limits specified in 10 CFR 20. However, all releases must be kept within the concentration limits specified in 10 CFR 20. It is ex-pected that using this operational flexibility under unusual operating conditions, the licensee shall exert every effort to keep levels of radioactive materials C:
released from the. pJant ;:iR h1w as prartfr.ahle anr1 th::it ;:inn11;ii. rplP.;:ii:;pi:; will nn1-exceed a small fraction of the annual average concentration limits specified in 10 CFR 20.
The limiting conditions* for operation contained in specification A-3 _above, which relates to the total number of curies which may be released in liquid effluents in any year, is based on the expected perfonnance of the Surry Power Station assuming both units are operating with 0.25 percent leaking fuel and each unit is experiencing a 20 gallon per day primary to secondary system leak rate.
The formula prescribed in specification B-1 takes atmospheric dilution into account and assures that at the point of maximum ground concentration at the site boundary, the req~irements of 10 CFR 20 will not be exceeded. The limit
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is based on the highest annual avera~~ value of ?(/Q which will occur at the