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Results
Other: ML20137H372, ML20197B076, ML20204G924, ML20205T170, ML20206B329, ML20206B459, ML20206F887, ML20207K386, ML20207K441, ML20207K446, ML20207K506, ML20207K512, ML20207P779, ML20207P991, ML20207P993, ML20209E329, ML20209F187, ML20209G043, ML20210A740, ML20210A748, ML20210A757, ML20210T436, ML20210T655, ML20210T686, ML20211D992, ML20211E058, ML20211E084, ML20211E110, ML20211G583, ML20211N368, ML20212M102, ML20214Q988, ML20214Q998, ML20214S836, ML20214S958, ML20215G796, ML20215H964, ML20215H973, ML20215J855, ML20215J871, ML20215L534, ML20216D570, ML20216E230, ML20234C109, ML20235E520, ML20235F508, ML20245C018
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MONTHYEARML20137H3721985-11-26026 November 1985 Memorandum & Order Granting Extension of 851130 Deadline for Environ Qualification of Electrical Equipment to 860531 & Approving Proposal to Allow Operation w/35% Reactor Power Limit During Interim.Served on 851127 Project stage: Other ML20211E0841986-02-20020 February 1986 Issue a to Fort St Vrain:Delayed Firewater Cooldown;Effect of Liner Cooling on Orifice Valve Temps Project stage: Other ML20209F1871986-03-18018 March 1986 Fort St Vrain Steam Generator Temps During Interruption of Forced Cooling from 105% Power Project stage: Other ML20214S9581986-09-25025 September 1986 Requests That Util Take Listed Actions Re Three Repts Submitted Concerning Firewater Cooldown Analysis,Per NRC Review & 860917 Telcon.Response to Item 1 Requested within 15 Days of Ltr Date & Item 2 at Time Analysis Submitted Project stage: Other ML20211E0581986-09-30030 September 1986 Effect of Delayed Firewater Cooldown W/Loss of Liner Cooling on Pcrv Temps Project stage: Other ML20215G7961986-10-10010 October 1986 Advises That Info Requested in NRC Re Best Estimate Schedule for Resolution of Problems Encountered W/Firewater Cooldown Analysis Will Be Submitted as Part of Util Graduated rise-to-power Program Project stage: Other ML20197B0761986-10-22022 October 1986 Informs That Util Will Update & Submit Rept on Chernobyl Accident by 861126.Update Will Ctr on Graphite Related Concerns,Including Analysis of Worst Case Explosive Gas Mixtures & Comparison of Reactor Kinetics Behavior Project stage: Other ML20211G5831986-10-22022 October 1986 Anticipates Completion of Steam Generator Analysis & App R Modeling Reanalysis Work by Feb 1987,per 860918 Telcon W/Nrc Re Steam Generator Cool Down Studies for App R Project stage: Other ML20215L5341986-10-23023 October 1986 Staff Requirements Memo Re Commission Briefing in Washington,Dc on Status of Facility Project stage: Other ML20207K5121986-11-13013 November 1986 Fort St Vrain Calculations for Circulator Temp-Related Operating Limits Project stage: Other ML20207K5011986-12-0404 December 1986 Effect of Firewater Cooldown Using Economizer-Evaporator- Superheater (EES) Bundle on Steam Generator Structural Integrity. Draft Rept of Steam Generator Ability to Withstand post-App R Firewater Cooldown Transient Encl Project stage: Draft Other ML20207K4461986-12-12012 December 1986 Issue a to Effect of Firewater Cooldown Using Reheater on Steam Generator Structural Integrity Project stage: Other ML20211N3681986-12-12012 December 1986 Forwards Restart Interaction Schedule,Per 861205 Request Project stage: Other ML20207K5061986-12-22022 December 1986 Issue a to Effect of Intentional Depressurization on Cooldown from 39% Power Using One Reheater Module (1-1/2 H Delay) Project stage: Other ML20207K4411986-12-23023 December 1986 Issue a to Economizer-Evaporator-Superheater (EES) Cooldown from 39% & 78% Power Using Condensate or Firewater (1.5 H Delay) Project stage: Other ML20207K3861986-12-30030 December 1986 Forwards Analyses Supporting Power Operation Up to 39% Power Based on Safe Shutdown Cooling Following 90 Min Interruption of Forced Circulation.Conclusions of Repts Listed.Corrective Actions for LERs 86-020 & 86-026 Also Listed Project stage: Other ML20207P7791987-01-0707 January 1987 Forwards Current Integrated Schedule for Restart & Power Ascension Activities.Schedule Incorporates Consolidated Schedular Info on Both Interaction Activities.Updates Will Be Provided Twice Per Month.W/One Oversize Graph Project stage: Other ML20207P9931987-01-13013 January 1987 SAR for Tech Spec Limiting Condition for Operation 4.3.1 Change Permitting Safe Shutdown Cooling W/Evaporator- Economizer-Superheater Project stage: Other ML20207P9911987-01-15015 January 1987 Proposed Tech Specs,Requiring Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power as Min Number of HXs Project stage: Other ML20207P9891987-01-15015 January 1987 Application for Amend to License DPR-34,requiring Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power as Min Number of Operable HXs Project stage: Request ML20207P9871987-01-15015 January 1987 Forwards Application for Amend to License DPR-34,changing Tech Specs to Require Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power.Fee Paid Project stage: Request ML20211E1101987-01-26026 January 1987 Rev a to Engineering Evaluation of Procedure to Recover from Actuation of Steam Line Rupture Detection/Isolation Sys for Power Levels Through P2 Project stage: Other IR 05000267/19870021987-01-30030 January 1987 Partially Withheld Insp Rept 50-267/87-02 on 870106-09 (Ref 10CFR73.21).No Violations or Deviations Noted.Major Areas Inspected:Matl Control & Accounting Project stage: Request ML20210A7571987-01-30030 January 1987 Fort St Vrain 1987 Power Ascension Plan Project stage: Other ML20210A7481987-01-30030 January 1987 Requests Concurrence to Start Up & Operate Facility Through Graduated Rise to Power Up to 100% of Rated Power,Subj to Listed Constraints. Fort St Vrain 1987 Power Ascension Plan Encl Project stage: Other ML20210A7401987-02-0202 February 1987 Forwards Updated Nrc/Public Svc Co of Colorado Restart Interaction Schedule, Reflecting Current Target Dates & Recently Completed Items Project stage: Other ML20209G0431987-02-0202 February 1987 Forwards Current Integrated Schedule for Plant Restart & Power Ascension Activities.W/One Oversize Encl Project stage: Other ML20210N8831987-02-0303 February 1987 Forwards Request for Addl Info on 861230 & 870115 Submittals Re Analysis of Firewater Cooldown from 82% of Full Power Project stage: RAI ML20210P0191987-02-0505 February 1987 Summary of 870113 Meeting W/Util Re Completion of Equipment Qualification Program & Program & Approvals Required for Plant Restart Project stage: Meeting ML20210T6861987-02-0505 February 1987 Rev a to Engineering Evaluation of Reanalysis of FSAR Accidents/Transients Relying on EES Cooling. W/Four Oversize Drawings Project stage: Other ML20211D9921987-02-0505 February 1987 Issue a to Economizer-Evaporator-Superheater Cooldowns for Equipment Qualification & App R Events W/Vent Lines (1.5 H Delay) Project stage: Other ML20210T6551987-02-0606 February 1987 Provides Results of Confirmatory Analyses for FSAR Accidents Which Utilize Either EES or Reheater Section of Steam Generator for DHR Project stage: Other ML20210T4361987-02-11011 February 1987 Requests Publication of Fr Notice of Consideration of Issuance of Amend to License DPR-34 & Proposed NSHC Determination & Opportunity for Hearing on 870115 Request Re Operation of evaporator-economizer-superheater Sections Project stage: Other ML20211E9791987-02-12012 February 1987 Forwards Proposed Agenda & Slides for 870226 Meeting W/ Commission & Staff to Secure Commission Approval for Full Power Operation of Facility Project stage: Meeting ML20211D8901987-02-17017 February 1987 Forwards Response to NRC 870203 Request for Addl Info Re Firewater Cooldown from 82% of Full Power,Per Util 861230 & s Project stage: Request ML20207Q7941987-03-0303 March 1987 Forwards Second Request for Addl Info Re Util Analysis of Firewater Cooldown from 82% of Full Power Operation,Based on Review of 861230,870115 & 0217 Submittals Project stage: Approval ML20212M1021987-03-0505 March 1987 Forwards Assessment of Addl Concerns Per 870226 Request Re Restart of Facility.Events & Circumstances Involving Contracts Had No Adverse Impacts Re Plant Safety or Safe Plant Operations Project stage: Other ML20204G9241987-03-20020 March 1987 Forwards Restart & Power Ascension Schedule,Incorporating Consolidated Schedular Info on NRC-util Interaction Activities.Brief Narrative Description of Scope of Each Line Item Activity Also Encl.W/One Oversize Encl Project stage: Other ML20205B3441987-03-20020 March 1987 Forwards Response to NRC 870303 Second Request for Addl Info Re Firewater Cooldown from 82% of Full Power (Safe Shutdown Cooling) Project stage: Request ML20205M8901987-03-30030 March 1987 Forwards Third Request for Addl Info Re Util 861230,870115 & 0217 Submittals Concerning Analysis of Firewater Cooldown from 82% of Full Power.Major Concerns Re Effects of Transient Loading Due to Seismic Motion or Flow Project stage: RAI ML20205T1701987-04-0101 April 1987 Forwards Oversize Current Integrated Schedule for Facility Restart & Power Ascension Activities Required for Equipment Qualification Completion Certification,Startup/Plant Criticality & Power Ascension to 82%.Related Info Encl Project stage: Other ML20206B6031987-04-0101 April 1987 Forwards Comments Re Implication of Chernobyl Reactor Accident.Design Differences Between Fort St Vrain & Chernobyl Preclude Accident Similar to Chernobyl from Occurring at Fort St Vrain Project stage: Approval ML20206B4591987-04-0303 April 1987 Forwards Summary of Equipment Qualification (EQ) Insp Conducted by NRR & IE on 870126-30.EQ Program Approved. Detailed Results of Insp Will Be Provided Project stage: Other ML20206B3291987-04-0707 April 1987 Submits Daily Highlight.Public Svc Co of Colorado Authorized to Restart & Operate Facility HTGR at Level of Up to 35% Full Power.Facility Out of Operation Since 860531,when Shut Down for Equipment Qualification Mods Project stage: Other ML20206F8871987-04-10010 April 1987 Submits Requested Addl Info for Analysis of Firewater Cooldown for 82% Power Operation,Per Project stage: Other ML20209E3291987-04-27027 April 1987 Provides Written Authorization to Operate Reactor at Up to 35% Full Power,Per Section IV of 870406 Confirmatory Order Modifying License DPR-34 Project stage: Other ML20215H9641987-04-30030 April 1987 Forwards Updated Ga Technologies Procedure 909410, Buckle Users Manual, Per 870330 Request.Manual Updated to Include Revs to Computer Code Required by High Temps & Short Times Assumed for Steam Generator Tube Stress Analysis Project stage: Other ML20215H9731987-04-30030 April 1987 Revised Buckle Users Manual:Creep Collapse of Thin-Walled Circular Cylindrical Shells Subj to Radial Pressure & Thermal Gradients Project stage: Other ML20215J8711987-05-0404 May 1987 Rev a to Evaluation of Test Data for Confirmation of Fire Water Flow Rate to Circulator Water Turbine During EES Cooldown for Safe Shutdown Cooling Project stage: Other ML20215J8551987-05-0404 May 1987 Forwards Rev a to EE-EQ-0057, Evaluation of Test Data for Confirmation of Fire Water Flow Rate to Circulator Water Turbine During EES Cooldown for Safe Shutdown Cooling Project stage: Other 1987-01-07
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20205G8771999-03-26026 March 1999 Forwards Copy of Cover Page from Fort St-Vrain Welding Manual,Which Had Been Listed as Encl on Page 4 of 990325 Reply to EA 98-081.Cover Page Had Been Inadvertently Left Out with Original Reply ML20197H8811998-12-0101 December 1998 Forwards Proposed Change to Fsv ISFSI Physical Protection Plan in Which Commitment Is Made to Provide Feature to Security Posture for Facility ML20236R9191998-07-20020 July 1998 Ltr Contract:Mod 4 to Task Order 27, Task Area No 4 of Basic Contract - Fort St Vrain Insp Under Contract NRC-02-95-003 ML20199H8141997-11-21021 November 1997 Responds to Requesting Clarification as to Whether Increase in Tritium & Iron-55 Contamination Limits That Were Approved for Plant Apply to All Licensees ML20198K1931997-10-10010 October 1997 Provides Supplemental Info in Support of Util Proposed Rev to Physical Security Plan for Plant Plant Isfsi.Plan Withheld,Per 10CFR2.790(d) & 10CFR73.21 ML20198H5601997-09-16016 September 1997 Final Response to FOIA Request for Documents.Documents Listed in App a Being Released in Entirety ML20141F3521997-05-14014 May 1997 Forwards Proposed Issue 4 of Physical Security Plan for Fort St Vrain ISFSI for Review & Approval.Encl Withheld,Per 10CFR2.790(d) ML20141C8611997-05-0909 May 1997 Informs of Approval of Fsv Final Survey Rept & Effluent Pathway Survey Plan & Supporting Analysis ML20141K9881997-05-0505 May 1997 Forwards Amend 89 to License DPR-34 & Supporting Safety Evaluation.Amend Designates All Elements of Approved Decommissioning Plan as License Termination Plan ML20138G2701997-04-28028 April 1997 Provides Response to NRC Comments Re Proposed Sampling & Survey Plan for Fsv Effluent Pathway.Response Documents Fsv Liquid Effluent Discharge Pathway Areas Are Acceptable for Release for Unrestricted Use IAW Draft NUREG/CR-5849 ML20148D4651997-04-24024 April 1997 Forwards Revised Interim Ltr Rept Which Describes Procedures & Results of Confirmatory Survey of Group E Effluent Discharge Pathway Areas at Fsv Station NUREG/CR-5849, Requests That Licensee Provide Evidence That Average Contamination Levels in Group E Effluent Discharge Pathway Areas Meet Averaging Criteria in Draft NUREG/CR-58491997-04-23023 April 1997 Requests That Licensee Provide Evidence That Average Contamination Levels in Group E Effluent Discharge Pathway Areas Meet Averaging Criteria in Draft NUREG/CR-5849 ML20138B0511997-04-22022 April 1997 Forwards Copy of Proposed Amend to Fsv NPDES Permit, Wastewater Discharge Permit CO-0001121 Requested to Support Repowering Activities,Iaw Section 3.2.d of Fsv Non-Radiological Ts,App B to License DPR-34 ML20140E1061997-04-10010 April 1997 Forwards Confirmatory Survey of Group Effluent Discharge Pathway Areas for Fsv Nuclear Station,Platteville,Co ML20137S4481997-04-0808 April 1997 Informs That Decommissioning Activities at Fsv Are Complete & NRC Issued Exemption from Requirements of 10CFR50.54(w) in .Property Damage Insurance Policy Is Maintaned to Protect Fsv balance-of-plant Assets ML20137S0821997-04-0707 April 1997 Forwards Insp Rept 50-267/97-01 on 970310-11.No Violations Noted ML20137S1691997-04-0707 April 1997 Fifth Partial Response to FOIA Request for Documents. Forwards Documents Listed in App K ML20137S5421997-04-0707 April 1997 Forwards Final Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Fort St Vrain Nuclear Station ML20148D5951997-04-0404 April 1997 Forwards Confirmatory Survey for Fsv Nuclear Station, Psc,Platteville,Co, Final Rept ML20137R6921997-04-0404 April 1997 Informs of Approval for Request for Addl 45 Days to Remedy Deficiencies Identified in NRC Re Financial Assurance Mechanism for Fsv Decommissioning Costs ML20137J8051997-03-31031 March 1997 Third Partial Response to FOIA Request for Documents.Records in App F Encl & Will Be Available in Pdr.App G & H Records Withheld in Part (Ref FOIA Exemptions 5 & 7) ML20148D5811997-03-26026 March 1997 Forwards Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station, Covered in Final Survey Rept,Vol 6 ML20137G7361997-03-25025 March 1997 Requests Addl Time for Util to Respond to NRC Comments in Re Financial Assurance Mechanism for Fort St Vrain Decommissioning Costs ML20137G9521997-03-24024 March 1997 Forwards Quarterly 10CFR50.59 Rept for Period 961201-970228 Re Changes,Tests & Experiments for Fort St Vrain Decommissioning ML20137H1131997-03-24024 March 1997 Second Partial Response to FOIA Request for Documents. Forwards Documents Listed in App D.Documents Also Available in Pdr.Documents Listed in App E Withheld in Part (Ref FOIA Exemption 6) ML20137C0181997-03-18018 March 1997 Documents That No Personnel Has Received Radiation Exposure at Fsv in 1997 or at Any Time Subsequent to ML20137C0061997-03-18018 March 1997 Documents That There Have Been No Activities Involving Release of Radioactive Matls from Fsv Nuclear Station That Potentially Could Have Affected Environ,Subsequent to Previous Radiological Envion Operating Rept ML20136G1201997-03-11011 March 1997 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1996 & Jan-Mar 1997. All Effluent Releases Completed as of 960703.Repts on Activities After 960703 Reflect Disposal of Solid Waste ML20136B1331997-02-28028 February 1997 First Partial Response to FOIA Request for Documents. Documents Listed in App a Already Available in Pdr.Forwards App B Documents.App C Documents Being Withheld in Entirety (Ref FOIA Exemption 5) ML20135D7891997-02-27027 February 1997 Forwards Responses to Comments Re Fort St Vrain Final Survey Rept ML20135D9531997-02-27027 February 1997 Forwards Copy of Amend to Util Npdes,Wastewater Discharge Permit CO-0001121,which Clarifies That Monitoring of Farm Pond Outlet Required When Industrial Wastewater Being Discharged Through Upstream Goosequill Ditch ML20135A8711997-02-14014 February 1997 Requests That Encl Deficiencies Identified in Financial Assurance Mechanism for Fort St Vrain Decommissioning Cost Be Addressed within 45 Days ML20134D1551997-01-31031 January 1997 Forwards Util Responses to NRC Comments Provided in NRC Ltr Re Sampling & Survey Plan Used for Final Radiological Survey of Liquid Effluent Pathway at Ft St Vrain ML20134C8481997-01-30030 January 1997 Forwards Draft Confirmatory Survey Rept for Fsv Nuclear Station,Psc,Platteville,Co Providing Info on Essap Activities on 960930-1003 ML20133L4961997-01-0707 January 1997 Forwards Comments That Need to Be Resolved Before Final Approval of Util Submittal Entitled, Proposed Sampling & Survey Plan for Effluent Pathway,Ft St Vrain Final Survey Program ML20133E0481997-01-0202 January 1997 Forwards Comments to Fsv Nuclear Station, Decommissioning Project Final Survey Rept (Volumes 4-11), for Consideration ML20132G0421996-12-23023 December 1996 Forwards Insp Rept 50-267/96-05 on 961203-05.No Violations Noted ML20132F2841996-12-19019 December 1996 Forwards Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Plant,Covering Period of 960901-1130 ML20133A8591996-12-16016 December 1996 Forwards Original & Copy Transcripts of Public Hearing,Held on 961203 in Platteville,Co Re Decommissioning & License Termination of Util Ft Saint Vrain Nuclear Generating Station ML20133N0011996-12-0404 December 1996 Recommends That NRC Require License to Modify Submission of Unexecuted Draft Trust Agreement Remaining Decommissioning Costs for Ft St Vrain Nuclear Generating Station in Listed Ways ML20135B3861996-11-25025 November 1996 Informs That NRC Reviewed Util 961114 Submittal (P-96096) Entitled, Fort St Vrain Final Emergency Response Plan, & Meets Requirements of 10CFR50.54(q) ML20135A5861996-11-25025 November 1996 Submits Suppl Info Re Annual Environ Rept for 1995 Operation of Fsv ISFSI ML20135A6361996-11-20020 November 1996 Submits Copy of Describing Discharge Practices for Groundwater Seeping Into Fsv'S Reactor Building Sump ML20134L4721996-11-14014 November 1996 Notifies NRC That Util Adopted Fsv ISFSI Emergency Response Plan to Direct Emergency Response for Radiological Accidents Occuring at Site,Until 10CFR50 License Is Terminated ML20134F4351996-10-30030 October 1996 Forwards Sections 1,2,6 & 8 from Survey Packages F0015, F0039 & F0126 & Sections 1,2 & 6 from Survey Package F0115 to Support on-site NRC Insp ML20134G5991996-10-30030 October 1996 Forwards Volumes 1-12 to Final Survey Rept for Groups A,B,C Rev 1,D Rev 1,E,F Rev 1 & G-J for NRC Approval in Support of Forthcoming Request for Termination of Fsv 10CFR50 License ML20133D7691996-10-22022 October 1996 Forwards Preliminary Rept Re Orise Support of NRC License Insp at Fsv on 960930-1003 ML20136B1411996-10-15015 October 1996 FOIA Request for Documents Re NOV Addressed to Scientific Ecology Group Re NRC Insp Rept 50-267/94-03 & OI Investigation Repts 4-94-010 & 4-95-015 ML20128M6181996-10-0404 October 1996 Forwards Ltr from PSC to Co Dept of Public Health & Environ Describing Monitoring Practices at Plant ML20128G8041996-10-0101 October 1996 Forwards Fsv Decommissioning Fire Protection Plan Update 1999-03-26
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20205G8771999-03-26026 March 1999 Forwards Copy of Cover Page from Fort St-Vrain Welding Manual,Which Had Been Listed as Encl on Page 4 of 990325 Reply to EA 98-081.Cover Page Had Been Inadvertently Left Out with Original Reply ML20197H8811998-12-0101 December 1998 Forwards Proposed Change to Fsv ISFSI Physical Protection Plan in Which Commitment Is Made to Provide Feature to Security Posture for Facility ML20198K1931997-10-10010 October 1997 Provides Supplemental Info in Support of Util Proposed Rev to Physical Security Plan for Plant Plant Isfsi.Plan Withheld,Per 10CFR2.790(d) & 10CFR73.21 ML20141F3521997-05-14014 May 1997 Forwards Proposed Issue 4 of Physical Security Plan for Fort St Vrain ISFSI for Review & Approval.Encl Withheld,Per 10CFR2.790(d) ML20138G2701997-04-28028 April 1997 Provides Response to NRC Comments Re Proposed Sampling & Survey Plan for Fsv Effluent Pathway.Response Documents Fsv Liquid Effluent Discharge Pathway Areas Are Acceptable for Release for Unrestricted Use IAW Draft NUREG/CR-5849 ML20148D4651997-04-24024 April 1997 Forwards Revised Interim Ltr Rept Which Describes Procedures & Results of Confirmatory Survey of Group E Effluent Discharge Pathway Areas at Fsv Station ML20138B0511997-04-22022 April 1997 Forwards Copy of Proposed Amend to Fsv NPDES Permit, Wastewater Discharge Permit CO-0001121 Requested to Support Repowering Activities,Iaw Section 3.2.d of Fsv Non-Radiological Ts,App B to License DPR-34 ML20140E1061997-04-10010 April 1997 Forwards Confirmatory Survey of Group Effluent Discharge Pathway Areas for Fsv Nuclear Station,Platteville,Co ML20137S4481997-04-0808 April 1997 Informs That Decommissioning Activities at Fsv Are Complete & NRC Issued Exemption from Requirements of 10CFR50.54(w) in .Property Damage Insurance Policy Is Maintaned to Protect Fsv balance-of-plant Assets ML20137S5421997-04-0707 April 1997 Forwards Final Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Fort St Vrain Nuclear Station ML20148D5951997-04-0404 April 1997 Forwards Confirmatory Survey for Fsv Nuclear Station, Psc,Platteville,Co, Final Rept ML20148D5811997-03-26026 March 1997 Forwards Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station, Covered in Final Survey Rept,Vol 6 ML20137G7361997-03-25025 March 1997 Requests Addl Time for Util to Respond to NRC Comments in Re Financial Assurance Mechanism for Fort St Vrain Decommissioning Costs ML20137G9521997-03-24024 March 1997 Forwards Quarterly 10CFR50.59 Rept for Period 961201-970228 Re Changes,Tests & Experiments for Fort St Vrain Decommissioning ML20137C0061997-03-18018 March 1997 Documents That There Have Been No Activities Involving Release of Radioactive Matls from Fsv Nuclear Station That Potentially Could Have Affected Environ,Subsequent to Previous Radiological Envion Operating Rept ML20137C0181997-03-18018 March 1997 Documents That No Personnel Has Received Radiation Exposure at Fsv in 1997 or at Any Time Subsequent to ML20136G1201997-03-11011 March 1997 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1996 & Jan-Mar 1997. All Effluent Releases Completed as of 960703.Repts on Activities After 960703 Reflect Disposal of Solid Waste ML20135D7891997-02-27027 February 1997 Forwards Responses to Comments Re Fort St Vrain Final Survey Rept ML20135D9531997-02-27027 February 1997 Forwards Copy of Amend to Util Npdes,Wastewater Discharge Permit CO-0001121,which Clarifies That Monitoring of Farm Pond Outlet Required When Industrial Wastewater Being Discharged Through Upstream Goosequill Ditch ML20134D1551997-01-31031 January 1997 Forwards Util Responses to NRC Comments Provided in NRC Ltr Re Sampling & Survey Plan Used for Final Radiological Survey of Liquid Effluent Pathway at Ft St Vrain ML20134C8481997-01-30030 January 1997 Forwards Draft Confirmatory Survey Rept for Fsv Nuclear Station,Psc,Platteville,Co Providing Info on Essap Activities on 960930-1003 ML20133E0481997-01-0202 January 1997 Forwards Comments to Fsv Nuclear Station, Decommissioning Project Final Survey Rept (Volumes 4-11), for Consideration ML20132F2841996-12-19019 December 1996 Forwards Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Plant,Covering Period of 960901-1130 ML20133A8591996-12-16016 December 1996 Forwards Original & Copy Transcripts of Public Hearing,Held on 961203 in Platteville,Co Re Decommissioning & License Termination of Util Ft Saint Vrain Nuclear Generating Station ML20133N0011996-12-0404 December 1996 Recommends That NRC Require License to Modify Submission of Unexecuted Draft Trust Agreement Remaining Decommissioning Costs for Ft St Vrain Nuclear Generating Station in Listed Ways ML20135A5861996-11-25025 November 1996 Submits Suppl Info Re Annual Environ Rept for 1995 Operation of Fsv ISFSI ML20135A6361996-11-20020 November 1996 Submits Copy of Describing Discharge Practices for Groundwater Seeping Into Fsv'S Reactor Building Sump ML20134L4721996-11-14014 November 1996 Notifies NRC That Util Adopted Fsv ISFSI Emergency Response Plan to Direct Emergency Response for Radiological Accidents Occuring at Site,Until 10CFR50 License Is Terminated ML20134F4351996-10-30030 October 1996 Forwards Sections 1,2,6 & 8 from Survey Packages F0015, F0039 & F0126 & Sections 1,2 & 6 from Survey Package F0115 to Support on-site NRC Insp ML20134G5991996-10-30030 October 1996 Forwards Volumes 1-12 to Final Survey Rept for Groups A,B,C Rev 1,D Rev 1,E,F Rev 1 & G-J for NRC Approval in Support of Forthcoming Request for Termination of Fsv 10CFR50 License ML20133D7691996-10-22022 October 1996 Forwards Preliminary Rept Re Orise Support of NRC License Insp at Fsv on 960930-1003 ML20136B1411996-10-15015 October 1996 FOIA Request for Documents Re NOV Addressed to Scientific Ecology Group Re NRC Insp Rept 50-267/94-03 & OI Investigation Repts 4-94-010 & 4-95-015 ML20128M6181996-10-0404 October 1996 Forwards Ltr from PSC to Co Dept of Public Health & Environ Describing Monitoring Practices at Plant ML20128G8041996-10-0101 October 1996 Forwards Fsv Decommissioning Fire Protection Plan Update ML20128G0481996-09-30030 September 1996 Submits Rev to Psco Definitions of Contents of Documentation Packages Re Fsv Final Survey Project ML20129C0421996-09-20020 September 1996 Forwards Quarterly Submittal of 10CFR50.59 Rept of Changes, Tests & Experiments for Facility Decommissioning,Covering Period of 960601-0831 ML20133D7601996-09-16016 September 1996 Forwards Confirmatory Survey Plan for Fsv Nuclear Station Decommissioning Project,First Final Survey Rept Submittal- Vols 1-5.NRC Comments Incorporated.Spending Plan Attached ML20117P0711996-09-13013 September 1996 Describes Util Plans to Remove Bldg 28 from Plant Facility ML20129A4431996-09-11011 September 1996 Describes Util Plans for Demonstrating That Liquid Effluent Pathway & Surrounding Open Land Areas Satisfy 10 Mrem/Yr Criteria Provided in Plant Final Survey Plan ML20117K5291996-09-0404 September 1996 Provides Notification That Util Will Be Revising Financial Assurance Mechanism That Will Be Used to Cover Remaining Costs of Decommissioning Plant ML20117C7281996-08-22022 August 1996 Discusses Impact of Final Decommissioning Rule & Requests NRC Concurrence That Requirements to Submit & Obtain Approval of License Termination Plan Have Been Satisfied ML20116P3431996-08-16016 August 1996 Describes Actions to Remove Structures & Equipment Items from Fort St Vrain Facility for NRC Info.Requests That NRC Advise Util of Wishes to Perform Confirmatory Survey of Any Parts of New Fuel Storage Building Before 960903 ML20133D7551996-08-14014 August 1996 Provides Environ Survey & Site Assessment Program'S (Essap) Comments Re Review of Fsv Nuclear Station Decommissioning Project Final Survey Rept ML20116M0771996-08-14014 August 1996 Provides Suppl Response to Re Insp Rept 50-267/96-01 in Jan 1996 Re NRC Concerns About Fsv Final Survey Program.Specifically,Bias in Instrumentation Response Overestimating Amount of Contamination Present ML20116M1841996-08-13013 August 1996 Forwards Util Responses to NRC Comments in Re Use of in-situ Gamma Spectroscopy to Measure Exposure Rates During Plant Final Survey.Approval to Use in-situ Gamma Spectroscopic instrument,Microspec-2,requested ML20116K0061996-08-0909 August 1996 Submits Fort St Vrain Nuclear Station Decommissioning Project Final Survey Rept ML20116M1241996-08-0808 August 1996 Responds to NRC Bulletin 96-004, Chemical,Galvanic,Or Other Reactions in Spent Fuel Storage & Transportation. Informs That Modular Vault Dry Storage Sys Is Not Susceptible to Problems Addressed in Bulletin ML20116F3611996-08-0202 August 1996 Submits Revised Documentation for Fort St Vrain Final Survey Program ML20116F8141996-08-0202 August 1996 Informs of Util Intent to Modify Fort St Vrain Control Room,Which Will Make Certain Final Survey Locations Unavailable for Further Review.Final Survey Efforts Are Complete ML20116A4511996-07-19019 July 1996 Requests NRC Approval of Proposed Method to Fsv Final Survey Plan to Determine Exposure Rates in Prestressed Concrete Reactor Vessel 1999-03-26
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059H9471990-09-14014 September 1990 Forwards Rev to Proposed Physical Security Plan,Per .Encl Withheld (Ref 10CFR73.21) ML20064A6841990-09-14014 September 1990 Notifies That Ms Frazier License OP-5608-1 Terminated Effective 900906,per 10CFR50.74 ML20059K2091990-09-14014 September 1990 Requests That Dj Trumblee Reactor Operator License Be Terminated,Effective 900912 ML20059L5831990-09-14014 September 1990 Forwards Application for Amend to License DPR-34,changing Tech Spec Design Features Section to Remove CRD & Orifice Assemblies from Core Regions Defueled in Support of Plant Closure Activities ML20059D3231990-08-15015 August 1990 Forwards fitness-for-duty Performance Data Rept for Plant ML20058N5661990-08-0808 August 1990 Requests Exemption from 10CFR50.54(w)(1) Requirements for Min Property Insurance Coverage ML20058L8031990-08-0303 August 1990 Forwards GA-C18103, Fort St Vrain Cycle 3 Core Performance, Per Request for Addl Info Re Proposed Amend to Tech Spec Which Would Allow Util to Complete Defueling of Facility ML20058M1931990-07-31031 July 1990 Advises of Termination of Kj Einig Reactor Operator License, Effective 900727 ML20056A1791990-07-30030 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-267/90-07.Corrective Actions:Electrical & Mechanical Craft Personnel Assigned to Planning & Scheduling Organization ML20055H8091990-07-23023 July 1990 Forwards Addl Info in Support of 900622 SAR for Isfsi. Proprietary SAR Withheld (Ref 10CFR2.790) ML20056A1711990-07-20020 July 1990 Discusses Decommissioning Funding for Plant.Util Revised Decommissioning Financial Plan,Based on Safstor,Currently in Effect & Funds Will Be Accumulated Per Plan Provisions ML20055H6571990-07-20020 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Model 1153 or 1154 Transmitters Presently in Use at Plant ML20058P6891990-07-20020 July 1990 Forwards Security Plan for Proposed ISFSI for NRC Review,Per 900622 & 0412 Ltrs.Plan Withheld (Ref 10CFR73.21) ML20055G8601990-07-20020 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-267/90-07.Corrective Actions:Site Mgt Recognized Applicability of 10CFR50.72(b)(2)(vi) Re Previous Day Notification & Immediately Notified NRC ML20055H2811990-07-20020 July 1990 Forwards Rev 4 to Updated Fire Protection Plan ML20055G7061990-07-20020 July 1990 Forwards Application for Amend to License DPR-34,revising Tech Spec 6.1 to Remove Equipment to Support Plant Closure ML20059D1361990-07-13013 July 1990 Applies for Exemption from Payment of Annual Fees for FY90, Per 10CFR171 ML20055E0401990-07-0303 July 1990 Forwards Issue 3 to FPOR-18, Fire Protection Operability Requirements, as Result of Rev 3 to Facility Fire Protection Plan ML20055E0331990-07-0202 July 1990 Informs of Plans to Remove B Helium Circulator from Plant Prestressed Concrete Reactor Vessel in Mannner Similar to Previous Circulator Maint Outages.Requests Timely Ack of NRC Concurrence W/Undertaking ML20055E9171990-06-29029 June 1990 Resubmits Semiannual Radioactive Effluent Release Rept Jul-Dec 1989, as Result of Omitted Tables ML20044A8281990-06-29029 June 1990 Notifies of Termination of Wj Ashmore,License SOP-43244 & Requests That Reactor Operator License Be Terminated Effective 900620,per 10CFR50.74 ML20055D3931990-06-29029 June 1990 Forwards Response to Generic Ltr 90-04,requesting Info on Status of Generic Safety Issues.W/O Encl ML20055D1591990-06-22022 June 1990 Forwards Proposed Change to Updated Fsar,App B Re Use of Controls Other than Change Notice Sys to Implement Mods Such as Cutting & Capping of Piping,In Order to Obtain Positive Equipment Isolation ML20043J0581990-06-15015 June 1990 Forwards Proposed Issue 5 to Fort St Vrain Nuclear Generating Station Defueling Emergency Response Plan, Deleting Offsite Emergency Response Capabilities,Per Util 890816 Defueling SAR & NUREG-0654 ML20043D3291990-05-25025 May 1990 Forwards Response to NRC 900330 Ltr Re Violations Noted in Insp Rept 50-267/90-04.Encl Withheld (Ref 10CFR73.21) ML20055D2051990-05-25025 May 1990 Responds to Violations Noted in Insp Rept 50-267/90-06. Corrective Actions:Extensive Training on Radiological Emergency Response Plan Implementing Procedures Will Be Provided to Operating Crews Beginning in May 1990 ML20043C8981990-05-25025 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-267/90-05.Corrective Actions:Mgt Will Continue to Stress Requirement for Procedural Compliance on Ongoing Basis ML20043B3691990-05-22022 May 1990 Describes Util Current Plans for Removal of Certain Plant Equipment Items.Components to Be Removed Have No Required Nor Useful Function During Planned or Postulated Remaining Defueling or Shutdown Conditions ML20043A9491990-05-11011 May 1990 Forwards Issue 59 to AOP-K-1, Environ Disturbances - Earthquake. ML20042F9771990-05-0404 May 1990 Forwards Assessment of Possible Sources of Water in Plant Core,Means of Detecting Water in Pcrv & Effects of Water Ingress on Reactivity,For Info ML20042F9011990-05-0101 May 1990 Forwards Conceptual Plan & Cost Estimates for Early Dismantlement of Fort St Vrain Pcrv, Per NRC 900315 Request ML20042E8911990-04-27027 April 1990 Requests 30-day Extension to 900529 to Submit Response to Violations Noted in Insp Rept 50-267/90-04 ML20042F3111990-04-26026 April 1990 Forwards Application to Amend License DPR-34,providing Tech Spec Changes Needed to Complete Defueling ML20012D4711990-03-19019 March 1990 Requests Approval of Encl Exemption Request Allowing Util to Accumulate Decommissioning Funds Beyond Termination of Operations ML20012B7541990-03-0707 March 1990 Informs of Electronic Transfer of Fees on 900307 for Invoice IO547 ML20011F4691990-02-22022 February 1990 Responds to NRC 900116 Ltr Re Violations Noted in Insp Rept 50-267/89-23.Violation Re Fire Doors 13 & 9 Between 480-volt Ac Switchgear Room & Bldg 10 Propped Open on 891129 Disputed.Improved Communications Underway ML20006E7821990-02-15015 February 1990 Forwards Responses to 891004 Questions Re Decommissioning Financial Plan ML20011E8941990-02-0707 February 1990 Responds to 890719 NRC Bulletin 89-002 Re safety-related Anchor/Darling Model S350W Swing Check Valves.Licensee Reviewed Plant Data Base & Design Documents & Concludes That No Anchor/Darling Model S350W Valves Installed at Plant ML20011E7861990-02-0707 February 1990 Forwards Overview of Evaluation on Compliance W/Environ Qualification Regulation (10CFR50.49) During Defueling of Facility.During Defueling,Pcrv Will Be Maintained at Nearly Atmospheric Pressures & Rapid Depressurization Not Possible ML20011E5031990-02-0606 February 1990 Notifies of Termination of Ma Rhoton Employment W/Util, Effective 900206 ML20006D6621990-02-0606 February 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Currently Established Preventive Maint & Tech Spec Surveillance Activities on HXs Remaining in Svc Will Continue ML20006B2091990-01-26026 January 1990 Forwards Fire Operability Requirements 3 & 14,Issues 3 & 4, Respectively,To Fire Protection Program Plan,Section FP.6.1. Changes Cover Halon Sys & Basis & Records Ctr Halon Sys ML20006B7791990-01-25025 January 1990 Forwards Application for Amend to License DPR-34,changing Section 7.1 of Tech Specs.Specifically,Senior Vice President,Nuclear Operations Revised to Vice President, Nuclear Operations & Nuclear Training Manager Revised ML20006A5001990-01-19019 January 1990 Notifies of Util Intent to Bore 10 Concrete Core Drills Into Prestressed Concrete Reactor Vessel Barrel Section at Various Elevations Found Necessary for Decommissioning Planning.Drillings Wil Begin on 900205.Assessment Encl ML20006A5041990-01-19019 January 1990 Responds to NRC 891218 Ltr Re Weaknesses Noted in Insp Rept 50-267/89-22.Corrective Actions:During Power Operations Three Licensed Reactor Operators Will Be Present in Control & Short in Emergency Public Address Sys Repaired ML20005G5001990-01-12012 January 1990 Forwards Approved Proposed Change to NPDES Permit to Use Halocide Product for Microbiological Control in Smaller Cooling Tower ML19354D8581990-01-0505 January 1990 Forwards Safety Assessment of Proposed Changes to Fire Protection Program Plan,Per P Erickson Request.Changes Cover Acceptance Criteria for Shutdown Cooling Following Fire & New Fire Protection Cooldown Trains 1 & 2 ML20005F5531990-01-0303 January 1990 Certifies That Util Has Implemented fitness-for-duty Program Meeting 10CFR26 Requirements ML20005D8921989-12-21021 December 1989 Forwards Proposed Change to Plant NPDES Permit ML19354D7441989-12-0808 December 1989 Requests Relief from Requirements of Generic Ltr 89-10 Re motor-operated Valves Since Plant Permanently Shut Down Since 890818 & Defueling Initiated on 891127 1990-09-14
[Table view] |
Text
_
h Public Service- =TA .
2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 March 20, 1987 Fort St. Vrain Unit No. 1 P-87110 U. S. Nuclear. Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267
SUBJECT:
Additional Information for Analysis of Firewater Cooldown for 82% Power Operation
REFERENCE:
- 1) NRC letter Heitner to Williams, dated February 3, 1987(G-87031)
- 2) PSC letter Warembourg to Berkow, dated December 30, 1986(P-86683)
- 3) PSC letter Williams to Berkow, dated January 15, 1987 (P-87002)
- 4) PSC letter Brey to Berkow, dated February 17, 1987 (P-87055)
- 5) NRC letter Heitner to Williams, dated March 3, 1987(G-87060)
Dear Mr. Berkow:
Reference 1 requested that PSC provide additional information concerning the firewater cooldown from 82% of full power in addition to that which was presented by PSC in References 2 and 3. PSC provided responses to NRC's first request for additional information in Reference 4.
O h o 00$367 D "5 1
,p
P-87110 Paga 2 March.20, 1987 Attachment 1 provides PSC's response to the NRC's second request for information presented in Reference 5. This additional information is provided to. enable the NRC to complete its review of PSC's analysis of a firewater cooldown from 82% of full power.
If you have any questions about the responses in Attachment 1, please contact Mr. M.'H. Holmes at (303) 480-6960.
Very truly_yours,
Nuclear Licensing and Fuels Division HLB /AHW:jmt Attachment cc: Regional Administrator, Region IV Attn: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain
[-
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^ -. Attachment-
- to P-87110'
< -Page 1-
,. ; NRC-SECOND~ REQUEST FORl ADDITIONAL INFORMATION
- -DOCKET NO. 50-267:
NRC QUESTION 1 I . What is'~the status of.the hot and cold reheat header isolation. valves-during a firewater-cooldown?
PSC-RESPONSE'1 Firewater cooldown is postulated.following a High-Energy Line Break
's (HELB), an Appendix R fire, an Earthquake, or a Tornado. After.
+
- firewater- has-been aligned for cooldown and subcooled firewater flow , 5 L ' has been established at the outlet of a steam generator, _ forced circulation Jof helium through the reactor core is. recovered with one a circulatorf driven -by. firewater supplied to its water turbine.
Whenever reheat- . steam is not used for driving a helium circulators' l steam. turbine, its . reheat steam ~ (inlet and outlet) isolation -valves.
are normally closed, per System Operating Procedure S0P 21-01. .In -
~
, the event of a loop shutdown, the ' circulators' cold reheat steam
. header . isolation valves,; bypass block- valve and the hot reheat steam
~
t'
' ~
~stop check' valve are automatically closed, effectively isolating that-loop's reheater,
!- Actuation of SLRDIS' following a HELB would.close the reheat steam isolation valves on both loops. Manual action by the operators is
- required to reopen the valves, but the operators would keep these L - valves closed during a-firewater cooldown since they are not in the firewater flow pa th . .. If the HELB were'in the reheat steam piping, l the operators would prevent venting the remaining reheat steam
' inventory into the Reactor Building through the ruptured pipe by
< -- manually closing the reheat steam isolation valves or assuring that l they remain closed after being closed automatically by SLRDIS.
[ - For an Appendix R fire, seismic event, or tornado the status of the valves would (without operator action) be either open or closed
- depending. on the severity or location of the accident and the resulting automatic-actions of the plant control systems.
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~
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, Page 2, . .
-NRC-QUESTION 2:
~
- Are'thevalvesditherclosedoroperableforautomaticclosureshouldo l La leak develop in~ the reheater
- tubes?-
i PSC RESPONSE'2 _
Asi discusse'd in PSC Response ~ 5, it has been determined th'at the.
reheaterJtubes' will not rupture in the event of a firewater cooldown
. . following _ a ' 90.. minute . interruption of forcedJ circulation after
. freactor operation at 105% power. s
<If .aJ leak occurs in the reheater tubes of:a steam generator during-
' f~ ~ ~
normal operation, the primary coolant helium in the.PCRV . would . leak into the ' reheat steam piping-system, which is at a pressure lower
~
-than -the; primary- coolant pressure. Radioactive gases and particu.lates~ within the ' primary coolant > could then pass =into the (J
s . - reheat steam. piping system andtthen.out to the environment. via -the main condenser.-
^ '
. IR'adiation : detectors monitor. the fluid inside the loop hot reheat- -
3; steam headers::forl indications of. reheater tube ruptures.. Upon detection .of radioactivity above the trip setpoint.in the hot reheat piping,-a signal is sent.via the PPS system to automatically close- -
the reheat _ steam.' isolation valves and the feedwater and main' steam header isolation val.ves, thus completely' isolating the affected loopi The - reheat - steam isolation valves are automatically closed upon ,
actuation of thel SLRDIS following 'a HELB.- The SLRDIS- Recovery Procedure OPOP XII, Checklist Step 2, requires the. plant operators to L .
actuate'all valve hand: switches closed, which were closed-~by- SLRDIS actuation, to assure -they do not reopen when SLRDIS is reset.
Following a' HELB in the Turbine Building the SLRDIS recovery-
! -procedure, 0 POP-XII, Checklist Step 7, requires the plant operators to remove the six hot reheat activity monitor modules, in order to g a .. prevent postulated malfunction of these non-environmentally qualified monitors from causing a loop shutdown, including shutdown of the :
recovered':(operating EES)' loop and thus interferring with SLRDIS recovery.
The reheat steam isolation valves and the PPS hot reheat steam header radiation detectors are seismically qualified and would therefore
-remain operable following an earthquake or tornado. The reheat steam V'
isolation valves would be closed by the radiation detectors in the
- < event of a reheat tube break.
i: ' For- a: fire, the worst condition that could result is inoperable
' reheat steam isolation valves and hot reheat steam header radiation
~ '
detectors. That is, the reheat steam header isolation valves could be assumed to be open. This condition is considered acceptable because the acceptance criteria delineated in PSC Letter Lee to Johnson, dated August 17, 1984 (P-84281), for an Appendix R fire
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Attachment:
, :to.P-87110L '
- Page 3 _
a: . .
- v-outside the Congested Cable Areas:(CCA). include:-
. 21. "For any: single fire-in 'a non-congested cable area means shall be available to: shut'down and cool.down the reactor in a manner.such thath ._no fuel; damage. occurs . :(i .e. maximum : . fuel particle
~ temperature does not exceed 2900 degrees F). There.shall be 'no simultaneous _ rupture ( of both a' primary: coolant boundary.and the f; associated = secondary. containment. boundary.. such that :no
~ unmonitored radiological releases of primary coolant occur."
-2. Thel specific: criteria for. single failures in conjunction with a fire is:-
" Systems: used- to ensure the post fire safe . reactor
-shutdown /cooldown capability need not be designed to meet seismic Category -I : criteria,- single failure criteria, or other.. design basis accident criteria, except where required.for other-reasons,-
e.g., because of -interface with or impact on existing safety systems, or because of adverse valve actions due'to fire damage."- -
Asf discussed .in PSC Response 5, the reheater. tubes will not rupture
-in the event of a cooldown following a fire outside the CCAs; Even though the- secondary. containment boundary could be postulated to be
. breeched-(open reheat steam isolation valves) following.a fire, the-primary coolant boundary-would remain intact, meeting the acceptance-criteria for a fire outside the CCAs.
L i- ,
4 I
L-_ _ .
Attachment to-P-87110 Page 4 NRC QUESTION 3-Are-there initiating events for which this system is not. qualified or does-not meet regulatory requirements, such as an earthquake .or
-single fire?
-PSC RESPONSE 3 The radiation detectors and isolation valves described in PSC~
-Response 2 are seismically qualified. The radiation detectors are not environmentally qualified while the reheat steam isolation valves are environmentally qualified and would .be available for manual isolation in the event of a HELB, if SLRDIS has not already closed these valves. These instruments and valves are not relied upon as part~.of the ~ Appendix R . fire protection model. As discussed in PSC Response 5, it has been demonstrated that the PCRV pressure boundary is maintained during firewater cooldown and that the reheater tubes are not predicted to rupture following a HELB, an Appendix R fire, seismic event, or tornado.
l.
L l
a
. . [ ' .r g A JAttachment
- to P-87110-u, .Page 5.
NRC QUESTION 4 .
What .safetyE benefits .would come -from closing 'the valves before
- restoring helium circulation? '
PSC RESPONSE'4 Closing the: reheat = steam isolation valves 'following an accident-requiring the alignment.of a firewater- cooldown, if not already closed; by.' automatic action, would. in' the event'of a subsequent
. - reheater tube rupture result'in containing the. primary coolant helium i ;in the'PCRV since the closed reheat steam isolation valves would form a secondary pressure boundary.
Closing the reheat steam isolation valves would establish 'two.
redundant barriers to prevent release of primary coolant._ Reheater tube ruptures during the cooldown with' firewater were evaluated and are- not expected to occur (see PSC Response 5). Because'of the many-operator actions occurring during the first few hours of cooldown on firewater, requiring Lthe . operators to close the reheati steam
. isolation valves during .this period ~might' introduce unnecessary complications.
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7 IAttachment
. Sto:P-87110- -
~
. Page 6; ,
..f '
LNRC QUESTION 5'
-What? sa'fety benefits. would come from.being able to pressurize the reheater with dry gas'before restoring helium circulation?-
PSC RESPONSE 5 Reversing' the: pressure differential 'across the reheater tube wall by
. pressurizing the reheater with dry gas would change' tube stresses from compressive to tensile and therefore eliminate the probability.
of tube buckling. But the analysis of the event has'shown that the- s
<- likelihood.of the reheater tube buckling is practically non-existant.
The reheater tube buckling ' analysis,- reported in GA~ Technologies 3
Report GA-909204,~- Issue N/C,.and submitted to-the-NRC as Attachment'7' ,
of PSC-letter.P-87055, utilized a maximum temperature of 1660 degrees
- F 'with a maximum . pressure differential of 350 psi- for a period of approximately two hours. This_ analysis concluded that'the potential
- for tube'. buckling arises when these temperature and pressure
' conditions are maintained for more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. This ' analysis is conservative since the: model assumes tube buckling occurs'when a single element of.the tube reaches yield stress. The 1660 degrees: F.
- and 350 psi. conditions- are representative of a 105 percent power
. case. The calculated maximum tube temperature at the 87.5 percent
- power . level is reduced to 1501 degrees F. From Table 4.2 of the GA-909204 r? port, the time for the reheater tube to reach: buckling at 1500 . degrees F using the maximum external-pressure used for the
~
- . buckling calculation of 832 psizis 196 hours0.00227 days <br />0.0544 hours <br />3.240741e-4 weeks <br />7.4578e-5 months <br />. As - the . temperature of 1500 degrees F is not expected to be exceeded for more than two hours-and the external pressure is expected to be less than 700 psi at any ;
time 'during the event, there is a significant margin of safety. It r
is noted that for the Appendix R cases (83.2% power level) reported -
in GA Technologies Report GA-909269, Issue A, this. temperature has -
been calculated to be slightly less than 1500 degrees F.
The' added complication of introducing the pressurized dry gas to the reheater seems to be an unnecessary precaution. Aside from any complications that a dry gas pressurizing system might introduce to
-the many actions that are already- required of the operators, the-design and ' operation of such a system may introduce more problems than the benefits it might provide. For instance, overpressurization
,' of a hotter than average reheater tube by the dry gas system could -
-result ~fn rupture of this tube.
l 4
Attachment to P-87110 Page 7 NRC QUESTION 6 Page 14 of Attachment 1 to P-87055 refers to 'the results of special testing of simulated firewater' supply to the steam generators. and circulator Pelton drives. The text of Unusual Event Report 76/05A indicates that Test T-30 yielded only 540 gpm supply to the steam generator. What was the circulator speed achieved during this test?
What were the steam generator supply flow and circulator speed during Test T-30A? Test RT-403 achieved circulator speeds in excess of 700 RPM on each circulator. What was the steam generator supply flow during RT-403? What circulator speeds and concurrent steam generator flow rates were obtained during any testing after installation of the j emergency water booster pumps? What were the simulated firewater L supply pressure (booster pump suction), steam generator back l pressure, and Pelton wheel supply pressure during each of these tests for which data are provided?
FSC RESPONSE 6 i
Table 6-1 is a summary of the T-30 test series. Figure 6-1 is the T-30 test flow schematic. In the initial T-30 test "A" circulator only was tested and this was for a low steam generator back pressure /high flow test case. The recorded circulator speed was 636 rpm which did not satisfy the approximate 700 rpm required speed for cold core conditions. Instrumentation upgrades were made and additional testing identified as T-30A were conducted. These are also summarized in Table 6-1. "A" circulator speed remained low for the one test while "B" circulator speed was observed to improve with subsequent tests. It was later concluded based upon RT-403 testing that the panel and data logger instrumentation used for T-30 speed readouts were significantly in error. RT-403 used the speed / wobble scope for circulator speed measurements. '
The T-30 series tests identified excessive pressure drop in valves HV-21257 through HV-21260 which are located in the flow path to the water turbine drives. These valves were subsequently modified to decrease pressure losses. Testing conducted in RT-403 evaluated circulator performance on water turbine drive on "A" and "B" circulators before the valve modifications and on all circulators after the valve modifications. The results of RT-403 testing are summarized in Table 6-2. The RT-403 flow schematic is shown in Figure 6-2. The test indicated that the valve pressure losses were i decreased by almost 50 psi down to 5-7 psi and that circulator speeds significantly increased satisfying the FSAR criteria at that time. ,
The criteria for the test was to simulate firewater cooling by I throttling the condensate pressure to maintain 113-115 psig at PI I 31204, which senses the pressure upstream of the Pelton wheel. The steam generator supply flow and other users of condensate and their flow rates during the RT-403 testing were not recorded.
The emergency water booster pumps were installed in 1979 to enhance the performance of the circulators when being driven by the water turbine drive on firewater. The primary requirement of the emergency
l' .
Attachment
-to P-87110-LPage 8
- water booster. pumps was'to assure a circulator performance when_being driven..by the water turbine drive .at least equivalent to that attainable with 175 psid.at the water turbine nozzle. -Typically this:
1-will provide. a primary coolant helium flow varying between 3 to 4%
under hot-- core ' accident - conditions.
~
r- r - Since . installation of the '
emergency . . water booster- pumps .all~ testing. of the water turbine-
~
Ldrives is performed to the requirements of Technical Specification SR 5.2.7.a. See. response 'to NRC-Question.7 for discussion of testing ~
. performed since the . emergency water.-booster pumps have been installed. .
.Upon review of the test data from Tests T-30 and.T-30A, PSC concludes that while the dataiis generally in' the range .to support recent analyses,' it: is not valid to rely on-the data for direct support of' these analyses cInstead, PSC proposes that a special test be run to-provide accurate and sufficient data, as described below.
'The tests T 30 and T-30A were conducted so'long ago that the results do not warrant current detailed evaluation. Since the time of these.
tests, several valves (HV-21257 through HV-21260).have been modified-to decrease excessive'-pressure drops. . Addltional.ly, the Emergency _
Water Booster Pumps .have been installed and then enlarged, and the.
new 6" vents-have been installed to allow: increased flow through 'the steam generators.
Therefore,- PSC considers a more prudent review path to be the running of.a special test of simulated firewater. to a circulator Pelton-drive. This test' is discussed in PSC Response 7. 'Any testing of firewater flow rates through the steam generators would still be an
, approximate simulation requiring significant analysis to correlate to actual conditions. . Therefore, PSC would still rely on the existing analysis to' verify those flow rates.
7
- ,u
., 9 Table 6-li 1.-
' Concurrent Circulator Speed and Steam Generator Testing Performed in T-30. Test. Series 1
Test T-30'6/2/76(2) T-30A6/16/76:(3)
Test Data PV-22129 Closed PV-22129 Closed. PV-22129 Open PV-2229 Open PV-2229 Open PV-2229 Open- '
Test Point Number 1 2 3 4 1 2 3 4 5 6 7- 8 Steam Generator 770 770 270 640 640- 640 640 690 700 --
'780. --
Loop Flow,(l) 103 lbm/hr ,
Steam Generator -- -- -- -- 1210 1235 1235 1320 1340 1360 1425- 1500 Loop Flow, gpm(l)
Emergency Cond. 115- 125 305 290 125 125 113 115 122 137 115 .'135 Header Pressure at PI 31204.
Elv. 4819, psig Steam Generator 160 75 75 0-25 58 60 60 64 63 50 51- --
Exhaust Pressure at PI-22129, psig Circulator Pelton -- -- -- 136 137. 138. 139 151 161 140 1501 160 Nozzle Supply Pressure at Test Gauge Elv. 4770, psig (same as Test Gauge 2 in RT-403 Testing)
Circulator -- -- --
636 0' 600 625' 666 750 731 '750 790 Speed, rpm "A" "A" "B" "B" "B"- "B" "B" "B" Notes (1) Two measurements available for T-30A '
(2) Helium temperature was 1030F and helium pressure'286 psia (3) Helium temperature was 1280F and helium pressure 334 psia-
Table 6-2 .
RT-403 Simulated. Firewater Circulator Speed Testing Prior HV21257 through Test HV21260 Valve Modifications (l) AfterValveModifica4jQns 10/3/76-and 10/5/76', 1 9/28/76 Test Gauge 1 114 115 114 114. 114' 114-psig (same location as PI 31204 in T-30 series tests)
Test Gauge 2, 136 136 133 131.5 -- --
psig Test Gauge 3, .78 81 128 124.5 -- --
psig Test Gauge 4, 74 86 130 124.5 -. --
1 psig j Circulator Speed, 630 650 760 790 790 ,770 rpm from "A" .
"B" "A" "B" "C" "D" speed scope Notes
- 1. Helium temperature was ll70F and helium pressure was 305 psia.
- 2. Helium temperature was 1190F and helium pressure was 338 psia for testing of "A" circulator..
Helium temperature was 1220F and helium pressure was 333 psia for testing of "B", "C", and "D" circulators.
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N' Attachinent to-P-87110
. -Page 9: <-
<NRC QUESTION 7 SR 5.2.7.2' is' indicated to..be -performed with 115 psig suction pressure at'the emergency booster pump from throttled- condensate.-.
' Provide .the results Lof SR 5.2.7.2 in -tems of booster pump suction pressure,L -booster: pump differential ' pressure, circulator-.. speed, simulated -firewater flow rates from the booster pump and to the'
. Pelton wheel, .and primary system flow at rated helium. inventory.- Has
-SR: 5.'2.7.2 been' performed in conjunction with simulated firewater flow to the steam generator? If so, provide ~the above results plus the steam generator flow and back pressure measurements.
PSC RESPONSE 7 SR_- 5.2.7.2 should: be SR 5.2.7.a. -Surveillance test SR 5.2.7.a has been replaced.with SR 5.2.7.a2-A effectiv'e November 30, 1984, after.
the Emergency Water Booster -Pumps .were upgraded. The following.
results are taken from SR 5.2.7.a2-A, conducted July 18, 1985:
WATER TURBINE USING CONDENSATE AT REDUCED PRESSURE TO SIMULATE
-FIRE PUMP PRESSURE ON LOOP I CIRCULATOR-P-2109 Booster Pump Suction Pressure - . . . . . . . . . . . . . . . . . . y. . . . . . . . . . . 115 psig Ci rcu l a to r' S peed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2100 rpm -
P-2110 Booster Pump-Suction Pressure .............................. 115 psig Circulator Speed............................................ 2200 rpm
?
WATER-TURBINE USING CONDENSATE AT REDUCED PRESSURE TO SIMULATE FIRE PUMP PRESSURE-0N LOOP II CIRCULATOR' ,
P-2109:
' Booster Pump Suction Pressure .............................. 113 psig Circulator Speed............................................ 2100 rpm P-2110
, Booster Pump Suction Pressure .............................. 114 psig i Circulator Speed............................................ 2000 rpm This surveillance test procedure does not require measurement of the Booster Pump differential pressure, the flow rate of the simulated firewater to the Pelton wheel, or primary system flow rate. During the performance of this test, the circulator inlet helium temperature i was 120 degrees F, and the reactor pressure was 64.3 psia. From this
- a helium density.of .04 lb/ cubic foot was calculated. A minimum
- . acceptable circulator speed of 1825 rpm was determined based on this l calculated density, and the circulator being driven by the water
! turbine drive with 175 psid across the water turbine nozzle. During
[ the test, the circulator speed was increased until an acceptable l speed above this minimum speed was attained. This is the speed noted l above.
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l._.. . _ .~ . _ . _ _ _ _ , _ _ _ _ . _ , _ _ _ , - , , _ _ _ _ _ _ . _ _ _ _ _ , , _ _ . . _ _ . . , _ . . _ _ _ . , _ _ _ _ _ , _ _ _ _ _
Attachment to P-87110 Page.10 SR 5.2.7a2-A has not'.been performed in conjunction with simulated firewater flow to the steam generators. Further, if this test had been performed the results would not be applicable -to the new firewater flow paths.
While SRS.2.7a2-A measures helium circulator speed with the simulated.
firewater supplied via the booster pumps, the helium flow rates and simulated firewater flow rates are not routinely determined.
Calculations were recently performed to determine the helium and simulated firewater flow rates involved in the July 18, 1985 test of the circulators. These calculated flows were within the range of the P2 shutdown analysis. However, the only way to verify the actual flow rates is to perform a special test, similar to SR 5.2.7.a2-A, with an increased amount of data being recorded. PSC will perform this test within the next few ' weeks and submit this additional information to the NRC in answer.to this question.
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