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Category:CORRESPONDENCE-LETTERS
MONTHYEARML18039A9021999-10-15015 October 1999 Forwards LER 99-010-00 Re Occurrence of Plant Reactor Scram Due to Main Turbine Trip Which Resulted in Main Steam Moisture Separator.All Plant Safety Systems Operated as Designed in Response to Event ML20217E0711999-10-14014 October 1999 Grants Approval for Util to Submit Original,One Signed Paper Copy & Six CD-ROM Copies of Updates to FSAR as Listed,Per 10CFR50.4(c),in Response to ML18039A8961999-10-14014 October 1999 Forwards LER 99-009-00,re Manual Reactor Scram on Unit 2 from 54% Power,Iaw 10CFR50.73(a)(2)(iv).All Plant Safety Sys Operated as Designed in Response to Event ML20217D3261999-10-0808 October 1999 Responds to Re Event Concerning Spent Fuel Pool Water Temperature Being Undetected for Approx Two Days at Browns Ferry Unit 3 ML20217F7751999-10-0808 October 1999 Confirms 991006 Telcon Between T Abney of Licensee Staff & a Belisle of NRC Re Meeting to Be Conducted on 991109 in Atlanta,Ga to Discuss Various Maintenance Issues ML18039A8931999-10-0808 October 1999 Forwards LER 99-008-00,concerning HPCI Sys Being Declared Inoperable,Iaw 10CFR50.73(a)(2)(v).There Are No Commitments Contained in Ltr ML18039A8881999-10-0808 October 1999 Provides Licensee Supplemental Response to NRC 980713 RAI Re GL 87-02,Suppl 1, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors. ML20217B5481999-10-0101 October 1999 Requests Exception to 10CFR50.4(c) Requirement to Provide Total of Twelve Paper Copies When Submitting Revs to BFN UFSAR ML20212M1481999-09-28028 September 1999 Refers to Management Meeting Conducted on 990927 at Region II for Presentation of Recent Plant Performance.List of Attendees & Copy of Presentation Handout Encl ML20212F7751999-09-22022 September 1999 Requests Operator & Senior Operator License Renewals for Listed Individuals and Licenses ML20212D3651999-09-20020 September 1999 Forwards SE Accepting Licensee 990430 Proposed Rev to Plant, Unit 3 Matl Surveillance Program ML18039A8721999-09-10010 September 1999 Informs of Licensee Decision to Withdraw Proposed Plant risk-informed Inservice Insp Program,Originally Transmitted in Util 981023 Ltr.Licensee Expects to Resubmit Revised Program within Approx 6 Wks ML20211Q5731999-09-0909 September 1999 Submits Response to Administrative Ltr 99-03 Re Preparation & Scheduling of Operator Licensing Exams.Completed NRC Form 536,operator Licensing Exam Data,Which Provides Plant Current Schedules for Specific Info Requested Encl ML20211G6491999-08-26026 August 1999 Confirms Telcon with T Abney on 990824 Re Mgt Meeting Which Has Been re-scheduled from 990830-0927.Purpose of Meeting to Discuss BFN Status & Performance ML20210Q6931999-08-0909 August 1999 Forwards Updated Changes to Distribution Lists for Browns Ferry & Bellefonte Nuclear Plants ML18039A8371999-08-0606 August 1999 Forwards BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts, for NRC Review.Corrected Inservice Insp Summary Rept for Unit 3 Cycle 8 Operation,Included in Rept ML20210Q4421999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006. Authorized Representative of Facility Must Submit Ltr with List of Individuals to Take exam,30 Days Before Exam Date ML20210N1051999-08-0202 August 1999 Forwards SE Accepting Licensee 990326 Request for Relief from ASME B&PV Code,Section XI Requirements.Request for Relief 3-ISI-7,pertains to Second 10-year Interval ISI for Plant,Unit 3 ML20210G8991999-07-28028 July 1999 Discusses 990726 Open Mgt Meeting for Discussion on Plant Engineering Status & Performance.List of Attendees & Presentation Handout Encl ML18039A8181999-07-26026 July 1999 Forwards LER 99-004-00 Re Inoperability of Two Divisions of Plant CSS Due to Personnel Error During Surveillance Testing.Event Reported Per 10CFR50.73(a)(2)(i)(B) ML20210G8051999-07-22022 July 1999 Discusses DOL Case DC Smith Vs TVA Investigation.Oi Concluded That There Was Not Sufficient Evidence Developed During Investigation to Substantiate Discrimination.Nrc Providing Results of OI Investigation to Parties ML20210F3031999-07-22022 July 1999 Submits Rept Re Impact of Changes or Errors in Methodology Used to Demonstrate Compliance with ECCS Requirements of 10CFR50.46.One Reportable non-significant Error Was Found During Time Period of 980601-990630 ML20209J0251999-07-16016 July 1999 Forwards SE Which Constitutes Staff Review & Approval of TVA Ampacity Derating Test & Analyses for Thermo-Lag Fire Barrier Configurations as Required in App K of Draft Temporary Instruction, Fpfi, ML20210B2671999-07-14014 July 1999 Confirms 990702 Telcon Between T Abney of Licensee Staff & Author Re Mgt Meeting Scheduled for 990830 at Licensee Request in Atlanta,Ga to Discuss Browns Ferry Nuclear Plant Status & Performance ML20209E3421999-07-0707 July 1999 Confirms Arrangements Made During 990628 Telephone Conversation to Hold Meeting on 990726 in Atlanta,Ga to Discuss Plant Engineering Status & Performance ML20209E5511999-07-0707 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 Rai,Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2.This Closes TACs MA1180,MA1181 & MA1179 ML20196J3531999-06-30030 June 1999 Responds to Re Boeing Rocket Booster Mfg Facility Being Constructed in Decatur,Al.Nrc Has No Unique Emergency Planning Concerns Re Proximity of Boeing Facility to BFN ML20196G9111999-06-28028 June 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-Related Logic Circuits ML18039A8081999-06-28028 June 1999 Forwards LER 99-004-00 Re Esfas That Occurred When RPS Motor Generator Tripped.Rept Is Submitted IAW Provisions of 10CFR50.73(a)(2)(iv) as Event of Condition That Resulted in Automatic Actuation of ESF ML18039A8111999-06-25025 June 1999 Requests Permanent Relief from Inservice Insp Requirements of 10CFR50.55a(g) for Volumetric Exam of Bfn,Unit 3 Circumferential RPV Welds,Per GL 98-05 ML20196F8741999-06-23023 June 1999 Forwards Safety Evaluation Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20196F8131999-06-22022 June 1999 Forwards Rev 24 to Security Personnel Training & Qualification Plan,Per 10CFR50.54(p).Rev Withheld ML18039A8051999-06-22022 June 1999 Forwards LER 99-003-00,re Automatic Reactor Scram Due to Turbine Trip.Rept Numbered 99-001 Should Be Deleted & Replaced with Encl Rept as Result of Error Noted in 990614 Rept ML18039A8031999-06-18018 June 1999 Responds to NRC Staff Verbal Request Re TS Change TS-376, Originally Submitted on 970312, & Proposed Changes to TS to Extend Current 7-day AOT for EDGs to 14 Days ML18039A7931999-06-0101 June 1999 Provides Summary of Major Activities Performed at BFN During Scheduled Unit 2 Cycle 10 Refueling Outage ML20195D3321999-06-0101 June 1999 Informs That Cb Fisher,License OP-5525-4,can No Longer Maintain License at Plant Because of Physical Condition That Causes Licensee to Fail to Meet Requirements of 10CFR55.21 ML18039A7911999-05-24024 May 1999 Informs That by Meeting Test Criteria Established by Test Based on Ansi/Ans 3.5-1985 (License Amends 254 & 214) power- Uprate Simulation Acceptable for Operator Training ML18039A7891999-05-24024 May 1999 Informs That Oscillation Power Range Monitor Module Has Been Enabled for Current Cycle of Operation Following Unit 2 Cycle 10 Refueling Outage Which Was Completed on 990509 ML20195B9361999-05-24024 May 1999 Informs That Do Elkins,License SOP-3392-6,no Longer Needs to Maintain License as Position Does Not Require License ML20206U6551999-05-14014 May 1999 Informs That ML Meek & Wd Dawson Will No Longer Need to Maintain SRO Licenses at Plant,Due to Termination of Employment,Effective 990521 ML20206Q8421999-05-10010 May 1999 Forwards Medical Info on DM Olive,License SOP-20540-2,in Response to NRC 990428 Telcon.Encl Withheld from Public Disclosure IAW 10CFR2.790(a)(6) ML18039A7771999-05-0606 May 1999 Forwards LER 99-003-00,providing Details Re Plant HPCI Sys Being Declared Inoperable Due to Loose Electrical Connection.Ltr Contains No Commitments ML20206G6611999-05-0404 May 1999 Forwards SE Accepting GL 88-20,submitted by TVA Re multi-unit Probabilistic Risk Assessement (Mupra) for Plant, Units 1,2 & 3 ML18039A7741999-04-30030 April 1999 Forwards Proposed Rev to BFN Unit 3 RPV Matl Surveillance Program,For NRC Approval ML20206H5901999-04-30030 April 1999 Forwards Notification of Revs to BFN Unit 2 Emergency Response Data Sys Data Point Library.Revs Were Implemented on 990413 DD-99-06, Informs That Time Provided by NRC within Which Commission May Act to Review Director'S Decision (DD-99-06) Has Expired.Decision Became Final Agency Action on 990423.With Certificate of Svc.Served on 9904281999-04-28028 April 1999 Informs That Time Provided by NRC within Which Commission May Act to Review Director'S Decision (DD-99-06) Has Expired.Decision Became Final Agency Action on 990423.With Certificate of Svc.Served on 990428 ML18039A7681999-04-27027 April 1999 Requests Relief from Specified Inservice Insp Requirements in Section XI of ASME Boiler & Pressure Vessel Code,Per 10CFR50.55a(a)(3)(i).Relief Requests 2-ISI-8 & 3-ISI-8,encl for NRC Review & Approval ML18039A7591999-04-27027 April 1999 Forwards Annual Radiological Environ Operating Rept Browns Ferry Nuclear Plant 1998. Rept Includes Results of Land Use Censuses,Summarized & Tabulated Results of Radiological Environ Samples in Format of Reg Guide 4.8 & NUREG-1302 ML18039A7651999-04-27027 April 1999 Forwards Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2,Cycle 11 Colr. ML18039A7541999-04-23023 April 1999 Requests Approval of Bfnp Unit 3 Risk-Informed ISI (RI-ISI) Program,Per 10CFR50.55(a)(3)(i) & GL 88-01.Encl RI-ISI Program Is Alternative to Current ASME Section XI ISI Requirments for Code Class 1,2 & 3 Piping 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML18039A9021999-10-15015 October 1999 Forwards LER 99-010-00 Re Occurrence of Plant Reactor Scram Due to Main Turbine Trip Which Resulted in Main Steam Moisture Separator.All Plant Safety Systems Operated as Designed in Response to Event ML18039A8961999-10-14014 October 1999 Forwards LER 99-009-00,re Manual Reactor Scram on Unit 2 from 54% Power,Iaw 10CFR50.73(a)(2)(iv).All Plant Safety Sys Operated as Designed in Response to Event ML18039A8881999-10-0808 October 1999 Provides Licensee Supplemental Response to NRC 980713 RAI Re GL 87-02,Suppl 1, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors. ML18039A8931999-10-0808 October 1999 Forwards LER 99-008-00,concerning HPCI Sys Being Declared Inoperable,Iaw 10CFR50.73(a)(2)(v).There Are No Commitments Contained in Ltr ML20217B5481999-10-0101 October 1999 Requests Exception to 10CFR50.4(c) Requirement to Provide Total of Twelve Paper Copies When Submitting Revs to BFN UFSAR ML20212F7751999-09-22022 September 1999 Requests Operator & Senior Operator License Renewals for Listed Individuals and Licenses ML18039A8721999-09-10010 September 1999 Informs of Licensee Decision to Withdraw Proposed Plant risk-informed Inservice Insp Program,Originally Transmitted in Util 981023 Ltr.Licensee Expects to Resubmit Revised Program within Approx 6 Wks ML20211Q5731999-09-0909 September 1999 Submits Response to Administrative Ltr 99-03 Re Preparation & Scheduling of Operator Licensing Exams.Completed NRC Form 536,operator Licensing Exam Data,Which Provides Plant Current Schedules for Specific Info Requested Encl ML20210Q6931999-08-0909 August 1999 Forwards Updated Changes to Distribution Lists for Browns Ferry & Bellefonte Nuclear Plants ML18039A8371999-08-0606 August 1999 Forwards BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts, for NRC Review.Corrected Inservice Insp Summary Rept for Unit 3 Cycle 8 Operation,Included in Rept ML18039A8181999-07-26026 July 1999 Forwards LER 99-004-00 Re Inoperability of Two Divisions of Plant CSS Due to Personnel Error During Surveillance Testing.Event Reported Per 10CFR50.73(a)(2)(i)(B) ML20210F3031999-07-22022 July 1999 Submits Rept Re Impact of Changes or Errors in Methodology Used to Demonstrate Compliance with ECCS Requirements of 10CFR50.46.One Reportable non-significant Error Was Found During Time Period of 980601-990630 ML18039A8081999-06-28028 June 1999 Forwards LER 99-004-00 Re Esfas That Occurred When RPS Motor Generator Tripped.Rept Is Submitted IAW Provisions of 10CFR50.73(a)(2)(iv) as Event of Condition That Resulted in Automatic Actuation of ESF ML18039A8111999-06-25025 June 1999 Requests Permanent Relief from Inservice Insp Requirements of 10CFR50.55a(g) for Volumetric Exam of Bfn,Unit 3 Circumferential RPV Welds,Per GL 98-05 ML20196F8131999-06-22022 June 1999 Forwards Rev 24 to Security Personnel Training & Qualification Plan,Per 10CFR50.54(p).Rev Withheld ML18039A8051999-06-22022 June 1999 Forwards LER 99-003-00,re Automatic Reactor Scram Due to Turbine Trip.Rept Numbered 99-001 Should Be Deleted & Replaced with Encl Rept as Result of Error Noted in 990614 Rept ML18039A8031999-06-18018 June 1999 Responds to NRC Staff Verbal Request Re TS Change TS-376, Originally Submitted on 970312, & Proposed Changes to TS to Extend Current 7-day AOT for EDGs to 14 Days ML18039A7931999-06-0101 June 1999 Provides Summary of Major Activities Performed at BFN During Scheduled Unit 2 Cycle 10 Refueling Outage ML20195D3321999-06-0101 June 1999 Informs That Cb Fisher,License OP-5525-4,can No Longer Maintain License at Plant Because of Physical Condition That Causes Licensee to Fail to Meet Requirements of 10CFR55.21 ML20195B9361999-05-24024 May 1999 Informs That Do Elkins,License SOP-3392-6,no Longer Needs to Maintain License as Position Does Not Require License ML18039A7911999-05-24024 May 1999 Informs That by Meeting Test Criteria Established by Test Based on Ansi/Ans 3.5-1985 (License Amends 254 & 214) power- Uprate Simulation Acceptable for Operator Training ML18039A7891999-05-24024 May 1999 Informs That Oscillation Power Range Monitor Module Has Been Enabled for Current Cycle of Operation Following Unit 2 Cycle 10 Refueling Outage Which Was Completed on 990509 ML20206U6551999-05-14014 May 1999 Informs That ML Meek & Wd Dawson Will No Longer Need to Maintain SRO Licenses at Plant,Due to Termination of Employment,Effective 990521 ML20206Q8421999-05-10010 May 1999 Forwards Medical Info on DM Olive,License SOP-20540-2,in Response to NRC 990428 Telcon.Encl Withheld from Public Disclosure IAW 10CFR2.790(a)(6) ML18039A7771999-05-0606 May 1999 Forwards LER 99-003-00,providing Details Re Plant HPCI Sys Being Declared Inoperable Due to Loose Electrical Connection.Ltr Contains No Commitments ML20206H5901999-04-30030 April 1999 Forwards Notification of Revs to BFN Unit 2 Emergency Response Data Sys Data Point Library.Revs Were Implemented on 990413 ML18039A7741999-04-30030 April 1999 Forwards Proposed Rev to BFN Unit 3 RPV Matl Surveillance Program,For NRC Approval ML18039A7681999-04-27027 April 1999 Requests Relief from Specified Inservice Insp Requirements in Section XI of ASME Boiler & Pressure Vessel Code,Per 10CFR50.55a(a)(3)(i).Relief Requests 2-ISI-8 & 3-ISI-8,encl for NRC Review & Approval ML18039A7591999-04-27027 April 1999 Forwards Annual Radiological Environ Operating Rept Browns Ferry Nuclear Plant 1998. Rept Includes Results of Land Use Censuses,Summarized & Tabulated Results of Radiological Environ Samples in Format of Reg Guide 4.8 & NUREG-1302 ML18039A7651999-04-27027 April 1999 Forwards Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2,Cycle 11 Colr. ML20206C8591999-04-23023 April 1999 Informs That Util Has Determined,Dr Bateman No Longer Needs to Maintain His License,Effective 990331,per Requirement of 10CFR55.55(a) ML18039A7541999-04-23023 April 1999 Requests Approval of Bfnp Unit 3 Risk-Informed ISI (RI-ISI) Program,Per 10CFR50.55(a)(3)(i) & GL 88-01.Encl RI-ISI Program Is Alternative to Current ASME Section XI ISI Requirments for Code Class 1,2 & 3 Piping ML18039A7581999-04-23023 April 1999 Responds to Item 4 of 981117 RAI Re TS Change Request 376 Re Extended EDG Allowed Outage Time,In Manner Consistent with Rgs 1.174 & 1.177 ML20206C1241999-04-21021 April 1999 Forwards Annual Occupational Radiation Exposure Rept for 1998, IAW TS Section 5.6.1.Rept Reflects Radiation Exposure Data as Tracked by Electronic Dosimeters on Radiation Work Permits ML20205T0971999-04-15015 April 1999 Submits Change in Medical Status for DM Olive in Accordance with 10CFR55.25,effective 990315.Encl Medical Info & Certification of Medical Exam,Considered by Util to Be of Personal Nature & to Be Withheld,Per 10CFR2.790(a)(6) ML18039A7441999-04-0707 April 1999 Forwards LER 99-001-00,providing Details Re Inoperability of Two Trains of Standby Gas Treatment Due to Breaker Trip on One Train in Conjunction with Planned Maint Activities on Other.Ltr Contains No New Commitments ML18039A7431999-03-30030 March 1999 Responds to NRC 990112 RAI Re BFN Program,Per GL 96-05, Periodic Verification of Design-Basis Capability of Safety- Related Movs. ML18039A7421999-03-30030 March 1999 Provides Results of Analysis of Design Basis Loca,As Required by License Condition Re Plants Power Uprate Operating License Amends 254 & 214 ML18039A7411999-03-30030 March 1999 Provides Partial Response to NRC 981117 RAI Re TS Change Request 376,proposing to Extend Current 7 Day AOT for EDG to 14 Days ML18039A7371999-03-26026 March 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME Boiler & Pressure Vessel Code,1989 Edition.Encl Contains Request for Relief 3-ISI-7,for NRC Review & Approval ML18039A7331999-03-26026 March 1999 Forwards Rev 4 to TVA-COLR-BF2C10, Bnfp,Unit 2,Cycle 10 COLR, IAW Requirements of TS 5.6.5.d.COLR Was Revised to Extend Max Allowable Nodal Exposure for GE GE7B Fuel Bundles ML18039A7291999-03-22022 March 1999 Forwards Revised Epips,Including Index,Rev 26A to EPIP-1, Emergency Classification Procedure & Rev 26A to EPIP-5, General Emergency. Rev 26A Includes All Changes Made in Rev 26 as Well as Identified Errors ML20204G8471999-03-19019 March 1999 Reports Change in Medical Status for Ma Morrow,In Accordance with 10CFR55.25.Encl Medical Info & Certification of Medical Exam,Considered by Util to Be of Personal Nature & to Be Withheld from Pdr,Per 10CFR2.790(a)(6).Without Encl ML20207M0611999-03-11011 March 1999 Forwards Goals & Objectives for May 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3,radiological Emergency Plan Exercise.Plant Exercise Is Currently Scheduled for Wk of 990524 ML18039A6971999-02-22022 February 1999 Forwards Typed TS Pages,Reflecting NRC Approved TS Change 354 Requiring Oscillation PRM to Be Integrated Into Approved Power uprate,24-month Operating Cycle & Single Recirculation Loop Operation ML18039A6961999-02-19019 February 1999 Provides Util Response to GL 95-07 Re RCIC Sys Injection Valves (2/3-FCV-71-39) for BFN Units 2 & 3.Previous Responses,Dtd 951215,1016 & 960730,0315 & 0213,supplemented ML18039A6911999-02-19019 February 1999 Forwards Rev 3 to Unit 2 Cycle 10 & Rev 1 to Unit 3 Cycle 9, Colr.Colrs for Each Unit Were Revised to Include OLs Consistent with Single Recirculation Loop Operation ML20203B6031999-02-0404 February 1999 Requests Temporary Partial Exemption from Requirements of 10CFR50.65,maint Rule for Unit 1.Util Requesting Exemption to Resolve Issue Initially Raised in NRC Insp Repts 50-259/97-04,50-260/97-04 & 50-296/97-04,dtd 970521 ML18039A6741999-01-21021 January 1999 Responds to NRC 981209 Ltr Re Violations Noted in Insp Repts 50-259/98-07,50-260/98-07 & 50-296/98-07,respectively. Corrective Actions:Will Revise Procedure NEPD-8 Re Vendor Nonconformance Documentation Submission to TVA ML20199F6951999-01-0808 January 1999 Submits Request for Relief from ASME Section XI Inservice Testing Valve Program to Extend Interval Between Disassembly of Check Valve,Within Group of Four Similar Check Valves for EECW Dgs,From 18 to 24 Months 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L5741990-09-19019 September 1990 Forwards Rev 2 to Browns Ferry Nuclear Plant Cable Issues Supplemental Rept Corrective Actions,Sept 1990. Rept Revised to Clarify Cable Bend Radius & Support of Vertical Cable & Document Resolution of Jamming Issues ML20064A6871990-09-18018 September 1990 Requests Closure of Confirmatory Order EA-84-054 Re Regulatory Performance Improvement Program ML20059L4931990-09-17017 September 1990 Provides Addl Info Re 900713 Tech Spec Change 290 Concerning Hpci/Rcic Steam Line Space Temp Isolations,Per Request ML18033B5171990-09-17017 September 1990 Forwards Addl Info Re 900524 Tech Spec Change 287 on Reactor Pressure Instrument Channel.Schematic Diagrams Provided in Encl 2 ML20064A6851990-09-17017 September 1990 Responds to NRC Recommendations Re Primary Containment Isolation at Facility.Background Info & Responses to Each Recommendation Listed in Encl 1 ML20059K2971990-09-14014 September 1990 Responds to NRC 900208 SER Re Conformance to Reg Guide 1.97, Rev 3, Neutron Flux Monitoring Instrumentation. TVA Endorses BWR Owners Group Appealing NRC Position Directing Installation of Upgraded Neutron Flux Sys ML20059H3861990-09-10010 September 1990 Forwards Corrective Actions Re Radiological Emergency Plan, Per Insp Repts 50-259/89-41,50-260/89-41 & 50-296/89-41. Corrective Action:Plant Manager Instruction 12.12,Section 4.11.3.1 Revised ML18033B5031990-08-31031 August 1990 Forwards Financial Info Required to Assure Retrospective Premiums,Per 10CFR140 & 771209 Ltr ML20059E1741990-08-31031 August 1990 Informs That Plant Restart Review Board & Related Functions Will Be Phased Out on Date Fuel Load Commences ML20059D7061990-08-28028 August 1990 Requests That Sims Be Updated to Reflect Implementation of Program to Satisfy Requirements of 10CFR50,App J.Changes & Improvements Will Continue to Be Made to Reflect Plant Mods, Tech Spec Amends & Recommendations from NRC ML18033B4931990-08-20020 August 1990 Suppls Response to Violations Noted in Insp Repts 50-259/90-14,50-260/90-14 & 50-296/90-14.Corrective Actions: TVA Developed Corporate Level std,STD-10.1.15 Re Independent Verification ML20063Q2431990-08-20020 August 1990 Responds to 900807 Telcon Re Rev to Commitment Due Date Per Insp Rept 50-260/89-59 Re Electrical Issues Program ML20063Q2451990-08-17017 August 1990 Provides Revised Response to Generic Ltr 89-19 Re Request for Action Concerning Resolution of USI A-47, Safety Implication of Control Sys in LWR Nuclear Power Plants & Notification of Commitment Completion ML20063Q2441990-08-17017 August 1990 Advises That IE Bulletin 80-11 Re Masonry Wall Design Implemented at Facilities.Design Finalized,Mods Completed, Procedures Issued & Necessary Training Completed.Sims Data Base Should Be Updated to Show Item Being Implemented ML20059A4861990-08-16016 August 1990 Responds to Verbal Commitment Made During 900801 Meeting W/Nrc Re Control Room Habitability.Calculations Performed to Support Util 900531 Submittal Listed in Encls 1 & 2 ML20059A5141990-08-16016 August 1990 Provides Response to NRC Bulletin 88-008,Suppl 3 Re Thermal Stresses in Piping Connected to Rcs.Util Does Not Anticipate Thermal Cyclic Fatique Induced Piping,Per Suppl 3 to Occur in Plant.Ltr Contains No Commitment ML18033B4821990-08-14014 August 1990 Submits Revised Response to Violations Noted in Insp Repts 50-259/89-16,50-260/89-16 & 50-296/89-16.Extends Completion Dates for Commitments to 901203 ML18033B4831990-08-13013 August 1990 Responds to NRC 900713 Ltr Re Violations & Deviations Noted in Insp Repts 50-259/90-18,50-260/90-18 & 50-296/90-18. Corrective Actions:Craft Foreman Suspended for Three Days & Relieved of Duties as Foreman ML18033B4811990-08-10010 August 1990 Responds to NRC 900710 Ltr Re Power Ascension Testing Program.Four Hold Points Selected by NRC Added to Unit 2 Restart Schedule ML18033B4801990-08-0808 August 1990 Forwards Response to SALP Repts 50-259/90-07,50-260/90-07 & 50-296/90-07 for Jul 1989 - Mar 1990 ML20044B2121990-07-13013 July 1990 Clarifies Util Position on Two Items from NRC 891221 Safety Evaluation Re TVA Supplemental Response to Generic Ltr 88-01 Concerning IGSCC in BWR Stainless Steel Piping.Insp Category for Nine Welds Will Be Changed from Category a to D ML18033B4371990-07-13013 July 1990 Forwards Corrected Tech Spec Page 3.2/4.2-45 to Util 900706 Application for Amend to License DPR-52 Re ADS ML18033B4331990-07-13013 July 1990 Requests Temporary Exemption from Simulator Certification Requirements of 10CFR55.45(b)(2)(iii) ML20055F6091990-07-12012 July 1990 Provides Response to NRC Bulletin 88-003 Re Insp Results. No Relays Found to Have Inadequate Latch Engagements. Therefore,No Corrective Repairs or Replacement of Relays Required ML18033B4251990-07-10010 July 1990 Forwards Cable Installation Supplemental Rept,In Response to NRC Request During 900506 Telcon.Rept Contains Results of Walkdowns & Testing Except Work on Ongoing Cable Pullby Issue ML18033B4241990-07-0606 July 1990 Advises That Util Expects to Complete Implementation of Rev 4 to Emergency Procedure Guidelines by Mar 1991.Response to NRC Comments on Draft Emergency Operating Instructions Encl ML18033B4201990-07-0505 July 1990 Provides Basis for Closure of Generic Ltr 83-28,Item 4.5.3. Util Has Concluded That Analyses Presented in BWR Owners Group Repts Acceptable for Resolving Issue,Subj to Listed Conditions ML18033B4091990-07-0202 July 1990 Responds to NRC 900601 Ltr Re Violations Noted in Insp Repts 50-259/89-53,50-260/89-53 & 50-296/89-53.Corrective Actions: Condition Adverse to Quality Rept Initiated & Issued to Track Disposition of Deficiency in Chilled Water Flow Rates ML20043G4901990-06-14014 June 1990 Forwards Tabs for Apps a & B to Be Inserted Into Util Consolidated Nuclear Power Radiological Emergency Plan ML20043H3511990-06-14014 June 1990 Forwards Corrected Pages to Rev 15 to Physical Security Contingency Plan,As Discussed During 900606 Telcon.Encl Withheld (Ref 10CFR73.21) ML20043F4951990-06-11011 June 1990 Advises That Facilities Ready for NRC Environ Qualification Audit.Only Remaining Required Binder in Review Process & Will Be Completed by 900615 ML18033B3651990-06-0808 June 1990 Forwards Revised Page 3.2/4.2-13 & Overleaf Page 3.2/4.2-12 to Tech Spec 289, RWCU Sys Temp Loops. ML18033B3391990-06-0404 June 1990 Responds to NRC 900504 Ltr Re Violations Noted in Insp Repts 50-259/90-08,50-260/90-08 & 50-296/90-08.Corrective Actions: Individual Involved Counseled on Importance of Complying W/Approved Plant Procedures When Performing Assigned Tasks ML20043D3251990-06-0101 June 1990 Responds to NRC 900502 Ltr Re Notice of Violation & Proposed Imposition of Civil Penalty.Corrective Actions:Snm Program Action Plan Being Developed & Implemented,Consisting of Improved Training for Control Personnel & Accountability ML18033B3551990-05-31031 May 1990 Forwards Response to 891219 Request for Addl Info on Hazardous Chemicals Re Control Room Habitability ML20043C1951990-05-30030 May 1990 Forwards Response to Generic Ltr 90-04 Re Status of Implementation of Generic Safety Issues ML20043C0601990-05-29029 May 1990 Forwards Response to Violations Noted in Insp Repts 50-259/90-12,50-260/90-12 & 50-296/90-12.Util Admits Violation Re Access Control to Vital Areas,But Denies Violation Re Backup Ammunicition for Responders ML18033B3351990-05-25025 May 1990 Provides Basis for Closure of Generic Ltr 83-28,Item 4.5.3, Reactor Trip Sys Reliability. Analyses Presented in BWR Owners Group Repts Acceptable for Resolving Issues Subj to Listed Conditions ML18033B3221990-05-21021 May 1990 Forwards Rev 1 to ED-Q2000-870135, Cable Ampacity Calculation - V4 & V5 Safety-Related Trays for Unit 2 Operation, as Followup to Electrical Insp Rept 50-260/90-13 Re Ampacity Program ML18033B3101990-05-18018 May 1990 Responds to NRC 900417 Ltr Re Violations Noted in Insp Repts 50-259/90-05,50-260/90-05 & 50-296/90-05.Corrective Action: Senior Reactor Operator Assigned to Fire Protection Staff for day-to-day Supervision of Fire Protection Program ML20043A6101990-05-15015 May 1990 Forwards Rev 16 to Security Personnel Training & Qualification Plan.Rev Withheld (Ref 10CFR2.790) ML20043A4091990-05-14014 May 1990 Forwards Rev 14 to Physical Security/Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20043A4081990-05-14014 May 1990 Forwards Rev 15 to Physical Security/Contingency Plan, Consisting of Changes for Provision of Positive Access Control During Major Maint & Refueling Operations to One of Two Boundaries.Rev Withheld (Ref 10CFR73.21) ML18033B2921990-05-0909 May 1990 Provides Info for NRC Consideration Re Plant Performance for Current SALP Rept Period of Jan 1989 - Mar 1990.Util Believes Corrective Actions Resulted in Positive Individual Changes & Programmatic Upgrades ML20042F7401990-05-0404 May 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.' TVA Will Finalize Calculations for Switch Setpoints Prior to Units Restart ML20042F7701990-05-0404 May 1990 Provides Results of Review of Util 890418 Submittal Re Supplemental Implementation of NUMARC 87-00 on Station Blackout.Implementation of 10CFR50.63 Consistent W/Guidance Provided by NUMARC 87-00 ML20042F3721990-05-0202 May 1990 Forwards Corrected Monthly Operating Repts for Jan-June 1989 & Aug 1989 - Jan 1990.Discrepancies Involve Cumulative Unit Svc Factors & Unit Availability Capacity Factors ML18033B2631990-04-12012 April 1990 Forwards Response to NRC 900212 Request for Info Re Power Ascension & Restart Test Program at Unit 2.Util Has Refined Power Ascension Program to Be More Integrated & Comprehensive ML18033B2551990-04-0909 April 1990 Responds to NRC 900309 Ltr Re Violations Noted in Insp Repts 50-259/89-16,50-260/89-16 & 50-296/89-16.Corrective Actions: Contractor Will Perform Another Check Function Review for Mechanical Calculations & Area Walkdowns Will Be Conducted ML18033B2431990-04-0202 April 1990 Responds to NRC 900302 Ltr Re Violations Noted in Insp Repts 50-259/89-43,50-260/89-43 & 50-296/89-43.Corrective Action: Surveillance Insp Revised to Prevent Removal of All Eight Emergency Equipment Cooling Water Pumps from Water 1990-09-19
[Table view] |
Text
I ACCELERATED DISIAUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8811300164 DOC.DATE: 87/06/22 NOTARIZED: NO DOCKET FACIL:50-259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION GRIDLEY,R.L. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
R
SUBJECT:
Forwards addi info re -App R draft safety evaluation rept 1 open items,per 870323 request.
DISTRIBUTION CODE: DF01D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: Direct Flow Distribution: 50 Docket (PD Avail)
NOTES:1 Copy each to: F.McCoy,J.G.Partlow, S.Richardson,S.Black, 05000259 S B. D.Llaw.
1 Copy each to: S.Black, B. D.Llaw,F.McCoy.
J.G.Partlow, S.Richardson 1 Copy each to: S. Black, J.G.Partlow,S.Richardson 05000260 05000296 g j
B. D.Liaw,F.McCoy.
D
'RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL INTERNAL: NUDOCS-ABSTRACT 1 1 ~Q~EG F~ 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NOTES: 5 5
/ S l
NVZE 'lO ALL "RIDS" D, RECIPIEÃIS'IZASE HELP US IO REDUCE WASTE! CONI'ACT 'IHE DOCGMEPZ COFZfRL DESK, D ROOM Pl-37 (EXT. 20079) TO ELIHINATE YOUR KQK FKM DISTEGBUTION LISTS KlR DOCUMEHX8 YOU DON'T NEED!
tt lI TOTAL NUMBER OF COPIES REQUIRED: LTTR ~ ENCL
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i 11/28/1988 18'12 t SITE LICENSING BFN L44 8706Z2 SOS t28S '729 3111 P.82 5N l578 I.ookout. Place JUN Srm8>
U.S. Nuclear Regulatory Commission ATTN. Document Control Desk Qaghington, D.C. 20555
%mtlemen'n 4he Matter of Docket Nos. 50-259 Tennessee Valley Authority 50-260 50 296 BROWNS FEISTY NUCLEAR PLANT (BFN) 10 CFR 50, APPENDIX R l
eq ested in the NRCITVA March 23,. 987 meeting held in Bethesda, Maryland, ve are providing the. enclosed material to resolve certain Appon x r .
safety evaluation report open items.
W'is expected Chat folloving the rev ev of this material a final safety reviev evaluation for BFS vill be issued. Shou ld additiona an orma
~bisect be required, please refer any questions to James D. o co Licensing at (205) 729-2689.
very truly yours>
TENNESSEE VALLEY AUTHORITY Qdghel Nyek 8f R. I Oddly'.
Gridley, Director Nuclear Safety and Licensing
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'OC1 See page ggn6
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F'DC 88il300i64 870622 PDR ADOCK C "4
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Owg~ FERR RESPONSE TO NRC REQUEST Foi{ ADDlTZONAL ZNFORHATlON FROM MARCH 23, 1987 APPENDIX R MEETING IN BETHESDA, MARYLAND Item 3.4. 3-1 Containme t Atmos her e Dilution CAD Valve Ali nrnont
- 1. Pz'ovide justification for changing manual action to align CAD system valves f rom one hour to two bours.
2, Show that alignment of CAD system is needed only for non-automatic depressurkxatkan system (ADS) valves.
4 3, Show that the operatoz'an achieve manual actions to align CAD system valves after a fice.
- 4. Shaw that each fice area has at least ane ADS valve available during a fico event.
Show that safe shutdown can be achieved vith one 'relief valve after initial reactor depressurixatian.
MSPJMSP<
The manual action of aligning the cAD system is to provide pneumatic.
(r6trogen) supply ta operate the main steam relief valves (MSRVs) for safe shutdown after a fkze event. The MSRVs have tvo safe shutdown functions. The MSRVs are required to depressurixe the zeactoz'essel which allovs the RHR system ta operate jn the low pz'essure coolant kn)ection {LPCZ) mo4e and maintain coolant inventory. For this function, three HSRVs are required within the first, 20 minutes of the fice event. The initial pneumatic supply for these MSRVs is from the tvo receiver tanks af the drywall air central system, each with a 57 ft3 capacity. MSRVs with the ADS function have their own accumulators as backup pneumatic supply. These ADS accumulators aze sixed for five valve operations. only one valve operation (to apen valve) is required of the MSRVs {bath ADS and non-ADS). It is estimated that the receiver tanks of the dzywell air control system az'e capable of holding the MSRVs open for at least one hour. The zeactoc'ill be sufficiently depressurixed by then. Tharefare, no manual action is required to provide pneumatic supply to the MSRVs for the first safe shutdown function of the HSRVso After the initial depressurixatkon, the MSRVs are used ta provide a flaw path foz removing decay heat from the reactor vessel to the suppression pool. Only one MSRV is required for this function (See table 4.1 of HEDC-31'Xl9). For each fire area or fire xone, at least one MSRV used foz'he initial, de ressurisatkon vill be an Aps valve. This valve vill be sufficient to satisfy the second safe shutdown function since it has its awn accumulator SL'xed far five valve operations. Except for minor leakage through the '
solenoid vaLve to the MsRU pneumatic actuator, the operate.an of h MSRV open will nat axpend any pneumatic supply. It is estimated that the ADs
28S 729 3111 P.84 tem 3.4.3-) Co tainment Atmos here Dilutian cAD velya All ament (Continued)
QEEPONSE (Continued) accumulator can keep the valve open for at least 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (refer to sER dated 7/24/85, Subject NUREC-0737, Xtem l?.K.3.28, "Qualification of ADS Accumulator"). Therefore, no manual action is required to ensure the flow path for residual hest xemovsl circulation from the reactor vessel to the suppression pool within the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the blawdown.
The manual action which slijns the CAD system to pxovide nitrogen ta operate the MSRVs within ane hour is s conservative requirement which allows the flexibility to use the non-ADS valves for the second safe shutdown function The required manual, action>> involve opening two one-inch manual valves snd clasing another two-inch manual valve which can be pex'formed within two to three minutes. The equivalent fixe loads for the fire zones in question are approximately 30 to 40 minutes. The area containing the valves (Elevation 565 of such Reactor Building, Fire Zones 1-1, 1-2, 2-1, 2-2, 3>>1, snd 3-2) hss ares detection snd suppression, emexgency lighting, limited combustibles in the axes, a high ceiling (20 to 30 feet) snd large open axes, snd the presence of an open hatch about 50 feet fram the CAD location, portable smoke ejectors, self-contained breathing apparatus, and portable lanterns <ill be available for use by the operators, if necesssxy, to minimize any potential problems with smoke obscurstian. To alleviate any possible concern on this matter, the one-hour requirement in the manual action tables will be changed to two hours. Therefore, the operator will have s significant amount af time after the fire to perform the manual action requiring only twa to three minutes.
Ttem 3.4.3-2 Manual Actions
- l. Show that each fire ares (with the exception of fire axes 16 af the Contxol Building) hss at least one emergency equipment cooling water (EEL) pump auto-start cspsbi.lity.
- 2. provide analysis to support extending the time limit for manual start of EECM pumps,
- 3. Commit to update correction to manual action calculation table for stax ting EECW pumps.
- 4. Show cost impact of performing modification to provide EEL pump auto-start feature for a Contral Building fire.
@LCPo~
The current post,-fire shutdown procedure for BFN has s manual action requirement to manually verify the EEcW pump availability in five minutes after the start. of a diesel generator, normally assumed to be at the initiation of the fire event.
11r28>1988 18r 14 S I TE L I CENS ING BFN 85 '729 3111 P @S
~pBVON F (Coflthtlu+4) hecent review of the plant configuration arrd the EECk'ump "tart logic indicates that the previous approach and assumptions went. beyond the design basis for Appendix. R fire event.s, The EECL pump logic includes an aut.omat.ic st aPt circuit r y based on diesel generator sr art recogni ti ori signals and a manual start. and stop circuitry fo". each pump, .he circu' y for the d esel ger>era or st a. t I ecogniti ari si gnalg, isfans jde he heac o.- buildings Diesel Generaror Buildings, and the respective shutdown board rooms. -.he c.'rcu;c."y for the manual st.art and stop signals is inside the cont.ro'uilding and t)ie respective shutdown board rooms and locally in the iotake pump "tat ion.
Analysis demonstrates that for a ire in any o. he fire areas (4 through '5 and 17 through 24), at, least two EEC4'umps would start. automatically upon a start signal wit;h its associated diesel generators, thus eliminating the need to manually verify the EEC)'ump sta. t from ou:side 'he cont:rol room for these areas.
Analysis also indicates that the auto-start circuitries associated <ith the diesel. generator run zecognition for the two divisions of t;he EECt4 pumps (A3 and C3 versus 33 and D3) are separated from each other by the wall between the unit 2 and unit 3 Reactor Buildings {figure 1). Thus, EECI'umps A3 and C3 would be available for a fire in unit 1 or 2. Fo" most loca ions in units 1 and 2, the A3 and C3 pumps wauld stazt. automatically. Howeves', for fire tones 1-3 and 2-3, because the cont, ol power feeder to 4 kV Shutdown Board 3EA is located in these two zones, the automatic start Of the A3 pump cannot be ensured. The A3 pump can be starzed by manual ac ion outside af the contz'ol room. The automatic start of the C3 pump is ensured for these two fire zones. Far fire zone 1-3, only unit 3 diesel generataz's az'e used .Or safe shutdown, The C3 yuzy provides dizect cooling water to the un't 3 diesel generators and no spurious operation of the sectionalizing val valves on the EECV headez would defeat this direct- cooling capability. Fo" fire zone 2-3, one of the unit 1 and 2 diesel generators (DG A) is used foz supply'ng power to
't unit 1 RHR pump and other required ac equ'pment. Since unit 11-is not a f're affected unit for a fire in .'re zone 2-3, the diesel generator DG A which provides power to un=t 1 w ll no. be reauireC fo. the 'r '
event. A spur ious operation o tne sec.iona i"'ng va ve 'nn un't . 2 (FCV-67-2')
would not affect the unit 3 diesel genezators w'.hach supply pawer to unit 2; however, it wzruld cut off the SEC)'low from EECM Pump C3 to "he uni s 1 anC 2 diesel generators. X this happens, the ape"atoz can tr'p the required dieso generator PG h. from the control room upon receiving the high temperature alarm. The operataz would then manually start the EEL pump before restartizrg the diesel generator at 30 ~utes, Ehould the operator ~ail ta trip diesel generator and should the diesel generator subsequently fail, the o p orator can align anathe" unit 3 diesel genewtor to supply t5e necessary ac powez'o support the safe shutdown .unc con in uni 1-3 and 2-3, M the KECM flow from EECW pump C3 does not provide adequate coaling water'o the requireC diesel genera a.s, M ope- o eratoz could close ~We
'un it 3 ssectianaJ.izivg valves from the main control room loads. The operator would then manua11y smrt the EEW pump A3 locally when i
oom to isolate unessential
iiz28ri988 i8:i5 SITE LLCENSING BFN 85 729 3iii P.86
~q gpggsz <continued) l required to opex'ate the unit and unit 2 diesel generators. Therefore, the Reactor Buildings (fire areas X through 3) will only require a manual action inside the control room'o ensure an adequate cooling water supply to the 4iesel generators.
The intake pumping station {part of fire area 25) will have a two-pump auto-start capability; however, a spurious trip of one pump would leave a single pump to supply the xequired diesel generators. To ensure adequate flow, the operatoz may be requixed to either close the EECW headex sectionalicing valves or manually start a second EECW pump at, the shutdown board. The circuits to sectionalicing valves would be unaffected by s fire in the intake pumping, station and could be opexated from the control room The Control Building (fire area l6) contains only the manual staxt and stop cix cuitry for all of the EEcW pumps. Consequently, s fire in the Contx'ol Building will not affect the auto-start circuitry of the EEL pumps and the EECW pumps would receive the automatic start, signal if the diesel generators should start. An EECW pump ~ould not. start during a Contr'ol Building fire only if the fire damages the manual stop cizcuitry causing the EECW pumps to spuriously stop. Zt would requize four simultaneous and identical spurious o ez'ations to cause a total loss of the EECW pumps, more spurious operations to cause s total loss of the E CW pump It would x'equire even fully loading up the diesel gsnezatars. These conditians are beyond the requiz'ed spurious operation assumptaons g1ven in Generic
~ ~
er4c Letter Le 86-%0.
One af the proposed fixes to ensuxe the diesel generator availability is to provide an automs ticc tri r p oof thee diesel,genexators based on high jacket water temperature. A simple trip system would have two sa feet, y gra radee temperature sensors with dual contacts at each diesel generator and connecting the logic l i l to' loca t erm na block oc for the tzip relays. The cost estimate foz this simple system is approximately 4589,770 for sll ree un s-includes all engineering, materials, and constru t ction costs. The estimated schedule is appx'oximately 30 ~eeks fox engineex'ing and 60 days for constructian. Howevez, this modification is nat desirable because he s e protect on for or thee diesel generstozs will also contribute to the unavailability of the diesel generators dur ng a a ~
high cost cannat be justified in view of the fact that the 'trip may reduce 1,'.
e reliability of the diesel genexators.
Based on the above, it is evident that the five minute manual act1on
.requirement for the EECW pumpss is unnecessary.
unnec Analysis demonstrates that, All fire zones m fme areas
~ ~
d'l for a fire in fire area locat1on s 4 throu g h 24, st less t two EKCW pumps would start automatically upon a start siign al from its assoc'.atet d 1es generator.
l-snd 3, except for cones -3 an 4 2-3 wauld auto start capability of at least two EECW pumps. Foz ha ve thee au or fire ire axes 25 and fire cones 1>>3 and 2-3 spurzous pump trip ri or valve closux'e would leave a single EECM pump w ~hich ustee until appropriate contz'ol room <or ich is aadeequs (or local) oca manual actions are taken. The manual actions ta will be changed to remove tablee w
ii~28<L988 i8: iS SITE LICENSIHG BFH 85 '729 3lla P.87 r~
t Continued) e'R~S)gag the 5-minute action accordingly making the 'soonest oper'atox ection 10 minutes, The 10-minute actions are either to vexify the scram and isolation functions ox to disable the HpCE system overfilling the reactor vessel.
if it is operating unreliably and t 3,4.6-2 << Credit for Dr all Coolers QQ~U~
- 1. State that drywall coolexs are not needed or taken credit for during an Appendix R event.
Z. State that drywell coolers will not be deliberately txipped unless the entxy conditions are met.
gg~pggf The drywell coolers are not requi.red for an Appendix R safe shutdown event.
However, the drywell cooling function will not be delibexately removed unless the following thx'ea entry conditions for the safe shutdown procedure are met:
1 A confirmed fire occurrence of significant severity in any plant location in which an immediate fire extinguishment cannot be achieved.
- 2. Loss of adequate high pressure makeup capability such that no x'easonable alternate exists but to proceed to a low pressure source of makeup.
- 3. Xnability to pxovide normal ox'mergency powex to vital equipment.
Since the dx'ywell. blowers are not an essential safety function, they were not A dix R xe uixements. To pxeclude the possibility of the drywall blower cables from contributing to a high impedance fault ul con condition the breakex'o the blower is tripped when its respective reactor motor i li operated valve b oar d s a gne d, For those cases in which the RHOV boaxd is
~
used fax the A endix R event and in which it is essential that thee drywall cooli,ng function be defeated, either powex' to thee blowers ox source power to the board vill be removed.
.4.3-4 - Scr Verification
~ggES'f State th a t shutdown s u pxoceduxes will allow scram vex'ific e ification to be verified by txipping reactor pressure system (RPS) mo o g enerator motor (HG) set breaker from the battery board room.
11r28r1988 18:16 SITE LICENSING BFN
~85 729 3111 P.88
~ (I R~PEM Th e Appen dix R Procedure will require either opening the RFS NG set generator u p ut breaker on the battery board ar opening the feeder breaker to the RFS out HC set mator on the 480V reactor motor operated valve board, This action will ensure power is removed from the reactor protection system which causes an automatic scram on loss of power and verifies the scram will take place.
2 em 4.0,h.3 - Tome Level and Tem eratura Instrumentation
~RUEST
- 1. Summarize the previously made case (HEDC 31119, November 21, 1986 submittal) that the instrument ci.rcuits will survive a fire event.
2, summarize the previously made case (NEDC 31119, November 21, 1986 submittal) for not requiring the use of the instruments,
- 3. Provide cost impact of performing modification to ensure instrument indication availability during a fire event on elevation 593'f reactor building.
QEs~PPsE Documentation has been provided in MEDC-31119 and in the November 21, 1986 submittal which demonstrates that the suppression pool level and temperature instrumentation are not required to mitigate the design basis Appendix R avant.
Xn particular, the analysis has shown that the instrumentation will not be re to determine any operator actions to preserve paol level or renduuired re temperature. All operator actions are based on reactor con ditions o {reactor pressure and reactor ~ster level) and preanalyzed worse-case conditions in whi c h a ctions are performed regardless of the torus canditions. Operator act on inn thee Appen endix R Shutdown Procedure will not depend upon owled 8 e of o knowl torus level or temperature. All sources which are caps a ble e oof overfilling or dra n ng th e toorus have ve been determined and these sources will be isalated.
Th e prace d ur es will direct the operators to depressurize thee r reactor and establish a heat removal path {alternate shutdown coaal ing) based u p on low reactor water level or high torus temperaturee or the inability to determine torus temp erature. Analysis demonstrates that the peak tor'us temperature b 1 t rus temperature limits given the worst Hx'arne assumed {two thee&'ct to three hours) for performance of these operator actions, ons, Therefore, torus temperature and level process variables are not n necessar y to perform or control the functions given in XXX.L.2 of Appendix R-f the cable routing far the torus instrumentation was provided in the November 21, 1986 submittal. The cables or e inside the Reactor Building stay within its own ~nit. Separat on e wee
~
11<28<1988 18: 16 SITE LICENSING BPN S 729 3111 P.89 R~{ SPOMSK {Continued) redundant trains exists on the 519 and 565 elevstians of each Reactor Building. However, the north area of khe 593 elevatian of the Reactor Buildings is the arse of the closest proximity for most of the torus instrumentation cables. Consequently, the situation is that for a fire in other than the north area (593 elevatjan) of each Reactor Building, the an)'rea torus instrumentation is available to the operator if hs desires ta canfirm the torus condition. Since a separate loop of the torus instrumentation is available at the backup control panel, the t.orus instrumentation is available for e fire event in the Control Building. For the north area of the 593 elevation, it is expected that sufficient instrumentation would survive a fire on this elevation at this particular lacatian where ths redundant trains of torus instrumentation are st the closest proximity. This is because the north area af 593 elevation is presently protected by an existing open head spray system and will be protected by e sprinkler system designed to NFPA-1$
requirements. The major combustible in this area is cable insulation which is either in conduit or coated liberally with Flamastic.
If the instrumentation did not survive, alternate means exist to determine ths torus conditions i.f the operator desires so. For example, the RHR eat exchanger inlet temperature can be used to determine the pool temperature.
The cables for the RHR heat exchanger inlet temperature sruti.th indication in the control raom is approximately 10 fest fram the closest proximity for the pool temperature indications. If the cables are eisa lost during the fire, thermowells are available to provide a local determination of the torus temperature. Direct torus water level indicatian can be determined by taking diffsrsnti.al pressure readings from the instrument taps at the local torus level transmitter. Therefare, various alternate means can be used to determine the torus conditions should the operator desire such information.
The approx ima teel y,,
cost to separate the tor'us instrumentation is estimated to be
$ 2 238 876 faro all three units, engineering, materials, and canstruction costs. The es This estimate includes all duration is approximately 30 weeks per unit for engineering which could be escalated by 50 percent due to Class 1B and safety-related pracurement and 30 to 60 days for construction per unit.
Xn summary~ thee high g costs of providing further protection for torus instrumentatian at the 593 elevation cannot be )ustif is{4 based on the f'allowing cansideratians.
- a. The benefit of the torus instrumentation has been demonstrated ta bs minimal.
- b. It is highly probable 'that the existi.ng instrumentatian would survive a fire in the critical area,
- c. Alternate means tions are available to Ms of determining torus condit'on operators.
xx/85/ JWCIO JO< J I F
Item 4.0.h-4 - Thr'ee Phase Shorts on HI-Lo pressure Interface Cir cuit for Residual Heat Removal RHR Shut Down Coolin SDC Fla<
Contr al Valve PCV 74-48
~RU EST Commit to tag out the breaker far Fcv 24-48 to ensure removal o~ mat "e power.
State what steps aze necessary t,o ensure the valve is in its P>>P<<
position prior to tagging out and what controls exist to ensure the breaker is periodically verified for proper tagged aut pasition
~Rl~gSE Far HI-LO pressure interface valve FCV 74-48 motive power will be removed by opening the breaker .disconnect switch and placing a valve hold ozdex tag, on the breaker'nd the cantrol zoom switch. Before tagging out the equipment>
.... position indicating,lights are available to ensure the valve's proper position. The tagout i.s verified by a Semiannual Hold Order Audit which tagged'ut includes verification of the position of the tagged equipment.
Zte 4.0.8.4 - Hanua ctions in Fire Areas 4 8 9 X2 and 1,3
- 1. ConfiarL that the manual actions performed in these fire areas do nat x'equi.re caid shutdown repairs ta accomplish.
2, Clarify what manual actions are performed in these fire areas
- 3. Commit to correcting mistakes i.dentified in manual action calculations (Enclosure 4 of November 21, 1986 submittal to NEDC 31119) for these fire areas'nclosure 4 of the November 21, 1986 submittal provided a table of manual actions for each fire area. The table fax fi.re areas 4, 8, 9, 12, and 13 mistakenly identified an action to be performed an a potentially fiz'e damaged circuit breaker. This error was identified duri.ng development of the operating procedure end will be corrected in manual action calculations. The valves will be di.sabled by verifying that pawez'o the board is removed from its pawer source located in a different fire area. The valves are disabled to valve using the local handwheel. 'o prevent spurious operation and allow the operator to manually positian the repair or repair proeeduxe is required to obtain hot or cold shutdown conditions foz'hese areas.
12~28>1988 18r 2.7 SITE LICENSING pFN 8S 729 3111 P.ii 26-0 20 <<9 i9 I
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11/28/1988 18: 18 S I TE L I(;hN'hlNU 'l-N Item 4.0.8 5 - III.G. 1 and III.G.3 Fi t'e Areac
~R~UES 1 ~ Confirm that III,G.l fixe ax'eas identified in November 21, 1986 submittal to NEDC 31119 do not contain any redundant cables, equipmont, ox'omponents.
- 2. Clax'ify the withdx'awal of the zeal.G.3 exemption for the turbine building.
R SPO gg The following information is provided to clarify fire area ZZZ.G designations.
A fixe area designated solely as III.G.1 will have availablo redundant equipment, cabling, and systems for Appendix R that, is not located in that fire area.
The November 2l, 1986 submittal clarified the definition for alternate shutdown such that the turbine building (part of fix'e ax'ea 25, subject of exemption i) would now be classified as IXI.G.2 fire area. Exemption i.
includes a XXX.C.3 exemption for the turbine building which is no longer needed. Therefoxe, i that portion of Exemption will be withdrawn.
The intake pumping station fire axea classification of XZZ.G.1 as shown in the november 21, 1986 enclosure 3 table should bo IXZ.G.2.
It 4.b .6 T sti oi Safe Shutdown S stems
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- 1. Describe the initial, postmodification, and periodic testing of remote shutdown panel 25-32.
QEJi~~S; The backup contx'ol panel was ini,tially tested in the SFN preoperational test program using General Electric test procedures. Shen modifications are performed, postmodification testing is required to ensure the affected equipment vill operate as designed, The x'emote shutdown panel and electrical distribution system are currently'ested each operating cycle. This testing includes checks to determine remoto opexati.on capability and circuit i.solation from the Control Building.
~ 11r28r1 Jtlv ill:1u hl I C L JVCIROllBQ DCI'V Xtem from Section 3.c ~ 3 oE Draft Sachet Evaluation Rc art REVEST State what manual actions are being taken to ensure the fallowing objecti<<s can be accomplished within 10 minutes.
- l. Transfer to local control {if needed) and ensure closure of main steam isolation valves (HSEV).
2 ~ Close the high pressure coolant injection {HpcX) steam supply (in the fire affected unit) to prevent water intrusion into the shutoff'valve main steam lines.
Manual actions have been provided to verify that the main steam isalation valves are closed following the automatic or manual control raam isolation.
This acti. on involves accessing the remote control panel (panel 25-32),
transferring the control switches, end placing the control switch for each MSZV to the closed position. There are eight valves (four inboard and four outboard) and the action on the panel requires only seconds to accomplish, The HPCZ system can potentially operate spuri.ously requiring an action to prevent filling the reactor vessel up to the main steam lines. To prevent this occurrence, the HPCZ steam line will be isolated by closing the steam supply valve to the HpcX turbine. This action requires the operator ta transfer the breaker to emergency on the 250V reactor motor operated valve board h and then operate the control switch to the closed position. The access and actian time should only require one or two minutes.