Letter Sequence Response to RAI |
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MONTHYEARML20141E0011997-06-26026 June 1997 Forwards RAI Re Licensee 961211 Request for Amends to Licenses DPR-33,DPR-52 & DPR-68.Request Proposed Changing TS for as-found Setpoint Tolerance for Main Steam Safety/Relief Valves Project stage: RAI ML20198P6161997-11-0303 November 1997 Forwards Response to RAI Re TS Change TS-386,which Increases Main Steam Safety/Relief Valve Setpoint Tolerance Project stage: Response to RAI 1997-11-03
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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
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I Tennessee Valley Authority, Post Offee Su 2000, Decatur, Alabamo 35009-2000
. November.3, 1997 4
TVA BFN TS-386
'U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-Gentlemen:
-In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 - ,
REQUEST FOR ADDITIONAL INFORMATION REGARDING INCREASE IN MAIN STEAM SAFETY / RELIEF VALVE (S/RV) SETPOINT TOLERANCE In (TAC. NOS. M97412, M97413, AND M97414) (TS-386) }
This. letter provides supplemental information for the review of' Technical Specificaticns (TS) change TS-386 as requested by NRC in a letter dated June 26, 1997. TS-586 was submitted on December 11, 1996, andproposeschangesto)'([3.
the existina custom TS to increase the S/RV setpoint tolerance from approximately 1% to 3%. As stated in +
.the original submittal, TS-386 also provides additional
-technical' justification for the corresponding 3% S/RV i setpoint tolerance proposed in TS-362. TS-362 was submitted-on_ September 6, 1996, and is the proposed conversion package.to Improved Standard Technical Specifications format.
'Re'sponses.to_the five NRC requests are provided in the
-Enclosure. There are no commitments in this letter.
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-U.S. Nuclear Regulatory Commission Page 2 November 3, 1997 If you have further questions on this TS change, please contact me at (205) 729-2636.
Sincerely, l Qj' Uw T. E.
ney-Manager o Licensi g and Industry. ffairs Enclosure cc (Enclosure):
Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Begion II Atlanta Federal Center 601 Forsyth St., Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager U.S. Nuclear Regulacory Commission One White Flint, North 11555 Rockzille Pike Rockville, Maryland 20852
- _ _ _ . . _m .m - ._ _ .m _ _ __
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- _ ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY: NUCLEAR. PLANT-(BFN)-
-UNITS-1,.2,-AND 3 l
TECHNICAL SPECIFICATIONS (TS) CHANGE TS-386 INCREASED -SETPOINT. TOLERANCE MAIN STEAM SAFETY / RELIEF VALVES y (S/RVs)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 26, 1997 Below are-responses'to the five NRC questions provided in the subject request-for additianal information on TS-386. TS-386 was.uuhmitted on December 11,_1996, and proposes changes to the-existing.TS te support increasing the main steam S/RV setpoint tolerance from approximately 1% to 3%. A detailed engineering evaluation was provided in TS-386 and the effects of an increased S/RV setpoint tolerance have been evaluated in 4 the cycle-specific reload-analyses for Units 2 and 3 as referenced in the TVA response to question 5.a.
LTS-386 also provided added technical justification for the
, corresponding. 3% S/RV.setpoint-tolerance proposed in TS-362.
TS-362 wan submitted on September 6, 1996, and is the proposed conversion package to Improved Standard Technical Specifications format. The TVA responses provided below are likewise applicable to TS-362.
NRC Question 1-Enclosure 5, page 13 of the December 11, 1996 submittal states that the hydrodynamic main steam safety / relief valves (S/RVs) '
loads resulting from the-3% set;sint tolerance are less than 1%
higher than the " loads used in design". However, it is the NRC staff's experience via review cf other analyses, such as those Esupporting power;uprate at several boiling water reactors that-theLloads vary almost linearly with respect to the opening.
- pressure of the S/RVs. Therefore, it is expected that an
' increase in the S/RV setpoint to 3% would result in a 3%
increasefin1the S/RV hydrodynamic loads, ai 'What S/RV setpoints and setpcint tolerances were cssumed forEdeveloping the above " loads used in design"?
- b. Describe the methodology for determining the S/RV
. hydrodynamic. loads and the analysis results which indicate
_ ._ . . _ - -_- 1 u ._ _ . _ _ . _ _ _ __ _2-
only a 1% load increase for the 3% setpoint tolerance. It is also stated that the resulting 1% increase in loads is negligible when combined with other loads; therefore, the effect of load increases on the plant structures were not evaluated.
- c. What is the available margin such that these lead increases are negligible?
,TVA Response Below are the responses to questions 1.a, 1.b, and 1.c. It should be noted that the revised calculations were performed at
+ 3% of the highest nominal S/RV setpoint. Previous analyses were performed at + 1% of the highest nominal S/RV setpoint.
Therefore, the new setpoint evaluation represents a net increase of 2% in analyzed S/RV pressure setpoints rather than the 3% cited by the staff in question 1.
- a. The S/RV " loads used in design" were those described in the BFN Long Term Torus Integrity Program (LTTIP) Plant Unique Analysis Report (PUAR) und considered in the associated NRC Safety' Evaluation Report (SER) [BFN FSAR Appendix C References 12 and 22]. The S/RV setpoints and secpoint tolerances used in the LTTIP analysis and design of modifications were 1125 psig at a + 3% tolerance for all valves (i.e., S/RV opening pressure of 1136 psig) which translates to a S/RV mass f]ow rate of 308 pounds mass per ,
second through each S/RV (Rererence PUAR Appendix H, page H-BNL-16-4R2]. There are 13 S/RVs on each BFN unit. Of the 13 S/RVs, 4 are set at 1105 psig: 4 are set at 1115 psjg; and 5 are set at 1125 psig in accordance with TS 2.2.A.
Therefore, the LTTIP analysis was conservative since the highest S/RV setpoint and tolerance was used to establish the flowcate for all 13 valves.
- b. The LTTIP analyses for S/RV nydrodynamic loads were based on loads defined in accordance with the general design criteria in PUAR Section 4.2.2. Of the controlling load combinations, including Loss of Coolant (LOCA), thermal, earthouake, deadweight, and pressure, the S/RV blowdown j loads represent only about 35% of the total load combinatdan. Thus, a 2% increase in S/RV blowdown loads can be expected to increase-the total loading by less than 1% on a first estimate basis.
Additionally, an in-plant S/RV confirmatory test was conducted for BEN Unit 2 after installation of LTTIP E-2
modifications, as described in Section 4.2.2.1, Section 4.6,
- and Appendix C of the PUAR. This test showed that significant margin existed between the analytically predicted (calculated) stresses and actual stresses. S/RV load reduction factors were conservatively justified and applied for the torus support system and the vent system downcomers. These load reduction factors were approved in the LTTIP SER.
To analytically account for the change in S/RV setpoint tolerance from 1% to 3%, the opening pressures and S/RV mass flow rates were linearly extrapolated by the ratio 1.03/1.01. This corresponds to an opening pressure of 1159 psig and a mass flow rate of 314 pounds mass per second.
The S/RV confirmatory test results were also linearly extrapolated based on the ratio of the test conditions (opening pressure ot 1000 psig and mass flow rate of 268 y pounds mass per second) to the + 3% setpoint tolerance conditions (1159 psig and 314 pounds mass per second).
Other differences in the tested conditions and the + 3%
setpoint tolerance conditions were also conservatively considered in the extrapolation methudology. These included differences in torus water level and drywell-to-wetwell differential pressure.
Margins betwcen analytically predicted S/RV loads and the extrapolated S/RV test results were reduced by the TVA evaluation, as permitted by the LTTIP PUAR and SER. As noted above, the S/RV loading is only one factor in the controlling equations for stress, so the increase has a proportionally small overall effect. As expected, TVA's evaluation concluded that the overall combined loads on torus structures, systems, and components increased by less than 1% due to the S/RV setpoint tolerance change.
- c. The statement in the December 1996 submittal was: "When S/RV loads are combined with other design basis loads including dead weight, thermal, LOCA, and earthquake, the total increase is negligible and the affected components meet design basis requirements."
The key load . , combinations are associated with the original LOCA and S/RV hydrodynamic loads analysis, and the in-plant S/RV confirmatory test which demonstrated significantly lower stresses than analytically predicted.
The TVA evaluation determined that the increase in overall loads resulting from the small increase in S/RV flow rate (314 versus 308 pounds mass per second) associated with the E-3
i setpoint tolerance increase js very small when combined with the other torus loads. The resulting increase in loads was evaluated against existing design basis allowables defined in the PUAR, which includes all LTTIP load combinations. It was concluded that the small increase in S/RV loads yields very small combined stress increases. The combined stresses remain within existing design basis stress allowables established by the LTTIP PUAR and SER.
NRC Question,2 Enclosure 5, page 13 of the December 11, 1996 submittal states that the maximum differential due to the ir. creased S/RV setpoint was evaluated for valves in the Generic Letter 89-10 program. Please verify that that the calculations which demonstrate the capability of these valves to function adequately have been revised for these differential pressures.
TVA Response As discussed in the TS-386 submittal, the Generic Letter (GL) 89-10 motor operated valves (MOVs ) affected by the proposed new S/RV tolerance were evaluated for increased differential pressure loads and found to be acceptable. Subsequently, on October 1, 1997, TVA submitted a TS amendment request (Power Uprate Project) to increase the authorized reactor thermal power by 5%. Since this Power Uprate Project will require a general revision of many of the calculations for GL 89-10 valves, including those listed in TS-386, the master MOV calculations are not being updated to account for the 3% S/RV setpoint tolerance at this time. The Power Uprate Project is in progress and is scheduled to be first implemented in Fall 1998 on Unit 3.
In the interim, the engineering evaluation performed for the proposed 31 S/RV TS setpoint tolerance change has been linked to the subject MOV calculations uti]izing TVA's configuration control system. This ensures that Engineering personnel are cognizant of the 3% S/RV tolerance evaluation pending final MOV calculation revisions under the Power Uprate Project.
NRC Question 3 What is the basis for the high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) design requirement to provide design flow to the reactor ressel at pressures up to 1120 psig? In other words, how was the 1120 psig value originally determined?
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TVA Response The 1120 psig value is documented in the vendor (GE) design
-specifications for the HPCI and RCIC pumps, and is shown on the original system process diagrams as the maximum reactor pressure for HPCI/RCIC operating modes. A review of the original design and safety analysis documentation did not identify a specific system operating limitation associated with the 1120 psig. Rather, the 1120 psig value appears to have been selected as a conservative higher pressure bound for system design considerations.
NRC Question 4 The engineering analysis describes the calculations performed to determine the required additional flow, power, and rpm needed for HPCI and RCIC.
- a. Do the HPCI and RCIC turbines have a maximum allowable pressure?
- b. If so, what is the value and is the value exceeded due to increased S/RV tolerance?
TVA Response
- a. Yes, the HPCI and RCIC system turbines have a maximum allowable pressure.
- b. The design pressure for both the HPCI and RCIC inlet nozzle and steam chest is 1250 psig which is not exceeded by the increased S.'RV tolerance.
NRC Ouestion 5 TVA's submittal includes plant-specific analyses supporting the increased S/RV setpoint tolerance for Unit 2. The cover letter states that analyses supporting the increased S/RV setpoint tolerance is being performed for the next Unit 3 reload cycle.
- a. Verify that the necessary plant-specific analyses have been completed for the current Unit 3 operating cycle.
- b. Confirm that any differences between Units 2 and Unit 3 that exist in approved alternative operating modes (e.g.,
increased core flow, extended load line limit, or final feedwater reduction) have been addressed for Unit 3.
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ac . Describe how TVA will ensure that the increased setpoint tolerance will be reflected in future operating cycles-for each unit?
TVA Response
- a. The unit-specific analysis-for the current Unit 3 operating cycle has been completed and a copy of the report is included in Appendix N of the Updated Final Safety Analysis Report. The Unit 2 reload analysis is also in Appendix N for the cycle just completed. General Electric performs the cycle-specific analyses for BFN.
- b. A listing of the operating flexibility options in effect for Units 2 and 3 are listed in Section 8 of the reload analyses referenced in item 5.a above. The selected operating flexibility options may and often do vary from cycle to cycle for either unit. As a result, it can be expected that the reload analyses for the two units may have differences in the operating flexibility options.
Since the reload analysas are unit-specific and cycle-specific, differences between the two units are immaterial.
Both the Unit 2 and the Unit 3 analyses were performed using a 3% S/RV tolerance with the selected flexibility options. The results of these analyses are shown in the above referenced reload reports,
- c. An input to the cycle-specific reload analyses are system setpoints as defined in the TS and TS be.ses. If the requested TS is approved by NRC, the 3% S/RV setpoint will become the minimum input value used for the cycle-specific analyses for transients affected by S/RV setpoints. In addition, prior to performing the cycle-specific analyses, TVA and General Electric review the analysis assumptions to verify the suitability of the ar:_ ysis inputs.
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