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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR05000327/LER-1999-002-03, :on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event1999-10-15015 October 1999
- on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event
05000327/LER-1998-003-01, :on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status1999-05-27027 May 1999
- on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status
05000327/LER-1999-001-04, :on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc1999-05-11011 May 1999
- on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc
05000327/LER-1998-004-02, :on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check1998-12-21021 December 1998
- on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check
05000327/LER-1998-003-04, :on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced1998-12-0909 December 1998
- on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced
05000328/LER-1998-002-05, :on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure1998-11-10010 November 1998
- on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure
05000328/LER-1998-001-05, :on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays1998-09-28028 September 1998
- on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays
05000327/LER-1998-002-03, :on 980716,inadequate Surveillance Testing Was Discovered.Caused by Misinterpretation of ANSI Standard. Revised Appropriate Procedures to Provide Required Guidance1998-08-14014 August 1998
- on 980716,inadequate Surveillance Testing Was Discovered.Caused by Misinterpretation of ANSI Standard. Revised Appropriate Procedures to Provide Required Guidance
ML20236P6441998-07-10010 July 1998 LER 98-S01-00:on 980610,failure of Safeguard Sys Occurred for Which Compensatory Measures Were Not Satisfied within Required Time Period.Caused by Inadequate Security Procedure.Licensee Revised Procedure MI-134 05000327/LER-1998-001-04, :on 980519,automatic Reactor Trip Occurred. Caused by Failure of Alternate Feedwater to 1A1-A 480-volt Shutdown Board.Normal Feedwater Breaker Placed in Svc & 480- Volt Shutdown Board Returned to Operation1998-06-18018 June 1998
- on 980519,automatic Reactor Trip Occurred. Caused by Failure of Alternate Feedwater to 1A1-A 480-volt Shutdown Board.Normal Feedwater Breaker Placed in Svc & 480- Volt Shutdown Board Returned to Operation
05000327/LER-1997-014-01, :on 971101,discovered That RCS PORVs Were Not Cycled in Mode 4 as Required by Ts.Caused by Inadequate Procedures.Procedure Revised & Unit Cooled Down to Mode 41997-12-0101 December 1997
- on 971101,discovered That RCS PORVs Were Not Cycled in Mode 4 as Required by Ts.Caused by Inadequate Procedures.Procedure Revised & Unit Cooled Down to Mode 4
ML20199C2951997-11-13013 November 1997 LER 97-S01-00:on 971017,vandalism of Electrical Cables Was Observed.Caused by Vandalism.Repaired Damaged Cables, Interviewed Personnel Having Potential for Being in Area at Time Damage Occurred & Walkdowns 05000328/LER-1997-001, :on 970117,missed Surveillance on Auxiliary Contacts of Reactor Trip Breakers Discovered.Caused by Adequate PMT for Reactor Trip Breaker Was Not Performed. Multi-disciplinary Team Formed to Perform Investigation1997-10-0606 October 1997
- on 970117,missed Surveillance on Auxiliary Contacts of Reactor Trip Breakers Discovered.Caused by Adequate PMT for Reactor Trip Breaker Was Not Performed. Multi-disciplinary Team Formed to Perform Investigation
05000327/LER-1997-011-01, :on 970725,operations Training Personnel Found 125 Vdc Vital Battery Board 4 Improperly Aligned.Caused by Personnel Error.Provided Appropriate Disciplinary Action for Individuals Involved in Event1997-09-17017 September 1997
- on 970725,operations Training Personnel Found 125 Vdc Vital Battery Board 4 Improperly Aligned.Caused by Personnel Error.Provided Appropriate Disciplinary Action for Individuals Involved in Event
05000327/LER-1997-012-01, :on 970801,manual Reactor Tripped Due to Loss of Control Air.Caused by Corrosion Products (Rust Debris) Inhibiting Full Closure of One of six-inch Gate Valves. Isolated Breached Control & Svc Air Sys Header1997-09-0202 September 1997
- on 970801,manual Reactor Tripped Due to Loss of Control Air.Caused by Corrosion Products (Rust Debris) Inhibiting Full Closure of One of six-inch Gate Valves. Isolated Breached Control & Svc Air Sys Header
05000327/LER-1997-011-01, :on 970725,vital Battery Board 4 Was Operating W/O Battery Source.Caused by Personnel Error.Revised Appropriate Instructions to Caution of Possibility of Breaker Misalignment1997-08-25025 August 1997
- on 970725,vital Battery Board 4 Was Operating W/O Battery Source.Caused by Personnel Error.Revised Appropriate Instructions to Caution of Possibility of Breaker Misalignment
05000327/LER-1997-010-01, :on 970626,failure to Properly Return Portion of Fire Protection Sys to Svc Following Mod Activities Was Noted.Caused by Failure to Follow Procedures.Manual Isolation Valve Was Opened1997-07-31031 July 1997
- on 970626,failure to Properly Return Portion of Fire Protection Sys to Svc Following Mod Activities Was Noted.Caused by Failure to Follow Procedures.Manual Isolation Valve Was Opened
05000327/LER-1997-009-02, :on 970521,failed to Perform Response Time Testing of Containment Radiation Monitor Following Maint Activities.Caused by Misinterpretation of Surveillance Requirements.Tested Radiation Monitor1997-06-20020 June 1997
- on 970521,failed to Perform Response Time Testing of Containment Radiation Monitor Following Maint Activities.Caused by Misinterpretation of Surveillance Requirements.Tested Radiation Monitor
05000328/LER-1997-009, :on 970401,licensed Failed to Maintain 2 Offsite Power Sources as Required by Ts.Caused Because Main Contact Compression Was Out of Tolerance.Start Bus 1A Was Transfered to Normal Feed Breaker1997-05-0606 May 1997
- on 970401,licensed Failed to Maintain 2 Offsite Power Sources as Required by Ts.Caused Because Main Contact Compression Was Out of Tolerance.Start Bus 1A Was Transfered to Normal Feed Breaker
05000327/LER-1997-007-01, :on 970404,DG Started When Drill Bit Being Used to Drill Into Electrical Panel in Main Control Room,Cut Into Cable.Caused by Drill Bit Penetrating Energized Wire Bundle Causing Short.Work Stopped on All Electrical Mods for S1997-05-0101 May 1997
- on 970404,DG Started When Drill Bit Being Used to Drill Into Electrical Panel in Main Control Room,Cut Into Cable.Caused by Drill Bit Penetrating Energized Wire Bundle Causing Short.Work Stopped on All Electrical Mods for Shift
05000327/LER-1997-004-01, :on 970320,failure to Properly Perform Surveillance Testing on Circuit Breakers Was Identified,Due to Inadequate Review of Surveillance Instruction Revs.Lcos Were Entered1997-04-21021 April 1997
- on 970320,failure to Properly Perform Surveillance Testing on Circuit Breakers Was Identified,Due to Inadequate Review of Surveillance Instruction Revs.Lcos Were Entered
05000327/LER-1997-006-01, :on 970322,failed to Perform Surveillance Requirement During Containment Entry Due to Personnel Error. RWP Was Closed to Prevent Further Use1997-04-21021 April 1997
- on 970322,failed to Perform Surveillance Requirement During Containment Entry Due to Personnel Error. RWP Was Closed to Prevent Further Use
05000327/LER-1997-005-01, :on 970319,two of Six Tested Main Steam Safety Relief Valves Not within TS Setpoint Tolerance.Appropriate LCOs Entered & Valves Found Outside TS Setpoint Tolerance Adjusted to within TS Tolerance & Retested1997-04-17017 April 1997
- on 970319,two of Six Tested Main Steam Safety Relief Valves Not within TS Setpoint Tolerance.Appropriate LCOs Entered & Valves Found Outside TS Setpoint Tolerance Adjusted to within TS Tolerance & Retested
05000327/LER-1997-003-02, :on 970305,failed to Properly Perform Surveillance Testing on Centrifugal Charging Pump Inlet Isolation Valve Logic.Caused by Inadequate Surveillance Instruction.Prepared Special Test Procedure1997-04-0404 April 1997
- on 970305,failed to Properly Perform Surveillance Testing on Centrifugal Charging Pump Inlet Isolation Valve Logic.Caused by Inadequate Surveillance Instruction.Prepared Special Test Procedure
05000328/LER-1996-004-02, :on 960919,reactor Trip Breaker Was Removed After It Was Found to Have Inoperable Auxiliary Contacts. Caused by Inadequate Procedure.Revised Breaker Procedure & Reemphasized Requirements for Working Steps Out of Sequence1997-03-28028 March 1997
- on 960919,reactor Trip Breaker Was Removed After It Was Found to Have Inoperable Auxiliary Contacts. Caused by Inadequate Procedure.Revised Breaker Procedure & Reemphasized Requirements for Working Steps Out of Sequence
05000327/LER-1997-002-01, :on 970214,TS AOT Exceeded for DG 2A-A.Caused by Mechanical Failure of DG 2A-A Governor Actuator on Engine 2.DG 2A-A Governor Actuators on Both Engines Replaced, Functional Tested & PMT Performed1997-03-13013 March 1997
- on 970214,TS AOT Exceeded for DG 2A-A.Caused by Mechanical Failure of DG 2A-A Governor Actuator on Engine 2.DG 2A-A Governor Actuators on Both Engines Replaced, Functional Tested & PMT Performed
05000327/LER-1997-001-02, :on 970125,failed to Perform Surveillance Testing on EDG Start Timer Relays Contained in Start Logic Circuitry.Caused by Inadequate Surveillance Procedures. Revised Procedures & Performed Testing1997-02-24024 February 1997
- on 970125,failed to Perform Surveillance Testing on EDG Start Timer Relays Contained in Start Logic Circuitry.Caused by Inadequate Surveillance Procedures. Revised Procedures & Performed Testing
05000328/LER-1997-001-01, :on 970117,failed to Perform Surveillance on Turbine Trip Contacts of Reactor Trip Breakers.Caused by Inadequate Procedure.Performed Test on Turbine Trip Contacts & Declared Rtb a Operable1997-02-18018 February 1997
- on 970117,failed to Perform Surveillance on Turbine Trip Contacts of Reactor Trip Breakers.Caused by Inadequate Procedure.Performed Test on Turbine Trip Contacts & Declared Rtb a Operable
05000327/LER-1995-001-01, :on 950118,accumulation of Gas in Residual Heat Removal Sys.Caused by Normal Leakage from Cold Leg Accumulators.Performed Monitoring & Venting of Gas Accumulation & Revised Quarterly Pump Tests1997-02-0606 February 1997
- on 950118,accumulation of Gas in Residual Heat Removal Sys.Caused by Normal Leakage from Cold Leg Accumulators.Performed Monitoring & Venting of Gas Accumulation & Revised Quarterly Pump Tests
05000328/LER-1996-007-03, :on 961207,ESF Actuation,Start of Feedwater Sys,Occurred as Result of Inadequate Return of Equipment to Svc.Refresher Training on Filling & Venting Fundamentals Will Be Conducted in Yrs Training Cycle1997-01-0606 January 1997
- on 961207,ESF Actuation,Start of Feedwater Sys,Occurred as Result of Inadequate Return of Equipment to Svc.Refresher Training on Filling & Venting Fundamentals Will Be Conducted in Yrs Training Cycle
05000328/LER-1996-006-04, :on 961206,automatic Reactor Trip Occurred. Caused by Loss of Power to Start Bus 2A,start of Four EDG & Loading of EDG 2B-B.Refurbished Breaker Installed in 2A Start Bus & Breaker Tested Acceptable1997-01-0202 January 1997
- on 961206,automatic Reactor Trip Occurred. Caused by Loss of Power to Start Bus 2A,start of Four EDG & Loading of EDG 2B-B.Refurbished Breaker Installed in 2A Start Bus & Breaker Tested Acceptable
05000327/LER-1996-012-01, :on 960330,two Cold Leg Accumulator Sample Isolation Valves & Missing Data Sheet in Surveillance Package Were Inoperable.Caused by LCO Not Being Entered. Isolation Valves Were Tested1997-01-0202 January 1997
- on 960330,two Cold Leg Accumulator Sample Isolation Valves & Missing Data Sheet in Surveillance Package Were Inoperable.Caused by LCO Not Being Entered. Isolation Valves Were Tested
05000327/LER-1996-011-01, :on 961118,discovered Rod Position Indication Sys Was Out of Step W/Demand Position Indication Sys.Caused by Incorrect Position Indication on Analog Rod Position. Began Dilution of Reactor Coolant Sys1996-12-18018 December 1996
- on 961118,discovered Rod Position Indication Sys Was Out of Step W/Demand Position Indication Sys.Caused by Incorrect Position Indication on Analog Rod Position. Began Dilution of Reactor Coolant Sys
05000327/LER-1996-010-01, :on 961116,manual Reactor Trip Occurred.Caused by Unexpected Feedwater Heater Isolation.Informed Personnel of Effects Changing Proportional Band on Operating Range of Associated Valve Controllers1996-12-16016 December 1996
- on 961116,manual Reactor Trip Occurred.Caused by Unexpected Feedwater Heater Isolation.Informed Personnel of Effects Changing Proportional Band on Operating Range of Associated Valve Controllers
05000328/LER-1996-005-04, :on 961011,manual Trip Occurred.Caused by Unexpected Loss of Load.Stabilized Unit,Replaced Turbine Impulse Pressure Switches & Replaced Failed Feedwater Isolation Valve Motor & Brake Assembly1996-11-12012 November 1996
- on 961011,manual Trip Occurred.Caused by Unexpected Loss of Load.Stabilized Unit,Replaced Turbine Impulse Pressure Switches & Replaced Failed Feedwater Isolation Valve Motor & Brake Assembly
05000328/LER-1996-004-04, :on 960919,after Reactor Trip Breaker Was Removed Breaker Found to Have Inoperable Auxiliary Contacts. Caused by Inadequate Evaluation of Procedural Change. Procedures Revised1996-10-23023 October 1996
- on 960919,after Reactor Trip Breaker Was Removed Breaker Found to Have Inoperable Auxiliary Contacts. Caused by Inadequate Evaluation of Procedural Change. Procedures Revised
05000327/LER-1996-009-01, :on 960718,auxiliary Building Secondary Containment Boundary/Fire Barrier Not Maintained as Required by Design.Caused by Failure to Follow Design Control Process1996-08-19019 August 1996
- on 960718,auxiliary Building Secondary Containment Boundary/Fire Barrier Not Maintained as Required by Design.Caused by Failure to Follow Design Control Process
05000327/LER-1996-008-01, :on 960709,failed to Perform Quarterly Backseat/Closure Test on Five Check Valves.Caused by Personnel Error.Test Procedure Written & Performed on Subject Check Valves1996-08-0808 August 1996
- on 960709,failed to Perform Quarterly Backseat/Closure Test on Five Check Valves.Caused by Personnel Error.Test Procedure Written & Performed on Subject Check Valves
05000327/LER-1996-006-02, :on 960623,failed Coupled Capacitor Potential Device Caused Actuation of Generator Backup/Transfomer Feeder Relay Tripping Turbine & Reactor.Removed & Replaced Ccpd1996-07-18018 July 1996
- on 960623,failed Coupled Capacitor Potential Device Caused Actuation of Generator Backup/Transfomer Feeder Relay Tripping Turbine & Reactor.Removed & Replaced Ccpd
05000327/LER-1996-007-01, :on 960614,rod Position Indicators for Control Bank D Rods M4 & M12 More than Required 12 Steps Out from Respective Demand Position Indicators.Caused by Incorrect Position Indication on Rod.Response Reduced1996-07-15015 July 1996
- on 960614,rod Position Indicators for Control Bank D Rods M4 & M12 More than Required 12 Steps Out from Respective Demand Position Indicators.Caused by Incorrect Position Indication on Rod.Response Reduced
05000328/LER-1996-003-05, :on 960605,reactor Trip Breakers Manually Opened Because Shutdown Bank D Dropped Into Reactor Core.On 960607 Manual Reactor Trip Initiated Due to Dropped Rod. Added Caution Statement to Apropriate Operations Prcedure1996-07-0505 July 1996
- on 960605,reactor Trip Breakers Manually Opened Because Shutdown Bank D Dropped Into Reactor Core.On 960607 Manual Reactor Trip Initiated Due to Dropped Rod. Added Caution Statement to Apropriate Operations Prcedure
05000327/LER-1995-010, :on 950717,turbine & Reactor Trips Occurred. Caused by Bellows Being Deformed When Sudden Pressure Relay Is Isolated & Heated.New Qualitrol Relays Installed & Placed in Service1996-07-0303 July 1996
- on 950717,turbine & Reactor Trips Occurred. Caused by Bellows Being Deformed When Sudden Pressure Relay Is Isolated & Heated.New Qualitrol Relays Installed & Placed in Service
05000327/LER-1996-005-02, :on 960526,ESF Actuation Occurred Resulting in DG Start.Cause Unknown.Diagnosed Condition,Reset Start Signal & Secured Actuated Equipment1996-06-21021 June 1996
- on 960526,ESF Actuation Occurred Resulting in DG Start.Cause Unknown.Diagnosed Condition,Reset Start Signal & Secured Actuated Equipment
05000328/LER-1996-002-05, :on 960822,discovered Failure to Properly Identify Steam Generator Tube May Have Exceeded Tech Spec Plugging Criteria.Caused by Misjudgement by Two Independent Analysts.Evaluation Performed1996-06-13013 June 1996
- on 960822,discovered Failure to Properly Identify Steam Generator Tube May Have Exceeded Tech Spec Plugging Criteria.Caused by Misjudgement by Two Independent Analysts.Evaluation Performed
05000327/LER-1996-004-02, :on 960509,normal Feeder Breaker Unexpectedly Tripped,Resulting in Loss of Power Signal & Start of Four Dgs.Caused by Failure of Breaker Mechanism.Spare Breaker Operating Mechanism Refurbished & Installed1996-06-0505 June 1996
- on 960509,normal Feeder Breaker Unexpectedly Tripped,Resulting in Loss of Power Signal & Start of Four Dgs.Caused by Failure of Breaker Mechanism.Spare Breaker Operating Mechanism Refurbished & Installed
05000328/LER-1996-001-05, :on 960418,inadvertent ESF Actuation & Loss of Power Signal & Load Sequencing Occurred During Maint. Caused by Personnel Error.Individual Involved in Event Was Counseled & Lessons Were Discussed1996-05-17017 May 1996
- on 960418,inadvertent ESF Actuation & Loss of Power Signal & Load Sequencing Occurred During Maint. Caused by Personnel Error.Individual Involved in Event Was Counseled & Lessons Were Discussed
05000327/LER-1996-003-01, :on 960325,failure to Perform Surveillance Requirements for Penetration Fire Barrier Insps as Required by Tech Specs Occurred.Caused by Personnel Error.Appropriate Disciplinary Action Taken W/Individuals1996-04-24024 April 1996
- on 960325,failure to Perform Surveillance Requirements for Penetration Fire Barrier Insps as Required by Tech Specs Occurred.Caused by Personnel Error.Appropriate Disciplinary Action Taken W/Individuals
05000327/LER-1996-002-01, :on 960215,SRs Associated W/Fire Protection Hose Stations Not Performed as Required by Ts.Caused by Inadequate Procedure Rev.Entered Action of LCO & Established Measures to Address Issue1996-03-15015 March 1996
- on 960215,SRs Associated W/Fire Protection Hose Stations Not Performed as Required by Ts.Caused by Inadequate Procedure Rev.Entered Action of LCO & Established Measures to Address Issue
05000327/LER-1996-001-02, :on 960121,discovered Fire Watch Patrol Did Not Patrol Some Assigned Areas in Control Bldg & on 960126,fire Watch Patrol Not Performed within Timeframe Required by Ts. Appropriate Disciplinary Action Taken1996-02-20020 February 1996
- on 960121,discovered Fire Watch Patrol Did Not Patrol Some Assigned Areas in Control Bldg & on 960126,fire Watch Patrol Not Performed within Timeframe Required by Ts. Appropriate Disciplinary Action Taken
05000328/LER-1992-008, :on 920627,reactor Tripped.Caused by Varying Resistance Readings Substandard Workmanship.Plant Instruction Written to Provide Guidance1996-02-0808 February 1996
- on 920627,reactor Tripped.Caused by Varying Resistance Readings Substandard Workmanship.Plant Instruction Written to Provide Guidance
1999-05-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEAR05000327/LER-1999-002-03, :on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event1999-10-15015 October 1999
- on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event
ML20217D2721999-10-12012 October 1999 Safety Evaluation Supporting Amends 248 & 239 to Licenses DPR-77 & DPR-79,respectively ML20217B3651999-10-0606 October 1999 Safety Evaluation Supporting Amends 247 & 238 to Licenses DPR-77 & DPR-79,respectively ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20217G3721999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Sequoyah Nuclear Plant.With ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20211N8891999-09-0707 September 1999 Safety Evaluation Supporting Amends 245 & 236 to Licenses DPR-77 & DPR-79,respectively ML20212C4761999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Sequoyah Nuclear Plant.With ML20212A1841999-08-25025 August 1999 Errata Pages for Rev 0 of WCAP-15224, Analysis of Capsule Y from TVA Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program ML20210L4361999-08-0202 August 1999 Cycle 9 12-Month SG Insp Rept ML20210L4451999-07-31031 July 1999 Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20216E3781999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20210G6631999-07-28028 July 1999 Cycle 9 90-Day ISI Summary Rept ML20209H3831999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Sequoyah Nuclear Plant.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20211F9031999-06-30030 June 1999 Cycle 9 Refueling Outage ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20195K2951999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-003-01, :on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status1999-05-27027 May 1999
- on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status
05000327/LER-1999-001-04, :on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc1999-05-11011 May 1999
- on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc
ML20206Q8951999-05-0505 May 1999 Rev 0 to L36 990415 802, COLR for Sequoyah Unit 2 Cycle 10 ML20206G3751999-05-0404 May 1999 Safety Evaluation Supporting Amends 244 & 235 to Licenses DPR-77 & DPR-79,respectively ML20206R5031999-04-30030 April 1999 Monthly Operating Repts for April 1999 for Sequoyah Units 1 & 2.With ML20205N0361999-04-12012 April 1999 Safety Evaluation Supporting Amend 234 to License DPR-79 ML20205P9811999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20204E8211999-03-16016 March 1999 Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively ML20205B6631999-02-28028 February 1999 Underground Storage Tank (Ust) Permanent Closure Rept, Sequoyah Nuclear Plant Security Backup DG Ust Sys ML20204C3111999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20203H7381999-02-18018 February 1999 Safety Evaluation of Topical Rept BAW-2328, Blended U Lead Test Assembly Design Rept. Rept Acceptable Subj to Listed Conditions ML20206U4331999-02-0909 February 1999 Safety Evaluation Supporting Amends 242 & 232 to Licenses DPR-77 & DPR-79,respectively ML20211A2021999-01-31031 January 1999 Non-proprietary TR WCAP-15129, Depth-Based SG Tube Repair Criteria for Axial PWSCC Dented TSP Intersections ML20198S7301998-12-31031 December 1998 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20199G3641998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-004-02, :on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check1998-12-21021 December 1998
- on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check
ML20198C0211998-12-16016 December 1998 Safety Evaluation Supporting Amends 241 & 231 to Licenses DPR-77 & DPR-79,respectively 05000327/LER-1998-003-04, :on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced1998-12-0909 December 1998
- on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced
ML20196J8021998-12-0707 December 1998 Safety Evaluation Supporting Amends 240 & 230 to Licenses DPR-77 & DPR-79,respectively ML20197J5621998-12-0303 December 1998 Unit 1 Cycle 9 90-Day ISI Summary Rept ML20197K1161998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20196C4091998-11-19019 November 1998 Safety Evaluation Supporting Amends 238 & 228 to Licenses DPR-77 & DPR-79,respectively ML20196B0231998-11-19019 November 1998 Safety Evaluation Supporting Amends 239 & 229 to Licenses DPR-77 & DPR-79,respectively ML20195H0831998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 226 to Licenses DPR-77 & DPR-79,respectively ML20195G3271998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 227 to Licenses DPR-77 & DPR-79,respectively 05000328/LER-1998-002-05, :on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure1998-11-10010 November 1998
- on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure
ML20195F8061998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Sequoyah Nuclear Plant.With ML20154H6091998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000328/LER-1998-001-05, :on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays1998-09-28028 September 1998
- on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays
ML20154H6251998-09-17017 September 1998 Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 Colr ML20153B0881998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Sequoyah Nuclear Plant.With 1999-09-07
[Table view] |
text
s Bl4 Tennessee Vaney Authonty, Post Office Hox 2000. Goddy-Daisy. Tennessee 37379-2000 Robert A. Fenech Vice President. Sequoyah Nuclear Plant July 14, 1993 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:-
TENNESSEE VALLEY AUTHORITY - SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 -
DOCKET NOS. 50-327 AND 50-328 - FACILITY OPERATING LICENSES DPR-77 AND DPR LICENSEE EVENT REPORT (LER) 50-327/93015 The enclosed LER provides details concerning the start of all four emergency diesel generators as the result of an incorrectly wired current transformer.
i This event is being reported in accordance with 10 CFR 50.73.b.2.iv as an event that resulted in the automatic actuation of an engineered safety feature.
Sincerely, Robert A. Fenech Enclosure cc: See page 2 i
1 9307220275 930714-
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PDR jADOCK 0500 7
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U.S. Nuclear Regulatory Commission Page 2 July 14,-1993 cc (Enclosure):
INP0 Records Center Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, Georgia 30339-5957 Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555-Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711
MRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 (6-89)
Empires 4/30/92 tICENSEE EVENT REPORT (tER)
FACIll*fY NAME (1) lDOCKETNUMBER(2) l_PAAGEW1_
_Sequoyah_!Mleitr Plant. Unit I lQlElll0jQ}L(Zj7 11]Dfl._Qj_6 TITLE (4)
_Di_eseLGenerator (D/G) St.RELR1_the Result of an Ipcorrectiv WiEed Current Tran1[ormer
_ENRLDALIS) l LER_t4 UMBER (6) l REPORT DATE (7) 1 OTHER_fACJ1111ES_lt0/DLVED_{8) l l
l l l SEQUENTIAL l l REVISION l l
l l FACILITY NAMES lDOCKETNUMBER(S)
[1QHlHj DAY jYEAR l YEAR l I NUM_9ER l l NUMBER IMONTHl DAY lYEAR I Secuoy & _ Unit 2 ID151010}H{3]Zja_
l l
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I I
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OPE RATING l lTHISREPORTISSUBMITTEDPURSUANTTOTHEREQUIREMENTSOF10CFR$:
MODE l l (Check one or more of the fo11owina)(11)
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IN l l20.402(b) l_l20.405(c) lKKl50.73(a)(2)(iv) l _.,l 73. 71 ( b) l POWER l l__l20.405(a)(1)(i) l_l50.36(c)(1) l_l50.73(a)(2)(v) l_l73.71(c)
LEVEL l l_l20.405(a)(1)(ii) l_l50.36(c)(2) l__l50.73(a)(2)(vii) l_l0THER(Specifyin
__.L10Llq_ja 10 l l20.405(a)(1)(iii)l_l50.73(a)(2)(i)
]_l50.73(a)(2)(viii)(A) l Abstract below and in l_l20.405(a)(1)(iv) l_l50.73(a)(2)(ii) l__l50.73(a)(2)(viii)(B) l Text, NRC Form 366A) l 120.405(a)(1)(v) 1 150.73(a)(2)(iii) l 150.73(a)(2)(x)
I LICENSEE CONTACT FOR THIS LER (12)
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A ABSTRACT (Limit to 1400 spaces, i.e., approximately fif teen single-space typewritten lines) (16)
On June 14, 1993, with Unit 1 defueled and Unit 2 in cold shutdown, Mode 5, the 1A start bus alternate feeder breaker tripped upon the start of the Unit I reactor coolant pump No. 1 motor. This resulted in the loss of voltage to the 1B-B 6.9 kilovolt shutdown board. The four emergency diesel generators (D/G) started and the 1B-B D/G supplied power to the 1B-B 6.9 kV shutdown board, as designed.
The event has been determined to have been caused by an incorrectly wired current transformer in the alternate feeder breaker protection circuitry. Plant equipment response during this event was consistent with that specified in the Final Safety Analysis Report for a loss of offsite power event.
I l
i4RC Form 366(6-89)
-,..U.S. NUCLEAR REGULATORY COMMISSION TApprovsd.0MB No. 3150-0104 (6-89)-
Exp1res 4/30/92
- - LICDeSEE EVDIT REPORT (LER) '
TEXT CONTINUATION FACILITY NAME (1) lDOCKETNUMBER(2)l LER NUMBER (6) l l
PAGE (31-
.l l-l l$EQUENTIAL;l l REVISION l l l.l l Sequoyah Nuclear Plant, Unit 1 l.
lYEAR I l NUMBER' -l
'l NUMBER l l l _ l,. l -
lDj5]A10j 913 12 17 19 13 l-l 0 l ' 1 1 5 l-l 0 1 0 l 01 2l0Fl 01 6' TEXT (If more space is required, use additional NRC Forn, 366A's) (17)
I.
PLANT CONDITIONS
4 Unit I was defueled and Unit'2 was in cold shutdown,- Mode 5.
II.
DESCRIPTION OF EVENT
A.
Event-On June 14,1993, the '1A start bus alternate feeder breaker '(EIIS Code EA).
tripped upon.the start of the Unit.1 reactor coolant pump-(RCP):No. 1 motor (EIIS Code AB). This resulted in the lossLof w itage to the 1B-B.6.9 kilovolt (kV) shutdown board (EIIS Code EB). 'The'four emergency-diesel. generators (D/Gs) (EIIS Code EK) started and the 1B-B D/G s'upplied power ~~to the11B-B.6.9-kV shutdown board,'as designed.
The 1A start bus had been transferred to the alternate feeder breaker. earlier
'the previous day as the result of maintenance.being performed on_the normal feeder breaker. A very light load existed on.the boards at that time as the result of the current-dual unit outage. On June 14, the, Unit 11 RCP No. 1-motor-was started af ter having been recently reinstalled from maintenance activities. Upon start of the RCP motor, a neutral-overcurrent' relay. actuated and, in turn, actuated the lock-out relay for the alternate feeder. breaker to the 1A start bus. As a result, the 1A and-1C 6.9 kV_ unit boards, as well as-the 1B-B shutdown board, experienced a loss of voltage and all four D/Gs started. The IB-B D/G tied on to the -1B-B shutdown board to ' supply power. - The-boards were verified to have no grounds, the RCP motor was not shorted, and there were no signs of damage to the shutdown board, unit boards, or start bus enclosure.
This event is being reported in accordance with 10 CFR 50.73.b.2.iv as an event that resulted in the automatic actuation of an engineered safety feature.
B.
Inapgrable Structures. Components or Systems That Contributed to the Event None.
C.
Dates and Approximate Times-of Maigr Occurrmacna October 15, 1992 Electrical Maintenance' discovered a significant crack on the "B" phase load current transformer (CT) for the 1A start bus alternate feeder breaker.
November 3, 1992 The cracked CT was replaced with a new CT from the same manufacturer.
NRC form 366(6-89)
f1RC Form 366A U.S. NUCLEAR REGULATORY COMMISSION Approvsd OMB No. 3150-0104 Expirss 4/30/92 (6-89)
+
LICENSEE EVENT REPORT (LER)
TEXT CONYINUATION Adll1YNAME(1) l000KETNUMBER(2)}
LIE _UUtfELR_16) l I
PAQi_{3) l l
l l SEQUENTIAL l l REVISION l l l l l Sequoyab Nuclear Plant, Unit 1 l
jlEARI I NQ@ER l l NUMBER I l l l~l IMMdd0l3 12 17 19 13 l-l 0 l 1 15l-l01 0101310f.}0l6 TLXT (If more space is required, use additional NRC Form 366A's) (17) l June 13, 1993 The 1A start bus is transferred to the alternate feeder breaker in order to perform maintenance on the normal feeder breaker.
l June 14, 1993 While attempting to start the Unit 1 RCP No. 1, the 1A start bus alternate feeder breaker tripped, causing a loss of voltage to the 1A and 1C unit' boards and the 1B-B shutdown board and subsequent start of all.four D/Gs.
l June 15, 1993 Investigation revealed that a CT on the "B" phase of the alternate feeder breaker was wired incorrectly. The CT had been replaced in November 1992. The replacement CT (same manufacturer) had the "X1" and "X2" positions in l
opposite locations from the original CT.
The i
electricians when replacing the CT and laid the leads down with the X1 and X2 wires in the same position as the original CT, not knowing the replacement CT was opposite the origit,a1 CT.
The CT wiring error went undetected i
until sufficient load existed (RCP *notor start) on the boards to create the phase imbalance that rerulted in the breaker trip. No records could be found that indicated j
the alternate feeder breaker had ever been placed in service until the day before this event occurred.
D.
Rther Systema _nr_S.econdary Functions Affrcted None.
l E.
Method of Discov_ery The D/G start was the result of the loss of voltage to the 1B-B shutdown board and was, thus, discovered by Operations personnel The incorrectly wired CT was discovered by Maintenance personnel investigating the cause of the 1A start bus alternate feeder breaker trip.
F.
QpSratDr Action Operations personnel, upon ensuring that a stable offsite power source existed, secured the three D/Gs that were not loaded to the shutdown boards. Once offsite power was restored to the 1A start bus, the final D/G was also secured.
G.
Safety Svatem_ Response The D/Gs started, as designed, upon the loss of voltage to the 1B-B shutdown board. The IB-B D/G properly tied onto the IB-B shutdown board and provided the power to the board until offsite power could be restored to the board.
NRC form 366(6-89)
iU.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
' TEXT CONTINUATION-j l
FACIllTY NAME (1) lDOCKETNUMBER(2)l LER NUMBER (6) l l
PAGE (3) l l
l l SEQUENTIAL l-l REVISION l l l l l Sequoyah Nuclear Plant, Unit 1 l
lYEAR l l NUMBER l l NUMBER l l l ll i
!Ol510[Qhj3 l2 17 19 13 l-l 0 l 1 1 5 l-l 0 l 0 l 01 410Fl 01 6 TEXT (If more space is required, use adfitional NRC form 366A's) (17) l III.
CAUSE OF EVENT
1 A.
Imediate Cause l
The immediate cause of this event was an' incorrectly wired CT on.the "B" phase i
of the 1A start bus alternate feeder breaker. - This caused the breaker to trip and resulted in.the loss of voltage to the 1B-B shutdown board.
l B.
Root CauSE The root cause of this event was the failure to uniquely identify.the CT secondary wiring in accordance with site procedures in order to ensure. correct-retermination of the wires. The individuals that. removed the. cracked CT indicated that the "B" phase CT control wiring was removed. Proper.
identification would have indicated that two uniquely identified wires were removed from the X2 pocition of the CT and one uniquely identified wire was removed from the X1' position.
C.
Conj;ributine Cauan A contributing cause was that the postmaintenance test (PMT) for the installation of the CT was ineffective'in detecting the wiring error. -This was the result of a lack of technical knowledge involving cts and a lack of l-specific guidance in site procedures concerning the proper PMT for a CT.
IV.
ANALYSIS OF EVENT
i One of the signals upon which the D/Gs are designed to start is a loss of voltage to the shutdown boards. The shutdown board, through the protective relays, will strip the load from the board, allow the D/G to come.up to speed, tie the D/G to the board, and then sequence the required loads back on the board..This is the sequence that occurred for this event. All safety-related equipment functioned as designed.
cts are designed to-monitor the current in a circuit. The cts used in-the 1A start bus alternate. feeder breaker are 4000/5-amp, wye-wired cts. The cts sum the current in all three. phases of.the circuit and send the resultant value to the breaker protective relays. The current total for a three-phase system should be zero for a normal circuit. However, with the "B" phase CT wired incorrectly, a current imbalance was sensed and a signal was sent to the protective relays to trip the 1A start bus alternate feedet breaker. Thus, with the CT wired incorrectly, the equipment functioned as designed.
Plant equipment response during this event was consistent with that specified in the Final Safety Analysis Report for a loss of offsite power event. Therefore, the event did not adversely affect the health and safety of the public.
i US Form 366(6-09) m
l HRC form 366A U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION fthklkHAME(1) lDOCKETNUMBER(2)}
LIR_tlUMBER (6) {
l PADE (3) l l
l l SEQUENTIAL l l REVISION l l l l l Sequoyah Nuclear Plant. Unit 1 l
jlEARl l_FUMBER I l NVtEIRJ l l l l IDMIE1DlD33d2 l7 l2_l3._l-l 0 l 1 l_5 l--! O l 0lOlljAFjJJJ_
TEXT (If more space is required, use additional NRC Form 366A's) (17)
V.
CORRECTIVE ACTION
i A.
IJInediate_ Correntive Action t
The immediate corrective action associated with this event included securing i
the three D/Gs that were not supplying power to the shutdown boards, restoring offsite power, and then securing the final D/G. An investigation was then l
I initiated to determine why the 1A start bus alternate feeder breaker tripped.
l Upon discovery of the incorrectly wired CT, all 6.9-kV cts that had been replaced were identified to ensure that proper installation had occurred. This i
investigation did not discover any further discrepancies. However, subsequent testing on a 6.9-kV common board CT revealed an incorrectly wired CT in the 2A
)
start bus normal feeder breaker circuitry.
Procedures were developed to j
perform phase testing on all 6.9-kV cts that had not been tested upon j
l replacement. The open work documents were also reviewed to ensure phase i
testing of any future replacement cts. The incorrectly wired cts were corrected.
l l
B.
Action to Prevent Recurrence The site procedure that governs configuration control of maintenance activities has been revised to clearly require unique identification of each configuration change.
Site Standard Practice (SSP) 6.31, " Maintenance Management System Pre-or Post-Maintenance Testing," will be revised to specify the proper PMT for CT replacement.
VI.
ADDITIONAL INFORMATION
A.
failesLcomponenta None.
B.
fERYinuS_Simi1ar Eventa A review of previous reportable events was conducted to identify any similar events. Several events were identified with similar causes, i.e., inadequate PMT, inattention to detail, and inadequate verification. Actions have been taken in response to previous events to ensure that management expectations were clearly conveyed, understood, and concurred with by site personnel.
It should be noted that the work for the replacement of the CT was planned in March 1992. This was before the specified corrective actions were in place.
Subsequent planning of work activities associated with CT replacement in February and again in April 1993 resulted in the proper PMT being specified.
N2C form 366(6-89)
.x
.1 NRC f orm 366A U.S. NUCLEAR REGULATORY COMMISSION Approvzd OMB No. 3150-0104 (6-89) s Expirss 4/30/92 LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACIl!TY NAME (1) lDOCKETNUMBER(2)l LEILN3tMDER (6) l l
PASE (3) l l
l l SEQUENTIAL l l REVISION l l l l l Sequoyah Nuclear Plant Unit 1 l
lyEARl I NUMBf1 l l NUMBER _l-l l l l 1015!0101013 12 17 19 13 1 - I o I 1 1 5 1--I o 1 0 I of 610rl of 6 TEXT (If more space is required, use additional NRC Form 366A's) (17)
VII.
COMMIIMENT SSP-6.31 will be revised by August 27, 1993, to ensure that the proper PMT for cts is specified.
NY form 366(6-89)
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05000328/LER-1993-001-04, :on 930301,extraction Steam Line Rupture Causes High Generator Output Voltage & Manual Reactor Trip.Caused by Programmatic Failure of Erosion/Corrosion Program.Insps, Repair & Replacements Will Be Performed |
- on 930301,extraction Steam Line Rupture Causes High Generator Output Voltage & Manual Reactor Trip.Caused by Programmatic Failure of Erosion/Corrosion Program.Insps, Repair & Replacements Will Be Performed
| | 05000327/LER-1993-001, :on 930124,Unit 1 Ice Bed Temperature Recorder in MCR Declared Inoperable.Caused by Ineffective Communication.Appropriate Personnel Involved W/Event Have Been Counseled |
- on 930124,Unit 1 Ice Bed Temperature Recorder in MCR Declared Inoperable.Caused by Ineffective Communication.Appropriate Personnel Involved W/Event Have Been Counseled
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-002-01, :on 930205,reactor Manually Tripped as Result of lock-up of Rod Control Sys.Caused by Problem in Rod Control Sys Circuitry.New I/O AC Amplifier Board Tested & Installed |
- on 930205,reactor Manually Tripped as Result of lock-up of Rod Control Sys.Caused by Problem in Rod Control Sys Circuitry.New I/O AC Amplifier Board Tested & Installed
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000328/LER-1993-002, Forwards Rev 1 of LER 93-002,providing Addl Info & Revised C/As Re Five Containment Boundary Isolation 1/2 Inch Drain Valves Found Improperly Configured.Event Originally Reported by ,as Operation Prohibited by TS | Forwards Rev 1 of LER 93-002,providing Addl Info & Revised C/As Re Five Containment Boundary Isolation 1/2 Inch Drain Valves Found Improperly Configured.Event Originally Reported by ,As Operation Prohibited by TS | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000328/LER-1993-002-03, :on 930318,during Integrity Surveillance,Five 1/2-inch Drain Valves Discovered to Be Misfigured.Caused by Valves Unlocked.Valves Placed in Correct Position & Secured |
- on 930318,during Integrity Surveillance,Five 1/2-inch Drain Valves Discovered to Be Misfigured.Caused by Valves Unlocked.Valves Placed in Correct Position & Secured
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-003, :on 930218,reactor Trip Occurred as Result of Exciter Field Breaker Opening & Turbine Tripped on Electrical Relay Protection Signal.Caused by Instructor Failing to Evaluate Risks.Disciplinary Action Taken |
- on 930218,reactor Trip Occurred as Result of Exciter Field Breaker Opening & Turbine Tripped on Electrical Relay Protection Signal.Caused by Instructor Failing to Evaluate Risks.Disciplinary Action Taken
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000328/LER-1993-003-01, :on 930321,containment Isolation Valve (EIIS Code Jm) Was Spuriously Actuated Resulting in Equipment Failure.Caused by Failure of RM Radiation Detector. Detector Replaced |
- on 930321,containment Isolation Valve (EIIS Code Jm) Was Spuriously Actuated Resulting in Equipment Failure.Caused by Failure of RM Radiation Detector. Detector Replaced
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000328/LER-1993-004-01, :on 930323,discovered That Surveillance Interval Specified in TS 4.6.5.1 Exceeded.Caused by Ineffective Communication Between Chemistry Personnel. Procedure Revised |
- on 930323,discovered That Surveillance Interval Specified in TS 4.6.5.1 Exceeded.Caused by Ineffective Communication Between Chemistry Personnel. Procedure Revised
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-004, :on 930222,determined That Blind Flange on Elevation 734 Personnel Airlock Outer Housing Leaking.Due to Improper Installation of Blind Flange.Evaluation Performed of Other 14 Double O-ring Blind Flanges |
- on 930222,determined That Blind Flange on Elevation 734 Personnel Airlock Outer Housing Leaking.Due to Improper Installation of Blind Flange.Evaluation Performed of Other 14 Double O-ring Blind Flanges
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000328/LER-1993-005, :on 931018,manual Actuations of RT Breakers Occurred Due to Demand Step Counter Malfunctions.Replaced Demand Step Counter |
- on 931018,manual Actuations of RT Breakers Occurred Due to Demand Step Counter Malfunctions.Replaced Demand Step Counter
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) | 05000327/LER-1993-005, :on 930320,containment Isolation Valve Inadvertently Actuated When Mechanic Inadvertently Shorted Test Leads After Verification of Equipment Setup.Caused by Blown Fuse.Test Leads Correctly Reconnected |
- on 930320,containment Isolation Valve Inadvertently Actuated When Mechanic Inadvertently Shorted Test Leads After Verification of Equipment Setup.Caused by Blown Fuse.Test Leads Correctly Reconnected
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000328/LER-1993-005-02, :on 931018,manual Actuations of RT Breakers Occurred Due to Malfunctioning.Demand Step Counters Manually Opened RT Breakers |
- on 931018,manual Actuations of RT Breakers Occurred Due to Malfunctioning.Demand Step Counters Manually Opened RT Breakers
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000328/LER-1993-006-01, :on 931203,reactor Tripped as Result of Generator Exciter Problems.Caused by Over Excitation of Generator.Exciter Replaced |
- on 931203,reactor Tripped as Result of Generator Exciter Problems.Caused by Over Excitation of Generator.Exciter Replaced
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-006, :on 930318,determined That Two Containment Isolation Valves in Fire Protection Sys Not Verified to Isolate Upon Receipt of Phase a Containment Signal.Caused by Use of Inadequate Procedure.Si Corrected |
- on 930318,determined That Two Containment Isolation Valves in Fire Protection Sys Not Verified to Isolate Upon Receipt of Phase a Containment Signal.Caused by Use of Inadequate Procedure.Si Corrected
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000328/LER-1993-007-02, :on 931203,reactor Trip Occurred as Result of Generator Exciter Problems.Caused by Overexcitation of Generator Due to Multiple Grounds in Generator Exciter. Exciter Replaced |
- on 931203,reactor Trip Occurred as Result of Generator Exciter Problems.Caused by Overexcitation of Generator Due to Multiple Grounds in Generator Exciter. Exciter Replaced
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-007, :on 930324,inadvertent Actuation of Containment Isolation Valve Occurred.Caused by Personnel Error. Individuals Involved Counseled & Operators Restored Affected Equipment to Svc |
- on 930324,inadvertent Actuation of Containment Isolation Valve Occurred.Caused by Personnel Error. Individuals Involved Counseled & Operators Restored Affected Equipment to Svc
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-008, :on 930326,determined That Flow Switches Controlling Operation of Emergency Gas Treatment Sys Decay Cooling Valves Not Calibr Since Apr 1993.Caused by Personnel Error.Instruction Revised |
- on 930326,determined That Flow Switches Controlling Operation of Emergency Gas Treatment Sys Decay Cooling Valves Not Calibr Since Apr 1993.Caused by Personnel Error.Instruction Revised
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-009, :on 930417,TS Surveillance Not Performed for Three Pipe Support Snubbers Because of Omission of Snubbers from Surveillance Instruction for Visual Insp.Snubbers Visually Inspected & Functionally Tested |
- on 930417,TS Surveillance Not Performed for Three Pipe Support Snubbers Because of Omission of Snubbers from Surveillance Instruction for Visual Insp.Snubbers Visually Inspected & Functionally Tested
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-010, :on 930430,Westinghouse Identified Error in Development of Calculations for Cold Overpressure Mitigation Sys Setpoints.Caused by Vendor Failure to Consider Elevation Difference.Engineering Evaluation Performed |
- on 930430,Westinghouse Identified Error in Development of Calculations for Cold Overpressure Mitigation Sys Setpoints.Caused by Vendor Failure to Consider Elevation Difference.Engineering Evaluation Performed
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-011, :on 930507,discovered That Fire Barrier Breached W/O Proper Compensatory Measures Established.On 930505,door Leading to Room Housing Containment Spray HX 1A Breached.Roving Fire Watch Established & LCO 3.7.12 Entered |
- on 930507,discovered That Fire Barrier Breached W/O Proper Compensatory Measures Established.On 930505,door Leading to Room Housing Containment Spray HX 1A Breached.Roving Fire Watch Established & LCO 3.7.12 Entered
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-012, :on 930504,apparent Failure to Properly Identify & Plug SG Tube Determined to Exceed TS Plugging Limit.Caused by eddy-current Coordinator Not Ensuring Task requirements.Eddy-current Procedure Revised |
- on 930504,apparent Failure to Properly Identify & Plug SG Tube Determined to Exceed TS Plugging Limit.Caused by eddy-current Coordinator Not Ensuring Task requirements.Eddy-current Procedure Revised
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-013, :on 930514,fire Watch Was Not Performed within Time Frame Required by Tech Specs Due to Inadequate Supervision by Fire Protection Foreman.Fire Watch Patrol Reestablished |
- on 930514,fire Watch Was Not Performed within Time Frame Required by Tech Specs Due to Inadequate Supervision by Fire Protection Foreman.Fire Watch Patrol Reestablished
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-014, :on 930611,determined That Inadequate Ventilation Design Resulted in Potential Inoperability of Vital Power Equipment.Design Being Modified |
- on 930611,determined That Inadequate Ventilation Design Resulted in Potential Inoperability of Vital Power Equipment.Design Being Modified
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) | 05000327/LER-1993-015, :on 930614,1A Start Bus Alternate Feeder Breaker Tripped Upon Start of Unit 1 RCP Which Resulted in Start of DG Due to Current Transformer Wired Incorrectly. Restored Offsite Power & Secured Final DG |
- on 930614,1A Start Bus Alternate Feeder Breaker Tripped Upon Start of Unit 1 RCP Which Resulted in Start of DG Due to Current Transformer Wired Incorrectly. Restored Offsite Power & Secured Final DG
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-016, :on 930619,Phase A,Auxiliary Bldg & Containment Isolations Manually Initiated as Result of Fuel Assembly Failing to Remain in Upright Position After Being Released. All Fuel Movement Stopped |
- on 930619,Phase A,Auxiliary Bldg & Containment Isolations Manually Initiated as Result of Fuel Assembly Failing to Remain in Upright Position After Being Released. All Fuel Movement Stopped
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000328/LER-1993-017, :on 930621,discovered 24-hour Telephone Notification Had Not Been Carried Out as Required by TS LCO 3.7.11.1 Action Statement (b)(2)(a) Due to Personnel Error. NRC Informed of Missed Notification |
- on 930621,discovered 24-hour Telephone Notification Had Not Been Carried Out as Required by TS LCO 3.7.11.1 Action Statement (b)(2)(a) Due to Personnel Error. NRC Informed of Missed Notification
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-018, :on 930704,DG Started Due to Improper WO Planning.Restored Power to 1BB Shutdown Board & Stopped Running DGs |
- on 930704,DG Started Due to Improper WO Planning.Restored Power to 1BB Shutdown Board & Stopped Running DGs
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-019, :on 930713,containment Isolation Occurred Due to Inaccurate Drawing Used to Develop Clearance Boundary. Returned Valves to Normal Positions & Restored Sys Air Pressure |
- on 930713,containment Isolation Occurred Due to Inaccurate Drawing Used to Develop Clearance Boundary. Returned Valves to Normal Positions & Restored Sys Air Pressure
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(viii)(A) | 05000327/LER-1993-020, :on 930719,Train B of Auxiliary Bldg Gas Treatment Sys Manually Initiated as Result of Fire in General Supply Fan Room.Sg Wet Layup Recirculation Pump Bearing & Motor Replaced |
- on 930719,Train B of Auxiliary Bldg Gas Treatment Sys Manually Initiated as Result of Fire in General Supply Fan Room.Sg Wet Layup Recirculation Pump Bearing & Motor Replaced
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-021, :on 930720,discovered That Boron Analysis of Refueling Canal Not Performed,As Required by TS 3.9.1 Due to Miscommunication.Lco 3.9.1 Entered.Chemistry Surveillance Instruction Will Be Revised |
- on 930720,discovered That Boron Analysis of Refueling Canal Not Performed,As Required by TS 3.9.1 Due to Miscommunication.Lco 3.9.1 Entered.Chemistry Surveillance Instruction Will Be Revised
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-026, :on 930917,determined That Design Basis Limit for Unqualified Coatings Inside Containment Had Been Exceeded.Caused by Inattention to Detail.Screens Will Be Installed at Openings on RCP Motor Stand |
- on 930917,determined That Design Basis Limit for Unqualified Coatings Inside Containment Had Been Exceeded.Caused by Inattention to Detail.Screens Will Be Installed at Openings on RCP Motor Stand
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(3)(ii) | 05000327/LER-1993-027, :on 930921,determined That Degraded Fire Dampers Condition Resulted from Failure to Install Damper in Accordance W/Design Drawings.Fire Watches Established in safety-related Plant Areas |
- on 930921,determined That Degraded Fire Dampers Condition Resulted from Failure to Install Damper in Accordance W/Design Drawings.Fire Watches Established in safety-related Plant Areas
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-028, :on 931104,identified Containment Isolation Valves Not Verified Closed or Secured Due to Inadequate Engineering Practices.Subj Plug & Vent Valves Verified Closed or Plugged by Operations Personnel |
- on 931104,identified Containment Isolation Valves Not Verified Closed or Secured Due to Inadequate Engineering Practices.Subj Plug & Vent Valves Verified Closed or Plugged by Operations Personnel
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) | 05000327/LER-1993-029, :on 931116,identified Inoperable Check Valves in Component Cooling Sys as Results of Buildup of Corrosion Product Between Valve Components.Valves Dissembled,Cleaned & Repaired as Needed |
- on 931116,identified Inoperable Check Valves in Component Cooling Sys as Results of Buildup of Corrosion Product Between Valve Components.Valves Dissembled,Cleaned & Repaired as Needed
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000327/LER-1993-030, :on 931221,failure to Perform Surveillance Requirements on Essential Raw Cooling Water Valve Identified.Caused by Personel Error.Valve Opened 35% & Power Removed from Valve |
- on 931221,failure to Perform Surveillance Requirements on Essential Raw Cooling Water Valve Identified.Caused by Personel Error.Valve Opened 35% & Power Removed from Valve
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(1) |
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