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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML18152B4411999-08-27027 August 1999 LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed ML18152B4421999-08-27027 August 1999 LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr ML18152B3771999-08-13013 August 1999 LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms ML18152B4181999-05-18018 May 1999 LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr ML18152B4111999-04-28028 April 1999 LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr ML18153A2741999-03-29029 March 1999 LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212 ML18153A2681999-03-19019 March 1999 LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr ML18152B7261999-01-21021 January 1999 LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable ML18152B5811998-12-16016 December 1998 LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B5781998-12-16016 December 1998 LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B7121998-12-0404 December 1998 LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened ML18152B7041998-12-0101 December 1998 LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr ML18152B6161998-11-0606 November 1998 LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable ML18152B6081998-10-23023 October 1998 LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely ML18152B7811998-07-31031 July 1998 LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status ML18153A2581998-06-0303 June 1998 LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed ML18152B8241998-05-22022 May 1998 LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed ML18152B7951998-04-29029 April 1998 LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6 ML18153A2521998-04-22022 April 1998 LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame ML18153A2511998-04-22022 April 1998 LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced ML18153A2391998-03-13013 March 1998 LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued ML18153A2341998-03-0909 March 1998 LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition ML18153A2301998-03-0606 March 1998 LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs ML18153A2251998-03-0404 March 1998 LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket ML18153A2201998-02-0606 February 1998 LER 98-001-00:on 980108,deficient Test Due to Faulty Test Equipment Resulted in TS Violation.Caused by Faulty Vibration Analyzer Cable or Loose Connection.Station Deviation Rept Was submitted.W/980206 Ltr ML18153A2071998-01-13013 January 1998 LER 97-012-01:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Two Breakers in Security Distribution Panel.Reset Affected Breakers Which Restored Power to Security Systems & Affected Doors ML18153A2101998-01-13013 January 1998 LER 97-009-01:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Caused by Inadequate Maint of Intake Canal Level Probes.Subject Probes Were Cleaned, Tested Satisfactorily & Returned to Operable Status ML18153A1911997-11-26026 November 1997 LER 97-011-00:on 971030,determined That Periodic Test Procedures for Testing Reactor Trip Bypass Breakers Did Not Test Manual Undervoltage Trip.Caused by mis-interpretation of Term in-service. Procedures Revised ML18153A1971997-11-26026 November 1997 LER 97-012-00:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Breaker in Security Distribution Panel in Central Alarm Station (CAS) Panel.Breakers in Affected CAS Panel Reset ML18153A1921997-11-25025 November 1997 LER 97-010-00:on 971028,discovered Missed Fire Protection Surveillance Pt.Caused by Personnel Error.Satisfactorily Completed PT Procedure 0-OPT-FP-009 & Diesel Driven Fire Pump 1-FP-P-2 Declared operable.W/971125 Ltr ML18153A1831997-11-12012 November 1997 LER 97-009-00:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Cause Indeterminate.Divers Inspected,Cleaned & Returned Probes to Operable Status & Initiated Interdepartmental Team to Investigate Cause ML18153A1791997-11-0707 November 1997 LER 97-008-00:on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset ML18153A1721997-10-30030 October 1997 LER 97-007-00:on 970930,determined That Plant Was Outside App R Design Basis Due to Vital Bus Isolation Issue.Caused by Personnel Error.Installed Circuit Protective Device During Oct 1997 Refueling Outage ML18153A1421997-06-10010 June 1997 LER 97-001-01:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation circuitry.W/970610 Ltr ML18153A1391997-05-28028 May 1997 LER 97-005-00:on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled ML18153A1291997-04-18018 April 1997 LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr ML18153A1281997-04-15015 April 1997 LER 97-004-00:on 970317,main Steam Safety Valve Was Outside as Found Setpoint Tolerance.Specific Cause Unknown,However, Minor Setpoint Drift Can Be Expected.No Immediate Corrective Actions performed.W/970415 Ltr ML18153A1241997-04-0808 April 1997 LER 97-002-01:on 970116,one Train of Auxiliary Ventilation Sys Was Inoperable Outside of Ts.Caused by Personnel Error. Submitted Deviation Rept Re Reverse Rotation of Fan & Work Request to Adjust linkage.W/970408 Ltr ML18153A1191997-03-19019 March 1997 LER 97-001-00:on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was replaced.W/970319 Ltr ML18153A1201997-03-19019 March 1997 LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr ML18153A1131997-02-20020 February 1997 LER 97-001-00:on 970123,shutdown Occurred Due to Steam Drain Line Weld Leak.Management Was Notified & Shift Supervisor Invoked Requirements of TS 4.15.C.1.W/undtd Ltr ML18153A1101997-02-13013 February 1997 LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr ML18153A0951997-01-0202 January 1997 LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown ML18153A0931996-12-12012 December 1996 LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr ML18153A0691996-09-19019 September 1996 LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr ML18153A0481996-08-26026 August 1996 LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr ML18153A0521996-08-20020 August 1996 LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr ML18153A0321996-07-30030 July 1996 LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr ML18153A0281996-07-17017 July 1996 LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr ML18153A0141996-07-0202 July 1996 LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr 1999-08-27
[Table view] Category:RO)
MONTHYEARML18152B4411999-08-27027 August 1999 LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed ML18152B4421999-08-27027 August 1999 LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr ML18152B3771999-08-13013 August 1999 LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms ML18152B4181999-05-18018 May 1999 LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr ML18152B4111999-04-28028 April 1999 LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr ML18153A2741999-03-29029 March 1999 LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212 ML18153A2681999-03-19019 March 1999 LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr ML18152B7261999-01-21021 January 1999 LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable ML18152B5811998-12-16016 December 1998 LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B5781998-12-16016 December 1998 LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B7121998-12-0404 December 1998 LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened ML18152B7041998-12-0101 December 1998 LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr ML18152B6161998-11-0606 November 1998 LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable ML18152B6081998-10-23023 October 1998 LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely ML18152B7811998-07-31031 July 1998 LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status ML18153A2581998-06-0303 June 1998 LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed ML18152B8241998-05-22022 May 1998 LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed ML18152B7951998-04-29029 April 1998 LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6 ML18153A2521998-04-22022 April 1998 LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame ML18153A2511998-04-22022 April 1998 LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced ML18153A2391998-03-13013 March 1998 LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued ML18153A2341998-03-0909 March 1998 LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition ML18153A2301998-03-0606 March 1998 LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs ML18153A2251998-03-0404 March 1998 LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket ML18153A2201998-02-0606 February 1998 LER 98-001-00:on 980108,deficient Test Due to Faulty Test Equipment Resulted in TS Violation.Caused by Faulty Vibration Analyzer Cable or Loose Connection.Station Deviation Rept Was submitted.W/980206 Ltr ML18153A2071998-01-13013 January 1998 LER 97-012-01:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Two Breakers in Security Distribution Panel.Reset Affected Breakers Which Restored Power to Security Systems & Affected Doors ML18153A2101998-01-13013 January 1998 LER 97-009-01:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Caused by Inadequate Maint of Intake Canal Level Probes.Subject Probes Were Cleaned, Tested Satisfactorily & Returned to Operable Status ML18153A1911997-11-26026 November 1997 LER 97-011-00:on 971030,determined That Periodic Test Procedures for Testing Reactor Trip Bypass Breakers Did Not Test Manual Undervoltage Trip.Caused by mis-interpretation of Term in-service. Procedures Revised ML18153A1971997-11-26026 November 1997 LER 97-012-00:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Breaker in Security Distribution Panel in Central Alarm Station (CAS) Panel.Breakers in Affected CAS Panel Reset ML18153A1921997-11-25025 November 1997 LER 97-010-00:on 971028,discovered Missed Fire Protection Surveillance Pt.Caused by Personnel Error.Satisfactorily Completed PT Procedure 0-OPT-FP-009 & Diesel Driven Fire Pump 1-FP-P-2 Declared operable.W/971125 Ltr ML18153A1831997-11-12012 November 1997 LER 97-009-00:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Cause Indeterminate.Divers Inspected,Cleaned & Returned Probes to Operable Status & Initiated Interdepartmental Team to Investigate Cause ML18153A1791997-11-0707 November 1997 LER 97-008-00:on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset ML18153A1721997-10-30030 October 1997 LER 97-007-00:on 970930,determined That Plant Was Outside App R Design Basis Due to Vital Bus Isolation Issue.Caused by Personnel Error.Installed Circuit Protective Device During Oct 1997 Refueling Outage ML18153A1421997-06-10010 June 1997 LER 97-001-01:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation circuitry.W/970610 Ltr ML18153A1391997-05-28028 May 1997 LER 97-005-00:on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled ML18153A1291997-04-18018 April 1997 LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr ML18153A1281997-04-15015 April 1997 LER 97-004-00:on 970317,main Steam Safety Valve Was Outside as Found Setpoint Tolerance.Specific Cause Unknown,However, Minor Setpoint Drift Can Be Expected.No Immediate Corrective Actions performed.W/970415 Ltr ML18153A1241997-04-0808 April 1997 LER 97-002-01:on 970116,one Train of Auxiliary Ventilation Sys Was Inoperable Outside of Ts.Caused by Personnel Error. Submitted Deviation Rept Re Reverse Rotation of Fan & Work Request to Adjust linkage.W/970408 Ltr ML18153A1191997-03-19019 March 1997 LER 97-001-00:on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was replaced.W/970319 Ltr ML18153A1201997-03-19019 March 1997 LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr ML18153A1131997-02-20020 February 1997 LER 97-001-00:on 970123,shutdown Occurred Due to Steam Drain Line Weld Leak.Management Was Notified & Shift Supervisor Invoked Requirements of TS 4.15.C.1.W/undtd Ltr ML18153A1101997-02-13013 February 1997 LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr ML18153A0951997-01-0202 January 1997 LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown ML18153A0931996-12-12012 December 1996 LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr ML18153A0691996-09-19019 September 1996 LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr ML18153A0481996-08-26026 August 1996 LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr ML18153A0521996-08-20020 August 1996 LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr ML18153A0321996-07-30030 July 1996 LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr ML18153A0281996-07-17017 July 1996 LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr ML18153A0141996-07-0202 July 1996 LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr 1999-08-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML18152A2811999-10-12012 October 1999 Technical Basis for Elimination of Nozzle Inner Radius Insps (for Nozzles Other than Reactor Vessel),Technical Basis for ASME Section XI Code Case N-619. ML18152B3531999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Surry Power Station,Units 1 & 2.With 991012 Ltr ML18152B6651999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Surry Power Station Units 1 & 2.With 990915 Ltr ML18152B4421999-08-27027 August 1999 LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr ML18152B4411999-08-27027 August 1999 LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed ML18151A3981999-08-13013 August 1999 SPS Unit 2 ISI Summary Rept for 1999 Refueling Outage. ML18152B3771999-08-13013 August 1999 LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms ML18152B3791999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Surry Power Station,Units 1 & 2.With 990811 Ltr ML18152B3911999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Surry Power Station,Units 1 & 2.With 990713 Ltr ML18152B4341999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Surry Power Station,Units 1 & 2.With 990614 Ltr ML20195E2401999-05-31031 May 1999 Rev 2 to COLR for SPS Unit 2 Cycle 16 Pattern Ag ML18152B4181999-05-18018 May 1999 LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr ML18152B4161999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Surry Power Station Units 1 & 2.With 990512 Ltr ML18152B4111999-04-28028 April 1999 LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr ML18152B6511999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Surry Power Station Units 1 & 2 ML18153A2741999-03-29029 March 1999 LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212 ML18153A2681999-03-19019 March 1999 LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr ML18152B7331999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Surry Power Station,Units 1 & 2.With 990310 Ltr ML18152B5421999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Surry Power Station,Units 1 & 2.With 990210 Ltr ML18151A3031999-01-29029 January 1999 ISI Summary Rept for 1998 Refueling Outage,Including Form NIS-1, Owners Rept for ISIs & Form NIS-2, Owners Rept for Repairs & Replacements. ML18152B7261999-01-21021 January 1999 LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable ML18152B6011998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Surry Power Station,Units 1 & 2.With 990115 Ltr ML18152B5781998-12-16016 December 1998 LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B5811998-12-16016 December 1998 LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition ML18152B7121998-12-0404 December 1998 LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened ML18152B7041998-12-0101 December 1998 LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr ML18152B7081998-11-30030 November 1998 Rev 0 to COLR for Surry 1 Cycle 16,Pattern Un. ML18152B5721998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Surry Power Station,Units 1 & 2.With 981214 Ltr ML18152B6161998-11-0606 November 1998 LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable ML18152B6241998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Surry Power Station Units 1 & 2.With 981111 Ltr ML18152B6081998-10-23023 October 1998 LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely ML18152B6881998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Surry Power Station Units 1 & 2.With 981012 Ltr ML18153A3271998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Surry Power Station,Units 1 & 2 ML18152B7811998-07-31031 July 1998 LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status ML18153A3161998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Surry Power Station Units 1 & 2.W/980807 Ltr ML18152B7621998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Surry Power Station,Units 1 & 2.W/980707 Ltr ML18153A2581998-06-0303 June 1998 LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML18153A3141998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Surry Power Station,Units 1 & 2.W/980610 ML18152B8241998-05-22022 May 1998 LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed ML18152B8161998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Surry Power Station Units 1 & 2.W/980508 Ltr ML18152B7951998-04-29029 April 1998 LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6 ML18153A2511998-04-22022 April 1998 LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced ML18153A2521998-04-22022 April 1998 LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame ML20217P9941998-04-0707 April 1998 Safety Evaluation Granting Licensee Third 10-yr Inservice Insp Program Relief Requests SR-018 - Sr-024 ML18153A2951998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Sps,Units 1 & 2.W/ 980408 Ltr ML18153A2391998-03-13013 March 1998 LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued ML18153A2341998-03-0909 March 1998 LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition ML18153A2301998-03-0606 March 1998 LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs ML18153A2251998-03-0404 March 1998 LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket 1999-09-30
[Table view] |
Text
. ACCELERATED DlrUTION DEMONS1\A.TION SYSTEM i* REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9109060144 DOC.DATE: 91/09/03 NOTARIZED: NO DOCKET#
FACIL: 50-.281 Surry Power Station, Unit 2; Virginia- Electric & Powe 05000281 AUTH.NAME AUTHOR AFFILIATION KANSLER,M.R. Virginia Power (Virginia Electric & Power Co.)
RECIP.N~ME RECIPIENT AFFILIATION ,
R
SUBJECT:
LER 91-007-00:on 910802,ESFactuated & safety injection & I
.reactor" trip occurred as result of high steam flow signal
,coincident w/low steam line pressure signal.Caused by loss D of voltage to protection instrumentation.W/910903 ltr. ~
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR _l ENCLr_._ SIZE: _ _'_/_ _ s TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.
I NOTES:lcy N11SS/IMSB/PM. 05000281 A
RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR. ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 D BUCKLEY,B 1 1 INTERNAL: ACNW 2 2 ACRS 2 2 s
AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP *2 2 *NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFBlO 1 1 NRR/DLPQ/LPEBlO 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPBll 2 2 NRR/DST/SELB SD 1 1 NRR/DST/SICB8H3 1 1
~:~ii~;~~
1 1 NRR/DST/SRXB SE 1
1 1 1 1 R,ES/DSIR/EIB 1 1 1 1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A "1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 R NOTES: 1 1 I D
s I
A D
D NOTE TO ALL "RIDS" RECIPIENTS:
s PLEASE HELP US TO REDUCE*WASTE! CONTACT THE DOCUMENT CONTROL DESK_,
ROOM Pl-37 (EXT. 20079) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34
Virginia Electric and Power Company Surry Power Station P. 0. Box 315 Surry, Virginia 23883 September_ 3, 1991 U. S. Nuclear Regulatory Commission Serial No~: 91-489 Document Control Desk Docket No.: 50-281 Washington, D. C. 20555 License No.: DPR-37 Gentlemen:
- -Pijrsuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Unit 2.
- REPORTNUMBER 91-007-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by the Corporate Management Safety Review Committee.
Very truly yours,
~n--=-rsler~
Station 1vlanager Enclosure cc: Regional Administrator Suite 2900 101 Marietta Street, NW Athmta, Georgia 30323 9109060144 910903 PDR ADOCK 05000281 S PDR
\
NRC FORM 366 (6-891 e U.S. NUCLEAR REGULATORY COMMISSION e APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE .TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-5301, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l' THE PAPERWORK REDUCTION 'PROJECT (3150-01041, OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME 111 IDOCKET NUMBER 121 .. . I PAGE 13)
Surry Power Station, Unit 2 o 15 Io Io Io 12 18 Id 1 loFb 16'
- 4 TITLE 1 1safety Injection/Reactor Trip From Vital Bus IVA Electrical Fault and Failure of Steam Generator Pressure Transmitter EVENT DATE (51 LER NUMBER (61 REPORT DATE 171
- OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR :\% SE~~t~i~AL t? ~t~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISl OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUlREMENTS OF 10 CFR §: {CMck one or more of the following/ (111-.
MODE (Bl N 20.402(bl 20.405(cl ,_ X 60,73(all2llivl 73.71(bl 20.405(*111 )Ill 50.311(cll1l 60.731all2lM 73.71(cl
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20.405(all1lliil 20.405(all1 llilil 50,38lcll2l 50.73(all2llil 60.73(all2llviil 60.73(al(2llviiil(Al
- 0TH ER {S~cify in Ab1tr*ct b*low *nd In* Toxt, NRC Form 366A/ .
- 20.405(all1 llivl 20.405(al111M
- 60,73(all2llill 50.731all2lliiil ---
LICENSEE CONTACT FOR THIS LER (121 60.731al(21(vilil(Bl 60.73(all2llxl NAME TELEPHONE NUMBER AREA CODE Mr.* M. R. Kansler_, Station Manager 81014 31c; 17 1-11111RIL.
COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MANUFAC . MANUFAC*
CAUSE SYSTEM COMPONENT , TURER TURER X SI B I I P 1T- R13 16 19 y I I I I *I I I I I .I I I I I I I I I *1 I I SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR EXPECTED SUBMISSION ITT YES /If yes. compf*te EXPECTED SUBMISSION DATE/
DATE 1151 111 115 911 ABSTRACT (Limit to 1400 spaces. i.tJ., approximately fiftetm sing/e-spactJ ryptJwritttJn lintJs) (16)
At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br /> on August 2, 1991, with Unit 1 at 100% power and Unit 2 at 92% power, an Engineered Safeguards Feature actuated in Unit 2. A safety injection/reactor trip occurred as a result of a high steam flow signal coincident with a low steam line press*ure signal. These conditions did not actually exist, but were generated by a combination of an erratic steam generator pressure channel and an electrical fault in a vital bus distribution panel. Troubleshooting was in progress on both problems, but the actuation occurred before corrective action could be completed. ESF functions performed as expected except fo.r the failure of Emergency Diesel Generator #3 to achieve rated speed and the failure of the Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) to fully shut. A Notification of Unusual Event (NOUE) was declared at 1710 hours0.0198 days <br />0.475 hours <br />0.00283 weeks <br />6.50655e-4 months <br />, and appropriate reports were made. On-shift operati.ng personnel promptly placed the plant in a stable condition. The NOUE was terminated at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />. A Root Cause Evaluation of the Emergency Diesel Generator problem is being conducted. The results of this evaluation will be reported in Licensee Event Report S.1-91-017 .. This event is reportable pursuant to 10CFR50.73(a)(2)(iv).
NRC Form 366 16-891
NRC FORM,366A U.S. NUCLEAR REGULATORY COMMISSION (6-89)
- APPROVED 0MB NO. :ii50-0104 EXPIRES: 4/30/92 .
- ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EV. REPORT (LER). INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE*
OF MANAGEMENT ANO BUDGET, WASHINGTON,_DC 20503.
FACILITY NAME (11 DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
Surry Power Station, Unit 2 o 1s Io Io Io 12 I s 11 .911 - o Io 11 - o I o o I 2 oF o 16 TEXT (ff mot9 -,,.C. la f'MIU.'rwl, UIIII MlditiofltJ! NRC Fann 3166A'a} (17)
LO DESCRIPTION OF THE EVENT At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br /> on August 2, 1991, with Unit 1 at 100% power and Unit 2 at 92% power, an Engineered Safety Feature (ESF) (EIIS-JE) actuated in Unit 2. A safety. injection/reactor trip occurred as a result of a high steam flow signal coincident with a low steam line pressure signal. These conditions did not actually exist, but were caused by an erratic steam generator pressure transmitter (02-MS-PT-2474) (EIIS-SB-PT) and an electrical fault in Vital Bus Distribution Panel IVA (EIIS-UJX-PL). Control and protection instrumentation .for steam flow, feed* flow, first stage turbine impulse pressure, steam generator (line) pressure, T-average control, pressurizer pressure control, steam dump control, and feed regulating valve "C" control are powered from the affected panel.
Problems were first noted at 1441 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.483005e-4 months <br /> when instrumentation powered by panel IVA began to experience perturbations.
- Concurrent with the problems with the vital bus IVA instrumentation, steam generator pressure channel II failed at 1506 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.73033e-4 months <br />. Corrective actions were initiated for the failed steam pressure transmitter, and a -
spare transmitter had been located, but work on replacement had not begun because of the continued voltage fluctuations on vital bus IVA.
The vital bus voltage fluctuations were under investigation at the distribution panel when the safety injection/reactor trip occurred. ESF functions performed as expected except for the failure of Emergency Diesel Generator #3 (EOG) (EIIS-EK-DG) to achieve rated speed (835 RPM vice a minimum acceptable speed of 870 RPM) and the failure of the Containment Sump Discharge Trip Valve (02-DA-TV-200A) (EIIS-BD-ISV) to fully shut. A Notification of Unusual Event (NOUE) was -declared at 171 O in accordance with the Emergency Action Level Table, Tab M, "Miscellaneous Abnormal Events", and appropriate reports were made.
Subsequent to the initial ESF actuation, several more actuations _
occurred as a result of continued spiking of the channel IV instrumentation coincident with an actual low value of Reactor Coolant System (RCS) average temperature*(< 543 degrees F). On-shift operating personnel acted promptly to place the plant in a stable condition. The NOUE was terminated at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />.
NRC Form 366A (6-89)
NRC FORM,366A
- U.S. NUCLEAR REGULATORY COMMISSION (6-89)
- APPROVED 0MB NO. 3150-0104 . * .
LICENSEE EVE.REPORT (LER) E-TEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD
- OOMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION
- AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1'HE PAPERWORK REDUCTION PROJECT (3160-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (21 . LER NUMBER (6) PAGE (31 Surry Power Station, Unit 2.
- TEXT (If mor. .,,.co m,-qul.TKI, UStl oddltianal NRC Form .BA :t) (171 2.0 SIG*NIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS The loss of a portion of one of the vital buses is of minor safety concern because of the redundancy of the power supplies for the different channels of control and protection instrumentation. Reactor protective schemes have redundant channels and the power sources are provided from redundant vital bus caqinets. Because of the fail-safe circuitry of the reactor protection instrumentation, a* power source failure to an instrument channel results in a trip signal from the affected channel.
Multiple power supplies are provided to prevent a single power supply failure from initiating a false trip. In this case, a problem occurred on a component powered from one vital bus concurrent with instrument perturbations caused by voltage fluctuations on a second vital bus satisfying the logic for safety injection even though actual plant conditions requiring a safety injection did not exist.* The system is designed so that if such a combination of events occurs, the Reactor
. Protective System (RPS) (EIIS-JC) acts automatically to place the plant in a safe condition.* If, for any reason, the RPS fails to function as designed, NRG-licensed operating personnel are trained to perform the necessary actions in accordance with the station's emergency operating procedures. During this event, ESF functions performed as expected except for the failure of EOG #3 to achieve rated speed and the failure of the Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) to fully shut.
Surry's emergency electric power system is designed to provide reliable power to engineered safety functions and other essential loads in the*
event of loss of off-site power (LOOP). The *system consists of three 100% capacity diesel generator sets for the two Units. One generator is used exclusively for Unit 1 (EOG #1 ), the second for Unit 2 (EOG #2), and the third (EOG #3) functions as a backup for either Unit. Each Unit has two emergency buses normally fed from independent off-site power sources, with the EDGs functioning as on-site backup power sources.
During this event, the off-site power sources remained available and fed the emergency busses. Although EDG #3 failed to achieve rated speed and would not have automatically loaded onto its bus, EOG #2 functioned as designed and could have carried its emergency bus had
- the need arisen. Also, existing training and procedures guide the operator to take manual control, raise spe.ed as necessary, and place EOG #3 on its bus.
NRC Form 366A (6-89)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION
(~-89) '. APPROVED 0MB NO. 3150-0104
- EXPIRES: 4/30/92 LICENSEE EVE. REPORT (LER). E-ATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING-BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
VEAR :::::::::: SEQUENTIAL *:*>:*:*:* REVISION Surry Power Station, Unit 2 :::::::::: NUMBER :-:*:*:*:* NUMBER 0 15 IO IO IO I 218 11 911 - 0 10 I 7 - 010 014 OF O I6 TEXT (ff mor. .,,ac. itJ m,ul,._,,, u.. *ddltional NRC Fam, 366A 'a) 1171 Because there was no actual plant condition requiring safety injection and plant conditions remained normal, the failure of the containment
- isolation valve (02-DA-TV-200A) to fully shut had no adverse affect. Also, the redundant isolation valve outside containment (02-DA-TV-200B) functioned properly.
The health and safety of the public were unaffected during the event.
3.0 CAUSE OF THE* EVENT The event was caused by the intermittent loss of voltage to channel IVA protection instrumentation and a _coincident low spiking of channel II -
steam generator pressure protection. The combination of these two conditions momentarily satisfied the high steam flow and low steam line pressure safety injection logic (two out of three high steam line flow signals coincident with two out of three low steam line pressure signals) and produced the ESF actuation .. The cause of the problem in vital bus channel IVA was an inadequate electrical connection at the panel's main breaker with evidence of arcing at the contacts. The suspected cause of the transmitter failure was an internal short causing the instrument to fai.1 high. Subsequently, the shorting condition cleared, producing a negative change in voltage sufficient to trip the low pressure bistable, and make up the safety injection logic matrix.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
Shift operating personnel placed the plant in a stable condition. The electrical connection to the input breaker to Vital* Bus Distribution Panel IVA was found. loose with intermittent contact occurring between the.
breaker and the vital bus panel bus bar. This connection was tightened and channel IVA instrumentation stabilized ..
5.10 ADDITIONAL CORRECTIVE ACTION($)
The Vital Bus Distribution Panel IVA input breaker was bench-tested satisfactorily and reinstalled. This breaker is a two-pole breaker with one pole in use at a time. The reconnection was made at the previously unused pole as a precaution. The three additional Unit 2 vital bus.panels which had been installed during the same plant modification were checked for evidence of a similar condition. They were found to be NRC Form 366A (6-89)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION
(~-89) .... - APPROVED 0MB NO. 3150-0104
.. EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE. REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 Surry Power Station, Unit 2 TEXT'" mom IIPI'"" i& mquimd, U/Jtl llddmona/ NRC Form .1151iA's) 1171 satisfactory. A Root Cause Analysis was initiated to investigate this event.
The failed pressure transmitter was replaced, and the defective instrument is being returned to the vendor for failure analysis.
EOG #3 was declared inoperable, and an. investigation into its performance was initiated. The engine's Woodward UG-8 governor (EIIS-DG-65) was found to be incorrectly set. The governor was returned to the proper setting, and the engine was tested and restored to operable status at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on August 3, 1991.
A separate root cause evaluation of the EOG #3 governor problem is being conducted by a team consisting of corporate and key station personnel. Licensee Event Report S1-91-017 will be submitted based on the results of this evaluation.
The Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) was binding due to excessive spring tension on its valve seat. The valve was repaired, tested, and returned to service at 2132 hours0.0247 days <br />0.592 hours <br />0.00353 weeks <br />8.11226e-4 months <br /> on August 5, 1991. A failure analysis is being conducted on this valve, and a supplement to this report will be submitted when the results of the analysis. have been received.
6.0 ACTIONS TO PREVENT RECURRENCE The vital bus distribution panel main breaker finger connectors and incoming feeds will be inspected and retorqued during the next refueling outage for each unit.
- Breaker maintenance proc~dures will be revised to include the manufacturers torque specifications for finger connectors.
The circuit breaker and associated switchgear preventive maintenance program will be revised to periodically inspect and retorque breaker electrical connections.
Any additional corrective action recommendations resulting from the Root Cause Evaluation will be evaluated and implemented as appropriate.
NRC Form 366A (6-89)
NRC FORM~366A U.S. NUCLEAR REGULATORY COMMISSION
\6.-89) ** ' APPROVEO 0MB NO. 3150-0104 EXPIRES: 4/30/92 .
LICENSEE EVE. REPORT (LER) ES, ..:.IATE.O BUR~EN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARO COMMENTS
- REGAROING BUROEN ESTIMATE TO THE RECOROS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, ANO TO 1'HE PAPERWORK REOUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT ANO BUOGET, WASHINGTON, OC 20503. '
FACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER 161 PAGE (31 Surry Power Station, Unit 2 o 1s I o Io I O ,2 1s 11 _9 11 _ o 10 I 7 _ o1o 016 oF o I 6 TEXT /ff mare _ . . la requ/1..J, U$t1 additiofll!! NRC Farm 35liA 's/ (17) 7 ..0 SIMILAR EVENTS Licensee Event Report S1 029-00, "Reactor Trip/Safety Injection Due to Spurious Hi CLS Signal as a Result of a Malfunctioning Relay".
8.0 ADDITIONAL INFORMATION Failed components:
Rosemount Transmitter (02-MS-PT-2474)
Model No. 1153GB9 NRC Form 366A 16-89)