Similar Documents at Surry |
---|
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
ION SYSTEM*
REGULA~ INFORMATION*DISTRIBUTIO~YSTEM (RIDS)
ACCESSION NBR:9202210408 DOC.DATE: 92/02/19 NOTARIZED: *No DOCKET #
FACIL:50-281 Surry Power Station, Unit 2, Virginia Elect~ic & Powe 05000281
,. AUTH.NAME AUTHOR AFFILIATION KANSLER,M.R.- Virginia Power (Virginia Electric & Power Co.)
RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 91-014-01:on 910724,determined that scaling for main ste~m flow transmitters incorrect,resulting -in setpoints in excess of max allowe~ value of 110%.Caused by use of incorrect methodolog~.Scaling program underway.W/920219 ltr.
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ! ENCL . / SIZE:
TITLE: 50. 7 3/50. 9 Licensee Event Report ( LER), IncidentRpt, et_c__.,.,""----
<5'.
NOTES:lcy NMSS/IMSB/PM. 05000281 RECIPIENT COPIES *RECIPIENT. COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2_PD 1 1 BUCKLEY,B 1 1 INTERNAL: ACNW 2 2 ACRS 2 2
. AEOD/DOA 1 1 .AEOD/DSP/TPAB 1 - 1 AEOD/ROAB/DSP 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFBlO 1 1 NRR/DLPQ/LPEBlO 1 1 NRR/DOEA/OEAB 1 1 NRR/DRE;P/PRPBll 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1
~l!IRR~L_SPLB8Dl 1 1 .NRR/DST/SRXB BE 1 1
~ 02 1 l RES/DSIR/EIB 1 1 RGN'2'=--***FILE 01 1 l.
EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PQR 1 1 NSIC MURPHY,G.A 1 1 J NSIC POORE,W. 1 1 NUDOCS .FULL TXT 1 1 NOTES: 1 1 I
l I
I NOTE TO ALL "RIDS" RECIPIENTS:
{
L 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 'I'EXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL 33
e e Virginia Electric and Power Company Surry Power Station
- P.O.Box315 Surry, Virginia 23883 Februatj'll, 199~
U.S. Nuclear Regulatory Commission Serial No.: 91-501A Document Control Desk Docket No.: 50-280 Washington, D. C. 20555 50-281 License No.: DPR-32 DPR-37
- Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following updated Licensee Event Report for Units 1 and 2.
REPORT NUMBER 91-014-01 This repo1-t has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed. by the Corporate Management Safety Review Committee.
Verytntly~
~ s l e1 r Stat~Manager Enclosure
- cc: Regional Administrator SuitE!2900 .
101 lv.larietta Street, NW Atlanta, Georgia 30323 9202210408 920219 PDR ADOCK 05000281 S PDR
e NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92
.MATED 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 IP-530), U.S. NUCLEAR
.. REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE I
OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACI.LITY NAME (1.1 DOCKET NUMBER (2) PAGET3J Surry Power Station, Unit 1 01s10101012 18 10 1 loF 014 TITLE 14) Main "Steam l:'l.OW .:::;etpoints in t:xcess or Technical. :::ipecirications Due to Flow Transmitter Scaling Inaccuracies Resulting From Incorrect Scaling Methodology EVENT DATE (!5) LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR /f) SEQUENTIAL NUMBER m REVlSION NUMBER MONTH DAY YEAR Surry, Unit 2 FACILITY NAMES DOCKET NUMBERISI 0 15 Io I o I o I 218 11 0 17 214 9 1 91 1 01 1 14
- 0 11 012 1 11 9 l2 0 1 s 1o IO I O I I I THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §:* (Check one or more of the following} (111 OPERATING MODE (9) N 20.402(b) 20.4051c) 50.73(o) (2)(iv) 73.71(b)
- 1-
. POWER LEVEL (10) Ii I o, 0
- 20.405(*ll1)(i) 20.4051oll1 llii) 50.38(cll1 I 50.38lcl(2) 50.73(oll2llv) 50.731oll21(vii)
I--
73.71 (cl OTHER (Specify in Abstr*ct lllill!iltll1lllli(tlll -,....._
20.406loll1 lliii) 20.405(oll1 )(iv) 20.4051*111llvl J
1---
50.731*ll2lli) 50.73(oll2llii) so.131,1121 om 1--
50.73(oll2)(viiil1Al 50.73(o)l2llvili) (Bl 50.731*ll2llxl btJlow ond in Texr, NRC Form 366A}
LICENSEE CONTACT FOR THIS LER 112)
NAME TELEPHONE NUMBER M. R. *Kansler, Station Manager AREA CODE 810 I 4 315 I 71- 13 I l 18 14 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS* REPORT 113) 1:,,:, *>:*:*::,:, ,',',',' .;,,:c:,:c:::::::
MANUFAC- REPORTABLE *.*.*.*. *:*: :-:,;,,.:;:;: MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT TO NPRDS :,;,:*:* *:*: ,',',',*:::: CAUSE SYSTEM COMPONENT TURER TURER TO NPRDS
.... *:*: ,;,;,;,;;,;, J!J::JJ::J:::::1:::::1:::1:i:11:::j:j:::i:1:1:::::J:
1,:,;.: :-::.; :::::;: :;:;;:: ;::::;: :::::;:::
,.. ,... *:*: ,',' .,., :::::1:::1::::1111::i::::::::::j:j:J::j:J::::i::::::
I I I I I I I ,,,; .*:* ,;,; I I I 1* I I I
. ,.,..;.; *ii:1::::JJ::jj!!::!:jjJjj!:jJljjj!J:::::1:::::1::::
I I I I I I I .......... I I I I I I I SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAY YEAR EXPECTED n ~
SUBMISSION DATE 1151 YES (If yes, comp/,-is EXPECTED SUBMISSION DATE} NO I I I ABSTRACT (Limi'. ro 1400 spaces. i.e .. approximately fifreBn single-sf)8ce typewritten lines) (16)
~
On July 24, 1991, with Unit 1 and Unit 2 operating at 100% power, it was determined after analyzing recent special test data that the scaling for the Unit 2 Main Steam (MS) flow transmitters was incorrect. These flow transmitters provide input to certain Reactor Protection System (RPS) and Engineered Safety Features (ESF) actuation signals. It was determined that the scaling resulted in Unit 2 setpoints in excess of the maximum allowed value of 110% (at full load) of full steam flow specified in Technical Specification (TS) 3.7.D. On July 22, 1991, voltage biases were placed on the uncompensated Unit 2 steam flow signals to reduce the Hi Steam Flow ESF function actuation to a value below 110%. This event was caused by the. use of incorrect scaling methodology, employed in 1977, to rescale the MS flow transmitters to correct an observed difference in the MS and feedwater flow indications. Since the same incorrect scaling information was used in 1977 for Unit 1 ; similar voltage biases were introduced on Unit 1 on July 24, 1991. Following .an evaluation of special test data, the Unit 2 steam flow and feedwater *flow transmitters were respanned and the steam flow circuitry rescaled. Independent testing is being performed to enable these measures to be implemented for Unit 1. The event is being reported, pursuant to 10 CFR 50. 73(a) (2)(i)(B).
NRC Form 366 (6-89)
L___
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) . APPROVED 0MB. NO, 3150-0104 EXPIRES: 4/30/92 LICENSEE EV. REPORT (LERI 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 COMMISSlbN, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY.NAME (1) DOCKET NUMBER (2) PAGE (3).
LEA NUMBER (6)
Surry Power Station, Unit 1 TEXT (If more SfJIICtJ is rec,*uired, use additional NRC Form 3!i'6A 's) (17) 1.10 DESCRIPTION OF THE EVENT On July 24, 1991, with Unit 1 and Uni.t 2 operating at 100% power, it was determined after analyzing recent special test data that the scaling for the Unit 2 Main Steam (MS) flow* transmitters {EliS-SB,FIT} was incorrect. These flow transmitters provide input to certain* Reactor Protection System (RPS) {EIIS-JC} and Engineered Safety Features (ESF) actuation {EIIS-JE} signals. It was determined that the installed scaling factors resulted in setpoints in excess of the maximum allowed value of 1 f0% (at full load) of. full steam flow specified in Technical Specification (TS) 3.7.D (Table 3.7-4, Item 5).
The special test began on July 4, 1991 and was governed by procedure 2-ST-302. The objective of the test was to collect data* at various power levels to verify the manufacturer's calibration curves *for feedwater flow venturis {EIIS-SJ} 2-FW-FE-2476, 2-FW-FE-2486, and 2-FW-FE-2496, which were installed during the Unit 2 Cycle 1O refu~ling outage. Data collected in this test
. was used to perform final scaling calculations for the feedwater and steam flow transmitters. The data collected during the special test was discussed by telephone wi.th NRC staff members on ~uly 12, 1991. *
- Following data reduction and analysis, it was determined that the Unit 2 high steam flow setpoint for ESF actuation was non-conservative due to incorrect scaling. It was also determined that the steam flow setpoint for the steam flow/feedwater flow mismatch reactor trip was non-conservative due to incorrect density compensation and incorrect scaling. The nonconservatism in the steam flow/feedwater flow mismatch reactor trip was insufficient to exceed the value for reactor trip protective instrumentation settings in Technical Specification 2.3.A.3.c.
The event is being reported, pursuant to 10 CFR 50.73(a)(2)(i)(B), since this*
condition is prohibited by TS 3.7.D (Table 3.7-4, Item 5).
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS MS flow signaJs are used as an input to two of the plant's protective functions.
High steam line flow coincident with either low Reactor Coolant System (RCS)
{EIIS-AB} *average temperature (Tave) or low steam line pressure results in a signal to the ESF protection logic, initiating Safety Injection (SI) and MS line isolation signals, Steam fiow is also an input to the RPS logic, initiating a
.reactor trip, if the system senses a steam flow/feedwater flow mismatch condition coincident with low steam generator {EIIS-SG} water level in any steam generator.
NRC For,y, 366A (6-89)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) -APPROVED OMS.NO. 3150-0104 .
- * - EXPIRES: 4/30/92 .
LICENSEE EV. REPORT (LERI -ATEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD
- COMMENTS REGARDING.BURD.EN- 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 (~) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
Surry Power Station, Unit 1 0 15 IO IO IO , 2 ,s IO 9 I 1 - o, 114 - 0 I 1 0 I3 OF O14 TEXT (ff more :,pa~ is mq,,ired, use tlddiriontll NRC Form 36tiA '*I (17)
The limiting steam line break analyses which take credit for the high steam flow channels are not sensitive to the steam flow setpoint. In the accident scenario, steam flow rapidly increases to a value well in excess of the setpoint. The Surry evaluation for small steam line breaks shows that break sizes in excess of 0.19 sq. -ft: per loop would exceed* the steam fiow setpoint. The limiting Surry break size is 1.4 sq. ft., well in excess of this value. Therefore, the limiting steam flow used in the accident analysis will exceed the setpoint by several hundred percent."
The recent review of scaling calculations indicated that the maximum error resu Its in a worst case steam flow of 115% vice 110% allowed by TS for generating the ESF signal. The Surry evaluation also. shows that, for smaller break sizes where the steam flow setpoint cannot be relied upon, adequate protection is provided by diverse channels (i.e.; high containment pressure, low RCS pressurizer {EIIS-PZR} pressure). --
The steam flow/feedwater flow mismatch reactor trip is a diverse, anticipatory trip which backs up the low-low steam generator level reactor protection -
function with one exception. The exception is a partial loss of feedwater due to failure of the steam generator level channel feeding the level. control system
{EIIS-LIC}. In this scenario, the two unaffected steam generators are adequate to remove the core heat until the reactor trip occurs on. low-low steam generator level in the affected steam generator, high RCS pressurizer level, or excessive difference between .RCS cold leg and hot leg temperatures (delta-T). In conjunction with the reactor trip, auxiliary feedwater is initiated on a low-low steam generator level signal prior to the onset of significant RCS heat-up. In -
addition, the nonconservatism in the steam flow/feedwater flow mismatch reactor trip was not sufficient to exceed the value for reactor trip protective instrumentation settings in -Technical Specification 2.3.A.3*.c.
Thus, the observed steam flow measurement uncertainties did not invalidate the acciderit analyses and the event has presented no adverse consequences to public -
health and safety.
3.0 CAUSE OF THE EVENT In 1977 the main steam flow transmitter spans were changed on both Unit 1 and Unit 2 using an inc9rrect scaling methodology. The respanning was undertaken to correct an observed difference in the values for steam flow and feedwater flow wherein the steam flow was reading higher than the feedwater flow. Based on
- evaluation of the special test 2-ST-302 data, it is now known that the difference in the steam flow and feed flow measurements observed in 1977 was due to an e*rror in the steam flow signal multiplier/divider scaling which caused an error in the steam flow indication ..
NRC Form 366A (6-89)
NRC F.ORM 366A' . U.S. NUCLEAR REGULATORY COMMISSION (6,89) . APPROVED 0MB NO. 3150-0104
- . . EXPIRES: 4/30/92
- LICENSEE EV * . REPORT (LERI - A T E D 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 THE PAPERWORK REDUCTION PROJECT (3150-01'04). OFFICE OF MANAGEMENT,AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)
Surry Powsr Station, Unit 1 o 1s .Io Io Io 12 1s Io 9 I 1 _ o 1114 _ o 1 1 o I 4 oF o 14 TEXT /ff more spa .. is nxiuired, u"" additional NRC Form 366A's/ (17)
- 4. O IMMEDIATE CORRECTIVE ACTION($)
On July 22, 1991 for Unit 2 and July 24, 1991 for Unit 1, voltage biases were placed on the uncompensated steam flow signals. Based on the data obtained from 2-ST-302, this action reduced the Hi Steam Flow ESF actuation on Unit 2 below 110%. Since the same incorrect scaling information was utilized on Unit 1 in 1977, the voltage biases were also placed on the Unit 1 signals to effectively reduce the 1977 scaling error, thereby reducing the Hi Steam Flow ESF actuation value. '
- 5. O ADDITIONAL* CORRECTIVE .ACTION(S)
Following an engineering evaluation of the special test data obtained from 2-ST-302, the Unit 2 steam flow and feedwater flow transmitters were respanned and the steam flow circuitry and associated Hi Steam Flow ESF actuation functions were rescaled in Septem_ber 1991. These actions corrected the steam flow/feedwater flow mismatch to acceptable values.
- The Unit 2 scaling effort was reviewed for applicability to Unit* 1. From this review, it was determined that Unit 1 testing would be required. Therefore,
. independent testing is being* performed for Unit 1 in conjunction with the 1992 refueling outage. The test results will be evaluated and, as appropriate, the Unit 1 . steam flow and feedwater flow transmitters will be respanned and the steam flow circuitry and associated Hi Steam Flow ESF actuation functions rescaled.
- 6. 0 ACTIONS TO PREVENT RECURRENCE A scaling program is being undertaken to verify the scaling for ESF and RPS instrument loops. The program includes the development of a standard scaling methodology and scaling procedure. The new scaling process will then be utilized to perform the calculations for an instrument loop. Any enhancements identified in performing the first calculation will be incorporated into the methodology.
The verification of the scaling for the ESF and RPS instrument loops will then be coordinated with the development of upgraded calibration pro.cedures being
- performed under the Technical Procedures Upgrade Program.
- 7. Qi SIMILAR EVENTS None
- 8. 0 MANUFACTURER/MODEL NUM.BER N/A NRC Form 366A (6-89)