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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR05000281/LER-1999-004-02, :on 981109,EDG Was Inoperable Longer than Allowed by TS Due to Governor Compensation Valve.Root Cause Evaluation Being Performed to Determine How Compensation Valve Became Closed1999-10-0101 October 1999
- on 981109,EDG Was Inoperable Longer than Allowed by TS Due to Governor Compensation Valve.Root Cause Evaluation Being Performed to Determine How Compensation Valve Became Closed
05000280/LER-1999-006, :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.With1999-08-27027 August 1999
- 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
05000280/LER-1999-005-01, :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 Installed1999-08-27027 August 1999
- 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
05000280/LER-1999-004-01, :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 Rooms1999-08-13013 August 1999
- 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
05000281/LER-1999-003-02, :on 990705,auto Reactor Trip on Low Coolant Flow,Occurred.Caused by Loop Stop Valve Failure.Approved RCE Recommendations,Designed to Prevent Recurrence of Similar Event Will Be Implemented Through CAP1999-07-30030 July 1999
- on 990705,auto Reactor Trip on Low Coolant Flow,Occurred.Caused by Loop Stop Valve Failure.Approved RCE Recommendations,Designed to Prevent Recurrence of Similar Event Will Be Implemented Through CAP
05000281/LER-1999-002-02, :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.With1999-05-18018 May 1999
- 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
05000280/LER-1999-003-01, :on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With1999-04-28028 April 1999
- on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With
05000281/LER-1999-001-02, :on 990301,RPS Relay Not Placed in Trip Resulted in Violation of TS 3.7.Caused by Lack of Procedural Guidance.Developed New Procedure to Provide More Explicit Instructions for Placing Stop Valve in Relay Trip1999-03-31031 March 1999
- on 990301,RPS Relay Not Placed in Trip Resulted in Violation of TS 3.7.Caused by Lack of Procedural Guidance.Developed New Procedure to Provide More Explicit Instructions for Placing Stop Valve in Relay Trip
05000280/LER-1999-002-01, :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 9902121999-03-29029 March 1999
- 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
05000280/LER-1998-013, :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 Ltr1999-03-19019 March 1999
- 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
05000280/LER-1999-001, :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 Acceptable1999-01-21021 January 1999
- 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
05000280/LER-1998-014, :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 Condition1998-12-16016 December 1998
- 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 05000280/LER-1998-012, :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.With1998-12-0101 December 1998
- 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
05000280/LER-1998-010, :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 Status1998-07-31031 July 1998
- 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
05000280/LER-1998-009, :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 Removed1998-06-0303 June 1998
- 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
05000280/LER-1998-008, :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 Needed1998-05-22022 May 1998
- 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
05000280/LER-1998-007, :on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-61998-04-29029 April 1998
- 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
05000280/LER-1998-006, :on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced1998-04-22022 April 1998
- on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced
05000280/LER-1998-005, :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 Frame1998-04-22022 April 1998
- 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
05000280/LER-1998-003, :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 Condition1998-03-0909 March 1998
- 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
05000280/LER-1998-004, :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 Repairs1998-03-0606 March 1998
- 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
05000280/LER-1998-002, :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 Socket1998-03-0404 March 1998
- 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
05000280/LER-1998-001-01, :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 Submitted1998-02-0606 February 1998
- 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
05000280/LER-1997-009, :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 Status1998-01-13013 January 1998
- 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
05000280/LER-1997-012, :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 Doors1998-01-13013 January 1998
- 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
05000281/LER-1997-004-02, :on 971202,invalid Mstv Indication Results in Manual Reactor Trip W/Esf Actuation Were Noted.Caused by Displaced Open Limit Switch Arms.Open Limit Switch for Mstv a Was Relocated Closer to Valve Position Bar1997-12-31031 December 1997
- on 971202,invalid Mstv Indication Results in Manual Reactor Trip W/Esf Actuation Were Noted.Caused by Displaced Open Limit Switch Arms.Open Limit Switch for Mstv a Was Relocated Closer to Valve Position Bar
05000281/LER-1997-002-01, :on 970713,main Steam High Range Radiation Monitor Was Declared Inoperable.Caused by Equipment Failure. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-61997-12-10010 December 1997
- on 970713,main Steam High Range Radiation Monitor Was Declared Inoperable.Caused by Equipment Failure. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6
05000280/LER-1997-011, :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 Revised1997-11-26026 November 1997
- 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
05000280/LER-1997-010, :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 Operable1997-11-25025 November 1997
- 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
05000281/LER-1997-003-02, :on 971014,Unit 2 MSSVs Revealed That Lift Setting for Two MSSVs Were Outside as Found Setpoint Tolerance.Caused by Minor Setpoint Drift.Repaired,Revised & Adjusted Safety Valves1997-11-13013 November 1997
- on 971014,Unit 2 MSSVs Revealed That Lift Setting for Two MSSVs Were Outside as Found Setpoint Tolerance.Caused by Minor Setpoint Drift.Repaired,Revised & Adjusted Safety Valves
05000280/LER-1997-008-01, :on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset1997-11-0707 November 1997
- 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
05000280/LER-1997-007-01, :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 Outage1997-10-30030 October 1997
- 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
05000281/LER-1997-002-03, :on 970713,CR annunciator,2-RMA-A-7 for Main Steam Line Effluent High Range Radiation Monitors Alarmed. Caused by Intermittent Component Failure.Preplanned Alternate Method of Monitoring Initiated1997-08-12012 August 1997
- on 970713,CR annunciator,2-RMA-A-7 for Main Steam Line Effluent High Range Radiation Monitors Alarmed. Caused by Intermittent Component Failure.Preplanned Alternate Method of Monitoring Initiated
05000280/LER-1997-001, :on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation Circuitry1997-06-10010 June 1997
- on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation Circuitry
05000280/LER-1997-005, :on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled1997-05-28028 May 1997
- on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled
05000280/LER-1997-006, :on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B1997-04-18018 April 1997
- 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
05000280/LER-1997-004, :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 Performed1997-04-15015 April 1997
- 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
05000280/LER-1997-002, :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 Linkage1997-04-0808 April 1997
- 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
05000281/LER-1997-001-01, :on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was Replaced1997-03-19019 March 1997
- on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was Replaced
05000280/LER-1997-003, :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 Open1997-03-19019 March 1997
- 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
05000280/LER-1997-002-01, :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 9701161997-02-13013 February 1997
- 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
05000281/LER-1997-002, :on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown1997-01-0202 January 1997
- 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
05000280/LER-1996-008-01, :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 Replaced1996-12-12012 December 1996
- 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
05000280/LER-1996-007, :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 Training1996-09-19019 September 1996
- 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
05000281/LER-1996-005-01, :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 Tubing1996-08-26026 August 1996
- 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
05000280/LER-1996-006, :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 Batteries1996-07-30030 July 1996
- 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
05000281/LER-1996-004-02, :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 Trees1996-07-0202 July 1996
- 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
05000280/LER-1996-004, :on 960510,noticed That Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies Due to Personnel Error.Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures Implemented1996-06-10010 June 1996
- on 960510,noticed That Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies Due to Personnel Error.Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures Implemented
05000281/LER-1996-003-01, :on 960512,Unit 2 Pressurizer Safety Valve as Found Lift Setting Out of Tolerance.Valve Was Reassembled & Lift Setting Was Established & Tested Satisfactorily1996-06-0707 June 1996
- on 960512,Unit 2 Pressurizer Safety Valve as Found Lift Setting Out of Tolerance.Valve Was Reassembled & Lift Setting Was Established & Tested Satisfactorily
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 05000281/LER-1999-004-02, :on 981109,EDG Was Inoperable Longer than Allowed by TS Due to Governor Compensation Valve.Root Cause Evaluation Being Performed to Determine How Compensation Valve Became Closed1999-10-0101 October 1999
- on 981109,EDG Was Inoperable Longer than Allowed by TS Due to Governor Compensation Valve.Root Cause Evaluation Being Performed to Determine How Compensation Valve Became Closed
ML18152B3531999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Surry Power Station,Units 1 & 2.With ML18152B3371999-09-24024 September 1999 SER Accepting Third 10-year Interval Inservice Insp Plan Request for Relief SR-026 for Surry Power Station Unit 2 ML18152B6651999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Surry Power Station Units 1 & 2.With 05000280/LER-1999-006, :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.With1999-08-27027 August 1999
- 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
05000280/LER-1999-005-01, :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 Installed1999-08-27027 August 1999
- 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
ML18152B3841999-08-23023 August 1999 Safety Evaluation Granting Relief Request from ASME Section Xa Requirements for Containment Insp ML18152B3631999-08-23023 August 1999 Safety Evaluation Supporting Eddy Current Techniques Used by VEPCO to Determine Depth of Degradation Evident in Units SG Tubing & VEPCO Approach for Dispositioning Tubes with Avb Wear Indications ML18152B3831999-08-23023 August 1999 Safety Evaluation Granting Relief Request from ASME Section XI Requirements for Containment Insp 05000280/LER-1999-004-01, :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 Rooms1999-08-13013 August 1999
- 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
ML18151A3981999-08-13013 August 1999 SPS Unit 2 ISI Summary Rept for 1999 Refueling Outage ML18152B3791999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Surry Power Station,Units 1 & 2.With 05000281/LER-1999-003-02, :on 990705,auto Reactor Trip on Low Coolant Flow,Occurred.Caused by Loop Stop Valve Failure.Approved RCE Recommendations,Designed to Prevent Recurrence of Similar Event Will Be Implemented Through CAP1999-07-30030 July 1999
- on 990705,auto Reactor Trip on Low Coolant Flow,Occurred.Caused by Loop Stop Valve Failure.Approved RCE Recommendations,Designed to Prevent Recurrence of Similar Event Will Be Implemented Through CAP
ML20196J4781999-07-0101 July 1999 Safety Evaluation Supporting Amends 221 & 221 to Licenses DPR-32 & DPR-37,respectively ML18152B3911999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Surry Power Station,Units 1 & 2.With ML20195D3571999-06-0707 June 1999 Safety Evaluation Supporting Amends 220 & 220 to Licenses DPR-32 & DPR-37,respectively ML20195E2401999-05-31031 May 1999 Rev 2 to COLR for SPS Unit 2 Cycle 16 Pattern Ag ML18152B4341999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Surry Power Station,Units 1 & 2.With 05000281/LER-1999-002-02, :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.With1999-05-18018 May 1999
- 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
ML18152B4161999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Surry Power Station Units 1 & 2.With 05000280/LER-1999-003-01, :on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With1999-04-28028 April 1999
- on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With
ML18152B6481999-04-14014 April 1999 Safety Evaluation Supporting Relief Requests IWE-2,4.5.6 & IWL-2 to Licenses DPR-32 & DPR-37 Respectively ML18152B6451999-04-13013 April 1999 SER Accepting Util Reactor Pressure Vessel Fluence Methodology for Surry Power Stations,Units 1 & 2 & North Anna Power Station,Units 1 & 2 Subject to Listed Limitations 05000281/LER-1999-001-02, :on 990301,RPS Relay Not Placed in Trip Resulted in Violation of TS 3.7.Caused by Lack of Procedural Guidance.Developed New Procedure to Provide More Explicit Instructions for Placing Stop Valve in Relay Trip1999-03-31031 March 1999
- on 990301,RPS Relay Not Placed in Trip Resulted in Violation of TS 3.7.Caused by Lack of Procedural Guidance.Developed New Procedure to Provide More Explicit Instructions for Placing Stop Valve in Relay Trip
ML18152B6511999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Surry Power Station Units 1 & 2 05000280/LER-1999-002-01, :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 9902121999-03-29029 March 1999
- 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
05000280/LER-1998-013, :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 Ltr1999-03-19019 March 1999
- 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
ML20207L8081999-03-12012 March 1999 Safety Evaluation Supporting Amends 219 & 219 to Licenses DPR-32 & DPR-37 ML18152B7331999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Surry Power Station,Units 1 & 2.With ML18152B5381999-02-16016 February 1999 SER Accepting Third 10-year Interval Inservice Insp Request for Relief for Surry Power Station,Unit 1.Staff Concludes That Licensee Proposed Alternative Will Provide Acceptable Level of Quality & Safety.Technical Ltr Rept Also Encl ML18152B5421999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Surry Power Station,Units 1 & 2.With 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 05000280/LER-1999-001, :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 Acceptable1999-01-21021 January 1999
- 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 ML18152B5861998-12-18018 December 1998 SER Approving Request Relief Related to Inservice Testing Program at Surry Power Station Unit 1 ML20198F9221998-12-16016 December 1998 Safety Evaluation Supporting Amends 217 & 217 to Licenses DPR-32 & DPR-37,respectively 05000280/LER-1998-014, :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 Condition1998-12-16016 December 1998
- 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
ML18152B5901998-12-16016 December 1998 Safety Evaluation Authorizing Request to Use Code Case N-577 as Alternative to Requirements of ASME Code Section XI for Surry Power Station,Unit 1 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 05000280/LER-1998-012, :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.With1998-12-0101 December 1998
- 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
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 ML18152B6241998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Surry Power Station Units 1 & 2.With ML18152B6881998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Surry Power Station Units 1 & 2.With ML20151U7261998-09-0303 September 1998 Safety Evaluation Approving Exemption from Certain 10CFR20 Requirements Re Use of self-contained Breathing Apparatus with Enriched Oxygen in Subatmospheric Containments at SPS ML18153A3271998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Surry Power Station,Units 1 & 2 ML20237E9721998-08-26026 August 1998 Safety Evaluation Supporting Amends 216 & 216 to Licenses DPR-32 & DPR-37,respectively ML18153A3161998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Surry Power Station Units 1 & 2 05000280/LER-1998-010, :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 Status1998-07-31031 July 1998
- 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
1999-09-30
[Table view] |
text
e CATEGORY Je REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
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ACCESSION NBR:9605200389 DOC.DATE: 96/05/07 NOTARIZED: NO DOCKET#
FACIL:50-281 Surry Power Station, Unit 2, Virginia Electric & Powe 05000281 AUTH.NAME AUTHOR AFFILIATION CHRISTIAN,D.A.
Virginia Power (Virginia Electric & Power Co.)
RECIP.NAME RECIPIENT AFFILIATION SUBJECT: LER 96-002-00:on 960408,EDG fire suppression sys declared inoperable due to personnel error.Submitted station Deviation Rept.W/960507 ltr.
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL J SIZE:
lo TITLE: 50.73/50.9 Licensee Event Report (LER),--Yncident Rpt, etc.
NOTES:
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NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-S(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27 1
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~7, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555
Dear Sirs:
e 10CFR50.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 Serial No.:
96-246 SPS:BAG Docket No.: 50-281 License No.: DPR-37 Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 2.
REPORT NUMBER 50-281 /96-002-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.
Very truly yours,
~300_
D. A. Christian Station Manager Enclosure cc:
Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 M. W. Branch NRG Senior Resident Inspector Surry Power Station 9605200389 960507 PDR ADOCK 05000281 S
PDR
e e
NRC FC9M 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5-92)
EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSING EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO (See reverse for required number of digits/characters for each block)
THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
ER (2)
~(3)
SURRY POWER STATION, Unit 2 05000 - 281 F 5 TITLE (4)
Inoperable EOG Fire Suppression Svstem Due to Personnel Error EVENT DATE 51 LEA NUMBER (6 REPORT DATE (71 OTHER FACILITIES INVOLVED (8}
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 05000
-- 002 --
00 05 07 96 FACILITY NAME DOCKET NUMBER 05000
- OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR:(Check one or more) (11)
MODE (9)
N 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(c)
POWER 20.405(a)(1 )(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
LEVEL (10) 100%
20.405(a)(1 )(ii) 50.36(c)(2) 50.73(a)(2)(vii)
OTHER 20.405(a)(1 )(iii)
X 50.73(a)(2)(i) 50.73(a)(2)(viii)(A)
(Specify in Abstract below and 20.405(a)(1 )(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) in Text, N RC Form 366A) 20.405(a)(1 )(v) 50.73(a)(2)(iii) 50.73(a)(2}(x)
LICENSEE CONTACT FOR THIS LER 12)
NAME I (~~4r3~;~; ~184ing Area Code)
D. A. Christian, Station Manager COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113)
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TONPRDS N
SUPPLEMENTAL REPORT EXPECTED 14)
I EXPECTED I MONTH I DAY I YEAR I YES (If ves comolete EXPECTED SUBMISSION DATE\\
XI NO I
SUBMISSION DATE (15)
I I
I ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On April 8, 1996, at approximately 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br />, with Unit 1 and Unit 2 operating at 100%, the rear exit door to Emergency Diesel Generator (EOG) room number 2 was found open.
The door was verified to be closed at approximately 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />.
Personnel exit through this door after approximately 1135 hours0.0131 days <br />0.315 hours <br />0.00188 weeks <br />4.318675e-4 months <br /> allowed a welding lead to become caught between the door and its doorjamb. This condition, which limited the ability of the EOG number 2 carbon dioxide fire suppression system to perform its intended function, had existed for a period in excess of that allowed by Technical Specification (TS) 3.21.8.4. At 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br />, Operations personnel entered EOG room number 2 to remove the welding lead from the doorjamb and close the door. The cause of the event was cognitive personnel error in failing to ensure that the door was closed upon exit. To prevent recurrence, construction management reviewed the event with craft involved in the construction work and the construction foreman. The doors were posted to restrict egress. The event resulted in negligible safety consequences since no fire had occurred, the condition existed for a short duration, and the swing EOG (number 3) was fully operable. Therefore, the health and safety of the public was not affected. This event is being reported pursuant to 1 OCFR50.73(a)(2)(i)(B), since the condition was prohibited by TS 3.21.B.4.
NRG FORM 366 (5*92)
NRG FORM 366 (5-92)
U.S. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
FACILITY NAME 111 DOCKET NUMBER 121 SURRY POWER STATION, Unit 2 05000 - 281 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
LEA NUMBER 161 PAGE (3).
YEAR SEQUENTIAL NUMBER REVISION NUMBER 96
-- 002 --
00 2~0F 5 Emergency Diesel Generator (EOG) [EIIS-DG] room number 2 doors are posted as Carbon Dioxide Boundary Fire Doors.
Blocking or holding the doors open is not permitted and Operations Shift Supervisor permission is required prior to putting the door on its automatic blow off device. Access into EOG number 2 is provided from the Unit 2 turbine building hallway and is restricted by the use of a keycard. The EOG number 2 rear exit door leads to the Unit 2 alleyway and provides no access from the outside.
During recent construction work, welding leads were supplied from the Unit 2 alleyway and supported along the outside wall above the rear exit door to EOG number 2. To support the tie-in of new EOG fuel oil supply lines, the welding leads were run through the doorway into EOG room number 2. While the door was open, a fire watch was posted as a compensatory measure to comply with Technical Specification (TS}
3.21.B.4. On April 8, 1996, following the completion of construction work, the welding leads were removed from the doorway and coiled outside the rear door for temporary storage. The door was closed and the Operations Shift Supervisor was notified. The fire watch was released at approximately-0850 hours.
At approximately 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />, the rear door to the EOG room number 2 was verified to be closed by Safety and Loss Prevention personnel. Another entry was made into the EOG room number 2 at 1135 hours0.0131 days <br />0.315 hours <br />0.00188 weeks <br />4.318675e-4 months <br /> by construction personnel and the individual exited the rear exit door at approximately 1145 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.356725e-4 months <br />.
When interviewed, the individual indicated he saw the door closing but did not verify that it was fully shut. It is believed that the welding leads supported above the rear exit door had loosened and, upon the exit, the leads fell between the door and doorjamb.
Other entries were made into the EOG room number 2 after this exit, however, none of the personnel exited through the rear exit door. At 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br />, an entry was made into EOG room number 2 and the rear exit door was found open approximately 2-3 inches.
Operations was notified and at 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br /> an operator entered EOG room number 2 to remove the welding leads and close the door.
NRC FORM 366 (5-92)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95.
LICENSING EVENT REPORT {LER)
(See reverse for required number of digits/characters for each block)
ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555-0001. AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
DOCKET NUMBER 12)
LEA NUMBER 16)
PAGEl3l YEAR SEQUENTIAL NUMBER REVISION NUMBER SURRY POWER STATION, Unit 2 05000 - 281 96
-- 002 --
00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) This event is being reported pursuant to 1 OCFR50. 73(a)(2)(i)(B), conditions prohibited by plant Technical Specifications, since the open rear exit door to EDG number 2 limited the ability of the carbon dioxide fire suppression system to perform its intended function and a continuous fire watch was not established within one hour.
2.0 SAFETY CONSEQUENCES AND IMPLICATIONS
The EDG number 2 room has a heat detection system [EIIS-IC] which alarms in the control room and a manually actuated total-flooding carbon dioxide fire suppression system [EIIS-KQ]. In addition, backup fire fighting capability is provided by portable extinguishers in the diesel generator room and manual hose stations serving this area are provided in the turbine building.
Had a fire occurred during this period, the heat system would have alarmed in the control room and an operator would have responded to investigate and actuate the fire suppression system. The control room would have also initiated a response by the station's fire brigade to provide additional suppression capability to backup the carbon dioxide fire suppression system.
This event resulted in negligible safety consequences and implications since a fire did not occur, the condition existed for a short duration, and the swing EDG (number 3) was available. Therefore, the health and safety of the public were not affected.
3.0 CAUSE
The cause of this event is attributed to cognitive personnel error on the part of the personnel who secured the welding leads above the EDG room number 2 rear exit door and by the individual who exited the door without ensuring that the door was fully closed. The rear exit doors are labeled as carbon dioxide boundary fire doors and instructions on the door state that the door is not to be blocked or held open. Planning and implementation of the EDG number 2 construction work recognized that the doors 3oF5
e NRG FORM 366 (5-92)
U.S. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
FACILITY NAME (1)
DOCKET NUMBER (2)
SURRY POWER STATION, Unit 2 05000 - 281 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 3.0 CAUSE (continued}
APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REVISION NUMBER 96
-- 002 --
00 4 OF 5 could not be blocked open without compensatory actions and all individuals interviewed concerning this event were aware of the requirements. The error leading to this event was created when the rear exit door was used as an egress doorway. No alarms are provided if the door is not secured in the closed position.
4.0 IMMEDIATE CORRECTIVE ACTIONS
.upon discovering that the rear exit door was not closed, the control room was notified and an operator responded to the EOG room number 2 to remove the welding leads from the doorway and close the door.
A station Deviation Report was submitted.
5.0 ADDITIONAL CORRECTIVE ACTIONS
The welding leads were removed from the area above the EOG room number 2 rear exit door. Construction management reviewed the event with personnel involved with EOG fuel oil line replacement construction work in the EOG rooms, the pipe fitter craft, and the construction foreman. This review stressed the requirement to maintain the carbon dioxide boundary fire door closed.
During the fuel oil line replacement construction work, no restrictions were placed on personnel exiting through the EOG rear exit doors. To reduce the risk of recurrence, the rear exit doors were posted as emergency exit only.
NRG FORM 366 (5-92)
. U.S. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
FACILITY NAME 111 DOCKET NUMBER 121 SURRY POWER STATION, Unit 2 05000 - 281 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 6.0 ACTION TO PREVENT RECURRENCE APPROVED BY 0MB NO. 3150*0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.
LER NUMBER 161 YEAR SEQUENTIAL NUMBER 96
.. 002..
REVISION NUMBER 00 PAGEl31 50F5 Prior to the fuel oil line construction work, personnel egress through the EOG room rear exit doors was prohibited. To expedite the fuel oil line replacement work, restrictions on personnel access were relaxed. Upon completion of the work, restrictions on personnel egress through the EOG room rear exit doors will be re-established.
7.0 SIMILAR EVENTS
91-021-00 Emergency Diesel Generator Number 1 Room Fire Suppression System Inoperable Due to Personnel Error in Administratively Controlling the Exit Door.
8.0 MANUFACTURER/MODEL #
None
9.0 ADDITIONAL INFORMATION
None
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05000280/LER-1996-001, :on 951213,ESW Pump Was Inoperable Due to Loss of Missile Protection for Piping.Revised DCP 91-025 |
- on 951213,ESW Pump Was Inoperable Due to Loss of Missile Protection for Piping.Revised DCP 91-025
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) | 05000281/LER-1996-001-02, :on 960222,through-wall Leak Identified in Unit 2 RHR Piping.Caused by General Intergranular Attack on Inside Surface of Piping.Conservative Leakage Rate of One Drop Every Ten Minutes Estimated |
- on 960222,through-wall Leak Identified in Unit 2 RHR Piping.Caused by General Intergranular Attack on Inside Surface of Piping.Conservative Leakage Rate of One Drop Every Ten Minutes Estimated
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) | 05000280/LER-1996-002-01, :on 960303,containment Isolation Valve Inoperable.Caused by Personnel Error.C/A:Individuals Counseled |
- on 960303,containment Isolation Valve Inoperable.Caused by Personnel Error.C/A:Individuals Counseled
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | 05000281/LER-1996-002-02, :on 960408,EDG Fire Suppression Sys Declared Inoperable Due to Personnel Error.Submitted Station Deviation Rept |
- on 960408,EDG Fire Suppression Sys Declared Inoperable Due to Personnel Error.Submitted Station Deviation Rept
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000281/LER-1996-003-01, :on 960512,Unit 2 Pressurizer Safety Valve as Found Lift Setting Out of Tolerance.Valve Was Reassembled & Lift Setting Was Established & Tested Satisfactorily |
- on 960512,Unit 2 Pressurizer Safety Valve as Found Lift Setting Out of Tolerance.Valve Was Reassembled & Lift Setting Was Established & Tested Satisfactorily
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) | 05000280/LER-1996-003-02, :on 960513,control Room Air Handling Units (AHU) Declared Inoperable.Caused by Mechanical Failure. Adjusted 1-VS-AC-2 Backdraft Dampers Counterweight Arm for AHU 1-VS-AC-2 |
- on 960513,control Room Air Handling Units (AHU) Declared Inoperable.Caused by Mechanical Failure. Adjusted 1-VS-AC-2 Backdraft Dampers Counterweight Arm for AHU 1-VS-AC-2
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000280/LER-1996-004, :on 960510,noticed That Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies Due to Personnel Error.Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures Implemented |
- on 960510,noticed That Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies Due to Personnel Error.Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures Implemented
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000281/LER-1996-004-02, :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 |
- 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000280/LER-1996-005, :on 960506,charging Pumps Declared Inoperable. Caused by Air Entering Service Water Sys Through Valve 2-SW-MOV-201B.Personnel Closed Valve 2SW-MOV-201B |
- on 960506,charging Pumps Declared Inoperable. Caused by Air Entering Service Water Sys Through Valve 2-SW-MOV-201B.Personnel Closed Valve 2SW-MOV-201B
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) | 05000281/LER-1996-005-01, :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 |
- 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) | 05000280/LER-1996-006, :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 |
- 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
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) | 05000280/LER-1996-007, :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 |
- 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000280/LER-1996-008-01, :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 |
- 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
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) |
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