|
---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20045E9961993-07-0101 July 1993 LER 93-012-00:on 930607,discovered That Surveillance Test on Fire Protection Sys Missed on 930601.Caused by Insufficient Degree of Attention Applied by Nonlicensed Individual.Test Satisfactorily completed.W/930701 Ltr ML20045E8661993-06-25025 June 1993 LER 93-011-00:on 930528,TS Violation Occurred When New Fuel Was Added to EDG Fuel Oil Storage Tank Prior to Completion of Chemical Analysis.Caused by Personnel Error.Individual counselled.W/930625 Ltr ML20044E7351993-05-20020 May 1993 LER 93-010-00:on 930425,noticed That Wide Range Reactor Level Indications Associated w/2A Condensing Chamber Drifted High than Level Instruments Associated w/2B Condensing Chamber.Caused by Level Instrument leaking.W/930520 Ltr ML20044E7341993-05-20020 May 1993 LER 93-008-00:on 930409,Tech Spec Violation Occurred.Caused by Setpoint Drift of Pressure Switch in Conjunction W/Less than Adequate Communication.Event Discussed W/Involved individuals.W/930520 Ltr ML20044E7381993-05-20020 May 1993 LER 93-009-00:on 930422,discovered Containment Sump Pump Collection & Flow Data Not Recorded in Surveillance Test. Caused by Personnel Error.Event Discussed W/Involved individuals.W/930520 Ltr ML20024G7391991-04-24024 April 1991 LER 91-004-00:on 910324,reactor Operator Failed to Initial Surveillance Test 5.3, Inoperable Valve Position Daily Log. Caused by Personnel Error Due to Failure to Follow Procedure.Operator counselled.W/910424 Ltr ML20028H3461990-12-10010 December 1990 Corrected LER 90-033-00:on 901108,discovered TS Limiting Condition of Operation Not Entered for Inoperable Containment Isolation Valve Due to Procedural Deficiency ML20044A6741990-06-25025 June 1990 LER 89-028-01:on 891108,determined That Standby Gas Treatment Sys Heater Control Relays Installed W/O Environ Qualification.Caused by Lack of Procedural Guidance.Relays relocated.W/900625 Ltr ML20043G8761990-06-14014 June 1990 LER 90-012-00:on 890815,discovered Valves Left Closed After Removal of Blocking Permit & on 890813,emergency Cooling Water Pump & Emergency Diesel Generator Removed from Svc. Caused by Inadequate procedures.W/900614 Ltr ML20043G0821990-06-11011 June 1990 LER 90-006-00:on 900511,blown Fuse from Battery Charger 3B Resulted in Declaring HPCI Sys,Core Spray B Logic,Rhr B Logic,Core Spray Subsystem B,Rhr Subsystem B & E2 & E4 Emergency Diesel Generators inoperable.W/900611 Ltr ML20043D7291990-06-0505 June 1990 LER 90-005-00:on 900507,Group 2A Primary Containment Isolation Sys Isolation Occurred During Surveillance Test. Caused by Inadequate Worker Practices.Blown Fuse Replaced & Personnel counseled.W/900605 Ltr ML20043D7211990-06-0404 June 1990 LER 90-011-00:on 900503,discovered That Tech Spec 3.4.1 Not Performed on 6-wk Frequency as Required.Caused by Personnel Error.Tracking of Surveillance Activities Scheduled to Be Transferred to Improved Software package.W/900604 Ltr ML20043C5921990-05-31031 May 1990 LER 90-010-00:on 900502,three Control Room Emergency Ventilation Actuations Occurred.Caused by Poor Electrical Continuity as Result of Oxidation Between plug-in Circuit Boards & Mating Electrical connections.W/900531 Ltr ML20043C5781990-05-30030 May 1990 LER 90-004-00:on 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test revised.W/900530 Ltr ML20043C5721990-05-30030 May 1990 LER 90-009-00:on 900430,discovered That Rod Block Monitor Not Been Proven Operable Prior to Exceeding 30% Power as Required by Tech Specs.Caused by Programmatic Deficiency. General Plant Procedures revised.W/900530 Ltr ML20043A4691990-05-16016 May 1990 LER 90-008-00:on 900417,discovered That Testing of LPCI Pumps & Core Spray Subsystems Not Performed When LPCI Pump D Declared Inoperable on 900414.Caused by Personnel Error. Procedures Declaring Pump Inoperable revised.W/900516 Ltr ML20043A7831990-05-14014 May 1990 LER 90-007-00:on 900412,evaluation Involving Seismic Qualification Performed Due to Postulated Failure of Condensate & Vacuum Pumps During Design Seismic Events. Caused by Design Oversight.Program updated.W/900514 Ltr ML20042F9581990-05-0707 May 1990 LER 90-001-01:on 900124,shift Surveillance Log Did Not Meet Requirements of Tech Specs.Caused by Deficient Procedure. Surveillance Log Revised to Include Daily Instrument Check. W/900507 Ltr ML20042F3211990-05-0202 May 1990 LER 90-006-00:on 900402,actuation of Emergency Diesel Generator Occurred.Caused by Personnel Miscommunication. Shift Mgt Will Be Reminded of Necessity to Control Activities in Control room.W/900502 Ltr ML20042E9811990-04-30030 April 1990 LER 90-002-01:on 900128,ESF Sys Actuations Occurred Due to Reactor Vessel Level Fluctuations After Manual Scram.Caused by Failure of O-ring on Fluid Inlet Port to Servo Valve for Hydraulically Operated Valve.Valve replaced.W/900430 Ltr ML20042E9101990-04-27027 April 1990 LER 90-005-00:on 900326,Tech Spec Surveillance Not Performed within Required Interval.Caused by Personnel Error.Personnel Counseled & Will Periodically Review Omitted Test Rept to Ensure Performance of Surveillance tests.W/900427 Ltr ML20042E6801990-04-23023 April 1990 LER 89-031-01:on 891206 & 900105,Agastat Relays Found Not Properly Secured by Seismic Support Straps.Caused by Inadequate Installation or Reinstallation of Seismic Straps. Straps Properly reconnected.W/900423 Ltr ML20042E6821990-04-19019 April 1990 LER 90-004-00:on 900321,discovered Potentially Inoperable Safety Sys Due to Inadequate Emergency Svc Water Cooling Flow Through Room Coolers.Caused by Gradual Buildup of Corrosion & Silt.Mod completed.W/900419 Ltr ML20012C4831990-03-12012 March 1990 LER 89-029-01:on 891117,primary Containment Isolation Sys Actuation Occurred During Performance of Refueling Floor Ventilation Exhaust Radiation Monitor Testing.Cause Unknown. Selector Switch & Relay Contacts cleaned.W/900312 Ltr ML20012C4821990-03-12012 March 1990 LER 89-024-01:on 891006,local Power Range Monitor Spike Caused Reactor Scram Signal While in Hot Shutdown.Caused by Design &/Or Mfg Process as Identified by Ge.Detector Placed in Bypass Position & Scram Signal reset.W/900312 Ltr ML20012A0171990-02-23023 February 1990 LER 90-002-00:on 900128,reactor Manually Scrammed Due to Leak of Electrohydraulic Control Sys Fluid.Caused by Lock Nut on Interlock Dump Valve Setting Adjustment Bolt Becoming Unsecured Due to Sys Vibration.Leak stopped.W/900223 Ltr ML20012A0021990-02-23023 February 1990 LER 90-001-00:on 900124,discovered That Daily Instrument Check of Main Stack Flow Rate Monitor Not Performed.Caused by Incomplete Procedure.Operating Shift Surveillance Log Revised to Include Daily Instrument check.W/900223 Ltr ML20012A0841990-02-0707 February 1990 LER 90-001-00:on 900108,HPCI Sys Declared Inoperable During Surveillance Testing When Start Time Exceeded 25 S.Caused by Inadequate Calibr Procedure Which Allowed Setting of 18 S. Ramp Generator & Signal Converter replaced.W/900207 Ltr ML20011F5791990-02-0707 February 1990 LER 89-007-01:on 890411,green Discoloration Discovered in Grease on Stabs of Several Control Fuses in 4 Kv Switchgear. Probably Caused by Incomplete Procedure.Maint Procedure M-054.004 Revised to Include Fuse insp.W/900207 Ltr ML20006A9621990-01-19019 January 1990 LER 89-033-00:on 891220,full Scram Signal Received When Technician Performed Surveillance on APRM D.Caused by Procedural Deficiencies & Inattention to Detail by Technician.Technician counseled.W/900119 Ltr ML19354E0121990-01-17017 January 1990 LER 89-032-00:on 891218,discovered That Weekly Surveillance Test Not Performed within Surveillance Interval.Caused by Inappropriate Action Based on Failure to Follow Procedure. Surveillance Test Coordinator counseled.W/900117 Ltr ML19354E0091990-01-16016 January 1990 LER 89-011-00:on 891213,discovered That Two Surveillance Tests of Turbine Stop & Control Valve Encl Not Performed Per Tech Specs.Caused by Incorrect Std Practice of Surveillance Testing.Programmatic Controls established.W/900116 Ltr ML19354E0101990-01-16016 January 1990 LER 89-015-01:on 890721,while Attempting to Remove Malfunctioning Reactor Pressure Vessel Regulator Set,Bypass & Control Valves Opened,Causing Steam Line Pressure to Increase to 480 Psig.Components replaced.W/900116 Ltr ML19354E0851990-01-11011 January 1990 LER 89-010-00:on 891211,monthly Surveillance Test ST 9.7 Not Performed within Surveillance Interval Established by Tech Spec Table 4.1.1.Caused by Combination of Programmatic Weaknesses.Review performed.W/900111 Ltr ML20005G2451990-01-11011 January 1990 LER 89-016-01:on 890720 & 22,LPRM Detector 4B-40-33 Spiked High,Resulting in Full Reactor Scram Signal While in Cold Shutdown.Caused by Design/Mfg Defect in GE Detector. Detector Placed in Bypass position.W/900111 Ltr ML20005G1901990-01-0808 January 1990 LER 89-009-00:on 891207,HPCI Sys Declared Inoperable When Sys Failed to Start During Pump,Valve & Flow Surveillance Test.Caused by Loose Lock Nut on HPCI Oil Sys Relief Valve. Lead Seal Wire to Be Placed on Valve caps.W/900108 Ltr ML20005F8711990-01-0505 January 1990 LER 89-031-00:on 891206,discovered That Four Agastat Relays Not Properly Secured by Seismic Support Straps.Root Cause Under Investigation & Will Be Reported in Rev to Ler.Support Straps Promptly reconnected.W/900105 Ltr ML20005E3911989-12-26026 December 1989 LER 89-030-00:0n 891126,steam Leak Discovered Coming from Packing on RCIC Injection Check Valve AO-22.Caused by Failure of Valve Stem Packing.Normal Reactor Level Restored & Mods of Valve Will Be pursued.W/891226 Ltr ML20011D2331989-12-18018 December 1989 LER 89-029-00:on 891117,Group III Primary Containment Isolation Sys Actuation Occurred During Refueling Floor Ventilation Exhaust Radiation Monitor Testing.Cause Unknown. Test Procedure to Be revised.W/891218 Ltr ML19332E7081989-12-0606 December 1989 LER 89-028-00:on 891108,review Determined That Standby Gas Treatment Sys Heater Control Relays Unqualified for post-LOCA Radiation Environ & Declared Inoperable.Cause Undetermined.Radiation Shielding installed.W/891206 Ltr ML19332E6331989-11-27027 November 1989 LER 89-007-00:on 891026,reactor Vessel Temp & Reactor Coolant Pressure Not Logged Every 15 Minutes as Required by Tech Spec 4.6.A.2 During Performance Integrated Leak Rate Testing.Caused by Procedure deficiency.W/891127 Ltr ML19332E7591989-11-27027 November 1989 LER 89-008-00:on 891027,primary Containment Isolation Sys Group II & III Isolations Occurred When Spurious Reactor Low Level Signal Sensed by Instruments.Possibly Caused by Air Bubble in Sensing Lines.Line backfilled.W/891127 Ltr ML19332D3401989-11-22022 November 1989 LER 89-006-00:on 891023,during Reactor Temp Adjustment, Reactor High Pressure Scram Occurred.Caused by Improper Planning & Coordination of Multiple Evolutions.Surveillance & Hydrostatic Test revised.W/891122 Ltr ML19332C4931989-11-20020 November 1989 LER 89-005-00:on 891020,reactor Protection Sys Actuation & Primary Containment Isolation Sys Actuation Occurred Due to False High Reactor Pressure Signal & lo-lo Reactor Vessel Level Signal,Respectively.Caused by spike.W/891120 Ltr ML19332C8191989-11-15015 November 1989 LER 89-027-00:on 891016,observation & Logging of Suppression Pool Temp as Required by Tech Spec 4.7.2 Not Met.Caused by Personnel Error.Operations Shift Team counseled.W/891115 Ltr ML19324C4361989-11-0808 November 1989 LER 89-026-00:on 891012,control Room Emergency Ventilation Sys Actuation Occurred Due to Momentary False High Radiation Signal from Control Room Radiation Monitor B.Caused by Sensitivity of Thumbwheel switch.W/891108 Ltr ML19324C1781989-11-0606 November 1989 LER 89-023-00:on 891005,outboard MSIV Ac Solenoid Pilot Valves de-energized,resulting in Expected Closure of Outboard MSIV D & Automatic Reactor Scram.Caused by Incomplete Guidance.Procedure revised.W/891106 Ltr ML19325F1801989-11-0606 November 1989 LER 89-024-00:on 891006,reactor Protection Sys Initiated Full Reactor Scram Signal.Caused by Output Signal for LPRM 40-33A Spiking High.Lprm Detector Placed in Bypass Position & Scram Signal reset.W/891106 Ltr ML19325F1811989-11-0202 November 1989 LER 89-022-00:on 891003,unterminated Lead in Circuit to HPCI Trip Solenoid Rendered HPCI Stop Valve Trip Functions Inoperable.Caused by Leads Loosely Hanging Inside Door Panel.Hanging Leads Secured & Panel inspected.W/891102 Ltr ML19325F1791989-11-0202 November 1989 LER 89-025-00:on 891007,determined That nonsafety-related Bellows Leak Detecting Pressure Switches Installed on Main Steam Relief Valves Could Prevent Opening During Design Basis Condition.Plant Alteration installed.W/891102 Ltr 1993-07-01
[Table view] Category:RO)
MONTHYEARML20045E9961993-07-0101 July 1993 LER 93-012-00:on 930607,discovered That Surveillance Test on Fire Protection Sys Missed on 930601.Caused by Insufficient Degree of Attention Applied by Nonlicensed Individual.Test Satisfactorily completed.W/930701 Ltr ML20045E8661993-06-25025 June 1993 LER 93-011-00:on 930528,TS Violation Occurred When New Fuel Was Added to EDG Fuel Oil Storage Tank Prior to Completion of Chemical Analysis.Caused by Personnel Error.Individual counselled.W/930625 Ltr ML20044E7351993-05-20020 May 1993 LER 93-010-00:on 930425,noticed That Wide Range Reactor Level Indications Associated w/2A Condensing Chamber Drifted High than Level Instruments Associated w/2B Condensing Chamber.Caused by Level Instrument leaking.W/930520 Ltr ML20044E7341993-05-20020 May 1993 LER 93-008-00:on 930409,Tech Spec Violation Occurred.Caused by Setpoint Drift of Pressure Switch in Conjunction W/Less than Adequate Communication.Event Discussed W/Involved individuals.W/930520 Ltr ML20044E7381993-05-20020 May 1993 LER 93-009-00:on 930422,discovered Containment Sump Pump Collection & Flow Data Not Recorded in Surveillance Test. Caused by Personnel Error.Event Discussed W/Involved individuals.W/930520 Ltr ML20024G7391991-04-24024 April 1991 LER 91-004-00:on 910324,reactor Operator Failed to Initial Surveillance Test 5.3, Inoperable Valve Position Daily Log. Caused by Personnel Error Due to Failure to Follow Procedure.Operator counselled.W/910424 Ltr ML20028H3461990-12-10010 December 1990 Corrected LER 90-033-00:on 901108,discovered TS Limiting Condition of Operation Not Entered for Inoperable Containment Isolation Valve Due to Procedural Deficiency ML20044A6741990-06-25025 June 1990 LER 89-028-01:on 891108,determined That Standby Gas Treatment Sys Heater Control Relays Installed W/O Environ Qualification.Caused by Lack of Procedural Guidance.Relays relocated.W/900625 Ltr ML20043G8761990-06-14014 June 1990 LER 90-012-00:on 890815,discovered Valves Left Closed After Removal of Blocking Permit & on 890813,emergency Cooling Water Pump & Emergency Diesel Generator Removed from Svc. Caused by Inadequate procedures.W/900614 Ltr ML20043G0821990-06-11011 June 1990 LER 90-006-00:on 900511,blown Fuse from Battery Charger 3B Resulted in Declaring HPCI Sys,Core Spray B Logic,Rhr B Logic,Core Spray Subsystem B,Rhr Subsystem B & E2 & E4 Emergency Diesel Generators inoperable.W/900611 Ltr ML20043D7291990-06-0505 June 1990 LER 90-005-00:on 900507,Group 2A Primary Containment Isolation Sys Isolation Occurred During Surveillance Test. Caused by Inadequate Worker Practices.Blown Fuse Replaced & Personnel counseled.W/900605 Ltr ML20043D7211990-06-0404 June 1990 LER 90-011-00:on 900503,discovered That Tech Spec 3.4.1 Not Performed on 6-wk Frequency as Required.Caused by Personnel Error.Tracking of Surveillance Activities Scheduled to Be Transferred to Improved Software package.W/900604 Ltr ML20043C5921990-05-31031 May 1990 LER 90-010-00:on 900502,three Control Room Emergency Ventilation Actuations Occurred.Caused by Poor Electrical Continuity as Result of Oxidation Between plug-in Circuit Boards & Mating Electrical connections.W/900531 Ltr ML20043C5781990-05-30030 May 1990 LER 90-004-00:on 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test revised.W/900530 Ltr ML20043C5721990-05-30030 May 1990 LER 90-009-00:on 900430,discovered That Rod Block Monitor Not Been Proven Operable Prior to Exceeding 30% Power as Required by Tech Specs.Caused by Programmatic Deficiency. General Plant Procedures revised.W/900530 Ltr ML20043A4691990-05-16016 May 1990 LER 90-008-00:on 900417,discovered That Testing of LPCI Pumps & Core Spray Subsystems Not Performed When LPCI Pump D Declared Inoperable on 900414.Caused by Personnel Error. Procedures Declaring Pump Inoperable revised.W/900516 Ltr ML20043A7831990-05-14014 May 1990 LER 90-007-00:on 900412,evaluation Involving Seismic Qualification Performed Due to Postulated Failure of Condensate & Vacuum Pumps During Design Seismic Events. Caused by Design Oversight.Program updated.W/900514 Ltr ML20042F9581990-05-0707 May 1990 LER 90-001-01:on 900124,shift Surveillance Log Did Not Meet Requirements of Tech Specs.Caused by Deficient Procedure. Surveillance Log Revised to Include Daily Instrument Check. W/900507 Ltr ML20042F3211990-05-0202 May 1990 LER 90-006-00:on 900402,actuation of Emergency Diesel Generator Occurred.Caused by Personnel Miscommunication. Shift Mgt Will Be Reminded of Necessity to Control Activities in Control room.W/900502 Ltr ML20042E9811990-04-30030 April 1990 LER 90-002-01:on 900128,ESF Sys Actuations Occurred Due to Reactor Vessel Level Fluctuations After Manual Scram.Caused by Failure of O-ring on Fluid Inlet Port to Servo Valve for Hydraulically Operated Valve.Valve replaced.W/900430 Ltr ML20042E9101990-04-27027 April 1990 LER 90-005-00:on 900326,Tech Spec Surveillance Not Performed within Required Interval.Caused by Personnel Error.Personnel Counseled & Will Periodically Review Omitted Test Rept to Ensure Performance of Surveillance tests.W/900427 Ltr ML20042E6801990-04-23023 April 1990 LER 89-031-01:on 891206 & 900105,Agastat Relays Found Not Properly Secured by Seismic Support Straps.Caused by Inadequate Installation or Reinstallation of Seismic Straps. Straps Properly reconnected.W/900423 Ltr ML20042E6821990-04-19019 April 1990 LER 90-004-00:on 900321,discovered Potentially Inoperable Safety Sys Due to Inadequate Emergency Svc Water Cooling Flow Through Room Coolers.Caused by Gradual Buildup of Corrosion & Silt.Mod completed.W/900419 Ltr ML20012C4831990-03-12012 March 1990 LER 89-029-01:on 891117,primary Containment Isolation Sys Actuation Occurred During Performance of Refueling Floor Ventilation Exhaust Radiation Monitor Testing.Cause Unknown. Selector Switch & Relay Contacts cleaned.W/900312 Ltr ML20012C4821990-03-12012 March 1990 LER 89-024-01:on 891006,local Power Range Monitor Spike Caused Reactor Scram Signal While in Hot Shutdown.Caused by Design &/Or Mfg Process as Identified by Ge.Detector Placed in Bypass Position & Scram Signal reset.W/900312 Ltr ML20012A0171990-02-23023 February 1990 LER 90-002-00:on 900128,reactor Manually Scrammed Due to Leak of Electrohydraulic Control Sys Fluid.Caused by Lock Nut on Interlock Dump Valve Setting Adjustment Bolt Becoming Unsecured Due to Sys Vibration.Leak stopped.W/900223 Ltr ML20012A0021990-02-23023 February 1990 LER 90-001-00:on 900124,discovered That Daily Instrument Check of Main Stack Flow Rate Monitor Not Performed.Caused by Incomplete Procedure.Operating Shift Surveillance Log Revised to Include Daily Instrument check.W/900223 Ltr ML20012A0841990-02-0707 February 1990 LER 90-001-00:on 900108,HPCI Sys Declared Inoperable During Surveillance Testing When Start Time Exceeded 25 S.Caused by Inadequate Calibr Procedure Which Allowed Setting of 18 S. Ramp Generator & Signal Converter replaced.W/900207 Ltr ML20011F5791990-02-0707 February 1990 LER 89-007-01:on 890411,green Discoloration Discovered in Grease on Stabs of Several Control Fuses in 4 Kv Switchgear. Probably Caused by Incomplete Procedure.Maint Procedure M-054.004 Revised to Include Fuse insp.W/900207 Ltr ML20006A9621990-01-19019 January 1990 LER 89-033-00:on 891220,full Scram Signal Received When Technician Performed Surveillance on APRM D.Caused by Procedural Deficiencies & Inattention to Detail by Technician.Technician counseled.W/900119 Ltr ML19354E0121990-01-17017 January 1990 LER 89-032-00:on 891218,discovered That Weekly Surveillance Test Not Performed within Surveillance Interval.Caused by Inappropriate Action Based on Failure to Follow Procedure. Surveillance Test Coordinator counseled.W/900117 Ltr ML19354E0091990-01-16016 January 1990 LER 89-011-00:on 891213,discovered That Two Surveillance Tests of Turbine Stop & Control Valve Encl Not Performed Per Tech Specs.Caused by Incorrect Std Practice of Surveillance Testing.Programmatic Controls established.W/900116 Ltr ML19354E0101990-01-16016 January 1990 LER 89-015-01:on 890721,while Attempting to Remove Malfunctioning Reactor Pressure Vessel Regulator Set,Bypass & Control Valves Opened,Causing Steam Line Pressure to Increase to 480 Psig.Components replaced.W/900116 Ltr ML19354E0851990-01-11011 January 1990 LER 89-010-00:on 891211,monthly Surveillance Test ST 9.7 Not Performed within Surveillance Interval Established by Tech Spec Table 4.1.1.Caused by Combination of Programmatic Weaknesses.Review performed.W/900111 Ltr ML20005G2451990-01-11011 January 1990 LER 89-016-01:on 890720 & 22,LPRM Detector 4B-40-33 Spiked High,Resulting in Full Reactor Scram Signal While in Cold Shutdown.Caused by Design/Mfg Defect in GE Detector. Detector Placed in Bypass position.W/900111 Ltr ML20005G1901990-01-0808 January 1990 LER 89-009-00:on 891207,HPCI Sys Declared Inoperable When Sys Failed to Start During Pump,Valve & Flow Surveillance Test.Caused by Loose Lock Nut on HPCI Oil Sys Relief Valve. Lead Seal Wire to Be Placed on Valve caps.W/900108 Ltr ML20005F8711990-01-0505 January 1990 LER 89-031-00:on 891206,discovered That Four Agastat Relays Not Properly Secured by Seismic Support Straps.Root Cause Under Investigation & Will Be Reported in Rev to Ler.Support Straps Promptly reconnected.W/900105 Ltr ML20005E3911989-12-26026 December 1989 LER 89-030-00:0n 891126,steam Leak Discovered Coming from Packing on RCIC Injection Check Valve AO-22.Caused by Failure of Valve Stem Packing.Normal Reactor Level Restored & Mods of Valve Will Be pursued.W/891226 Ltr ML20011D2331989-12-18018 December 1989 LER 89-029-00:on 891117,Group III Primary Containment Isolation Sys Actuation Occurred During Refueling Floor Ventilation Exhaust Radiation Monitor Testing.Cause Unknown. Test Procedure to Be revised.W/891218 Ltr ML19332E7081989-12-0606 December 1989 LER 89-028-00:on 891108,review Determined That Standby Gas Treatment Sys Heater Control Relays Unqualified for post-LOCA Radiation Environ & Declared Inoperable.Cause Undetermined.Radiation Shielding installed.W/891206 Ltr ML19332E6331989-11-27027 November 1989 LER 89-007-00:on 891026,reactor Vessel Temp & Reactor Coolant Pressure Not Logged Every 15 Minutes as Required by Tech Spec 4.6.A.2 During Performance Integrated Leak Rate Testing.Caused by Procedure deficiency.W/891127 Ltr ML19332E7591989-11-27027 November 1989 LER 89-008-00:on 891027,primary Containment Isolation Sys Group II & III Isolations Occurred When Spurious Reactor Low Level Signal Sensed by Instruments.Possibly Caused by Air Bubble in Sensing Lines.Line backfilled.W/891127 Ltr ML19332D3401989-11-22022 November 1989 LER 89-006-00:on 891023,during Reactor Temp Adjustment, Reactor High Pressure Scram Occurred.Caused by Improper Planning & Coordination of Multiple Evolutions.Surveillance & Hydrostatic Test revised.W/891122 Ltr ML19332C4931989-11-20020 November 1989 LER 89-005-00:on 891020,reactor Protection Sys Actuation & Primary Containment Isolation Sys Actuation Occurred Due to False High Reactor Pressure Signal & lo-lo Reactor Vessel Level Signal,Respectively.Caused by spike.W/891120 Ltr ML19332C8191989-11-15015 November 1989 LER 89-027-00:on 891016,observation & Logging of Suppression Pool Temp as Required by Tech Spec 4.7.2 Not Met.Caused by Personnel Error.Operations Shift Team counseled.W/891115 Ltr ML19324C4361989-11-0808 November 1989 LER 89-026-00:on 891012,control Room Emergency Ventilation Sys Actuation Occurred Due to Momentary False High Radiation Signal from Control Room Radiation Monitor B.Caused by Sensitivity of Thumbwheel switch.W/891108 Ltr ML19324C1781989-11-0606 November 1989 LER 89-023-00:on 891005,outboard MSIV Ac Solenoid Pilot Valves de-energized,resulting in Expected Closure of Outboard MSIV D & Automatic Reactor Scram.Caused by Incomplete Guidance.Procedure revised.W/891106 Ltr ML19325F1801989-11-0606 November 1989 LER 89-024-00:on 891006,reactor Protection Sys Initiated Full Reactor Scram Signal.Caused by Output Signal for LPRM 40-33A Spiking High.Lprm Detector Placed in Bypass Position & Scram Signal reset.W/891106 Ltr ML19325F1811989-11-0202 November 1989 LER 89-022-00:on 891003,unterminated Lead in Circuit to HPCI Trip Solenoid Rendered HPCI Stop Valve Trip Functions Inoperable.Caused by Leads Loosely Hanging Inside Door Panel.Hanging Leads Secured & Panel inspected.W/891102 Ltr ML19325F1791989-11-0202 November 1989 LER 89-025-00:on 891007,determined That nonsafety-related Bellows Leak Detecting Pressure Switches Installed on Main Steam Relief Valves Could Prevent Opening During Design Basis Condition.Plant Alteration installed.W/891102 Ltr 1993-07-01
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K9931999-10-14014 October 1999 Safety Evaluation Supporting Amend 234 to License DPR-56 ML20217B4331999-10-0505 October 1999 Safety Evaluation Supporting Amend 233 to License DPR-56 ML20217G3541999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbaps,Units 2 & 3. with ML20216H7091999-09-24024 September 1999 Safety Evaluation Supporting Amends 229 & 232 to Licenses DPR-44 & DPR-56,respectively ML20212D1281999-09-17017 September 1999 Safety Evaluation Supporting Proposed Alternatives CRR-03, 05,08,09,10 & 11 ML20212A5871999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Peach Bottom,Units 2 & 3.With ML20211D5501999-08-23023 August 1999 Safety Evaluation Supporting Amends 228 & 231 to Licenses DPR-44 & DPR-56,respectively ML20212H6311999-08-19019 August 1999 Rev 2 to PECO-COLR-P2C13, COLR for Pbaps,Unit 2,Reload 12 Cycle 13 ML20210N7641999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for PBAPS Units 2 & 3. with ML20209H1121999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbaps,Units 2 & 3. with ML20195H8841999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbaps,Units 2 & 3. with ML20206N1661999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pbaps,Units 2 & 3. with ML20206A2921999-04-20020 April 1999 Safety Evaluation Concluding That Proposed Changes to EALs for PBAPS Are Consistent with Guidance in NUMARC/NESP-007 & Identified Deviations Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20205K7411999-04-0707 April 1999 Safety Evaluation Supporting Amends 227 & 230 to Licenses DPR-44 & DPR-56,respectively ML20205P5851999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Peach Bottom Units 2 & 3.With ML20207G9971999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Peach Bottom Units 2 & 3.With ML20199E3471998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Peach Bottom,Units 1 & 2.With ML20205K0381998-12-31031 December 1998 PECO Energy 1998 Annual Rept. with ML20206P1651998-12-31031 December 1998 Fire Protection for Operating Nuclear Power Plants, Section Iii.F, Automatic Fire Detection ML20206D3651998-12-31031 December 1998 1998 PBAPS Annual 10CFR50.59 & Commitment Rev Rept. with ML20206D3591998-12-31031 December 1998 1998 PBAPS Annual 10CFR72.48 Rept. with ML20196G7021998-12-0202 December 1998 SER Authorizing Proposed Alternative to Delay Exam of Reactor Pressure Vessel Shell Circumferential Welds by Two Operating Cycles ML20196E8261998-11-30030 November 1998 Response to NRC RAI Re Reactor Pressure Vessel Structural Integrity at Peach Bottom Units 2 & 3 ML20198B8591998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Pbaps,Units 2 & 3. with ML20206R2571998-11-17017 November 1998 PBAPS Graded Exercise Scenario Manual (Sections 1.0 - 5.0) Emergency Preparedness 981117 Scenario P84 ML20198C6751998-11-0505 November 1998 Rev 3 to COLR for PBAPS Unit 3,Reload 11,Cycle 12 ML20195E5341998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Pbaps,Units 2 & 3. with ML20155C6071998-10-26026 October 1998 Safety Evaluation Supporting Amend 226 to License DPR-44 ML20155C1681998-10-22022 October 1998 Safety Evaluation Accepting Proposed Alternative Plan for Exam of Reactor Pressure Vessel Shell Longitudinal Welds ML20155H7721998-10-12012 October 1998 Rev 1 to COLR for Peach Bottom Atomic Power Station Unit 2, Reload 12,Cycle 13 ML20154J2401998-10-0505 October 1998 Safety Evaluation Supporting Amends 224 & 228 to Licenses DPR-44 & DPR-56,respectively ML20154H4771998-10-0505 October 1998 Safety Evaluation Supporting Amends 225 & 229 to Licenses DPR-44 & DPR-56,respectively ML20154G6821998-10-0101 October 1998 SER Related to Request for Relief 01A-VRR-1 Re Inservice Testing of Automatic Depressurization Sys Safety Relief Valves at Peach Bottom Atomic Power Station,Units 2 & 3 ML20154G6631998-10-0101 October 1998 Safety Evaluation Supporting Amends 223 & 227 to Licenses DPR-44 & DPR-56,respectively ML20154H5541998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Pbaps,Units 2 & 3. with ML20153B9651998-09-14014 September 1998 Safety Evaluation Supporting Amend 9 to License DPR-12 ML20151Y2901998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Pbaps,Units 2 & 3. with ML20238F2661998-08-24024 August 1998 Safety Evaluation Supporting Amend 222 to License DPR-44 ML20237B9531998-08-10010 August 1998 Specification for ISI Program Third Interval,Not Including Class Mc,Primary Containment for Bpaps Units 2 & 3 ML20237A7761998-08-10010 August 1998 SER Accepting Licensee Response to NRC Bulleting 95-002, Unexpected Clogging of RHR Pump Strainer While Operating in Suppression Pool Cooling Mode ML20237A5351998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Pbaps,Units 2 & 3 ML20236R8281998-07-15015 July 1998 Safety Evaluation Approving Proposed Alternative (one-time Temporary non-Code Repair) Pursuant to 10CFR50.55a(a)(3) (II) ML20236M3471998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Pbaps,Units 2 & 3 ML20249C4791998-06-0202 June 1998 Rev 6 to COLR for PBAPS Unit 2 Reload 11,Cycle 12 ML20248F4781998-06-0101 June 1998 Corrected Page 1 to SE Supporting Amends 221 & 226 to Licenses DPR-44 & DPR-56,respectively.Original Page 1 of SE Had Three Typos ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20248M3001998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Pbaps,Units 2 & 3 ML20247N5351998-05-11011 May 1998 SER Accepting Third 10-year Interval Inservice Program for Pump & Valves for Plant,Units 2 & 3 ML20249C4751998-05-0707 May 1998 Rev 5 to COLR for PBAPS Unit 2 Reload 11,Cycle 12 ML20247G0721998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Pbaps,Units 2 & 3 1999-09-30
[Table view] |
Text
-
.CCN-90 1401g
, =.
' PHILADELPHIA ELECTRIC COMPANY PIACilllOTIUM ATOMIC POWlR S1ATION R. D. I, llox 208 Delta, Itnnsylvania 17314 nua nomm-tm routa os a scantwt (717) 456-7014 D. M. Srnith Ykr Prnident January 16, 1990 Docket No. 50-277 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Licensee Event Report Peach Bottom Atomic Power Station - Unit 2 This LER involves a reactor scram resulting from an attempt to remove a malfunctioning Electro-Hydraulic Control System component from service. This revision provides additional information on the malfunctioning component.
Reference:
Docket No. 50-277
- Report Number: 2-89-015 Revision Number: 01 Event Date: 07/21/89 Report Date: 1/16/90 Facility: Peach Bottom Atomic Power Station RD 1. Box ~208, Delta, PA- 17314 The revision to this LER is being submitted pursuant to the requirements of10CFR50.73(a)(2)(iv).
i Sincerely, cc: J. J. Lyash, USNRC Senior Resident Inspector-W. T. Russell, USNRC, Region I 9001250197 900116 PDR ADOCK 05000277 k\
S PDC
i MC f 6rs 304 U S. NUCLE AR kt1ULATORY COMMIS$10N
+ MitOvt0 DMB ND 3164tr1M
' "';a' $ ' '8'
- LICENSEE EVENT REPORT (LER)
F ACILITY NAME til DOCalY NVuttR L21 ' AGE $3-Peach Bottom Atomic Power Station - Unit 2 0 lb l 0 l0 l 0 l2l 7l7 1 loFl Q Malfunctioning-Electro-llydraulic Control System Component Causes Scram When Removed Prom Service EVENT Daf t (Si LtR NUheRER gg# RIPOnf oaf t 471 CTMt R 9 ACILif tt s INv0LvtD lei DOC AS T NUMpt Rtle
%N L*4 Q,Wf7 *AciteTw havns MONTH DAY vtAR vtAR MONTM DAY vtAR 0 15101010 1 I I
~ ~
ol7 2l1 8 9 8l9 ol1 lL ol1 ol 2 1l6 9l 0 0 l 61 0 l0 1 0 i l l TMt3 REPORT 88 $UsurTTED PURSUANT TO TMt at OUIREMENTS Of to CF R h- (Caeca on, or me,e e,,ae ,eereneapp fin opt R Af tNG MODI m N ,,,,,,,, ,,,,,,,, y ,,,,n,,,,,, ,3.,,,,,
f 70 4061sH1Hd 90 38teH16 60 73teH2Hel T3.?tiel (10) 1 01719 so soni.HiH.i s02sisim _
50.73.H H.=>
_ gTMi, gsgga,*p,7,e, 20 406teH1 Heil 50.tStel(2 nd 50 73teH2Hveill Al J66A 8 20 608telt1 Hest 60.73 el(2 Heil 90 73 ell 2Hv4Mit!
20 totteH1 Hel 60.73 eH3H460 to.731eH2Hal LICIN$tt CONT ACT FOR THIS Lt R litt NAME TE L tPMOht NUMBER Asit & Copt T. E. Cribbe, Regulatory Engineer 71117 41516 l-17101114 COMPLET4 ONE LINE FOR 4 ACM COMPONENT F AILURE DESCRIDED IN 7 Mil R4 PORT (131 LE "A * " '
CAust 8vsTtM Cou'ONENT MAN g (AC R E 'OR,1,A CAust syst t M CoMooNtNT ut o >R X TIG l l PIE G 1018 l 4 Y I i l l l l l l l l 1 I I i 1 1 I I l l l
$UPPLEMtNT AL REPORT E K'tCTED 114l MONT H DAY vtAR 455 Ilf ,es remow,e $KPECTfD $UgstIS$10N OA TL) NO l l l
- . 1 R A C T m. . . e ,,,,.. - ., . . -, ,,,,-,,.. ,. , n .3 At 2231 on 7/21/89 with Unit 2 at 79% thermal power, an attempt was made to remove a malfunctioning Reactor Pressure Vessel (RPV) Pressure Regulator Set from the electronic portion of the Main Turbine (MT) Electro-Hydraulic Control (EHC) Pressure Regulating System. Immediately, the MT Bypass and Control Valves opened, causing main steam line pressure to decrease to approximately 480 psig. At 850 psig main steam line pressure a Group I Isolation occurred causing the Main Steam Isolation Valves (MSIV)toclose. As a result, a full reactor scram occurred. RPV level decrease due to shrink following MSIV closure resulted in a Group II and III isolation as level decreased below 0 inches. Two Main Steam Relief Valves (MSRV) lifted once. automatically, followed by manual Operator cycling of MSRVs to control RPV pressure between 930 psig and 1060 psig. The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems were placed in operation to control RPV pressure and level. The root cause of this event was a malfunction of
, the electronic portion of the "A" RPV Pressure Regulator Set. No actual safety consequences occurred as a result of this event. The majority of the "A" Regulator electronic components were replaced. This event has been reviewed with appropriate I plant personnel.. One previous similar LER was identified.
__ ._ _ _ ._ _ _ _ _ _ _ ~
anc v4 wucten4:vutony commissio=
" P-. aspa UCENSEE EVENT REPORT (LCJ TEAT CINTINUADON m novtooueim 3 % em -
- EXPatt:l'31C P ACILifv esaaet (16 o0CKti 88Uniten (#1 ten edumeth ten PA04 (31 Psach Bottom Atomic Power Station '8 '" l ' % 3" WU Unit 2 0 l5 J O l0 l0 l 2 l 7l7 8l 9 -
o ll l 5 -
ol l ol 2 or ol6 nxt u . o c w aww nn Requirements for the Report ~
This LER is being submitted pursuant to'10CFR50.73(a)(2)(iv) to report those conditions which resulted in the automatic actuation of an Engineered Safety Feature i.e., Reactor Protection System (EIIS JC) and Primary Containment Isolation System ,
(EIIS:JM). :
Unit Status at Time of Event !
Unit 2 was in the Run Mode at 79% thermal. power increasing generator (EIIS:TG) load l at a rate of 10 MWe per hour. Unit 3 was in the Refuel Mode with the core offloaded. i Unit 2 "A" RPV Pressure Regulator (EIIS:RG) Set output signal was drifting slowly upward at a constant rate (pressure decreasing).
Description of Event NOTE: TheMainTurbine(MT)(EIIS:TRB) Electro-HydraulicControl(EHC)(EIIS:TG) (
, Pressure Regulating System consists of two redundant Pressure Regulator (E!IS:RG) j Sets which maintain constant reactor (EIIS:RCT) pressure in coordination with MT i speed and load control.
On7/3/89whileperformingMTEHCPressureRegulatingSystemstabilitytesting-(with !
the "A" Pressure Regulator in service) it was noted that the signal output from "B" Pressure Regulator circuit was offset from the "A" approximately 15 psig (normal 3 psig). The offset was returned to the 3 psig setting by adjusting the pressure [
setpoint bias potentiometer. Continued observation revealed the pressure ;
differential between regulator circuits to be increasing requiring periodic bias :
potentiometer adjustments to maintain 3 psid between channels (EIIS:CHA).. Testing '
indicated the "A" Regulator signal output was increasing slowly (causing pressure to :
decrease 0.4 psig/ day), while at the same time "B" Regulator signal' output appeared stable. It was noted that the bias potentiometer was quickly-running out of adjustment requiring action before a loss of the ability to maintain the required 3 +
psid between regulators occurred. The 3 psid value is required to enable a ,
relatively smooth transfer to the redundant regulator in the event of failure of the -
regulator in service. On 7/20/89 the Plant Manager was informed of the problem and options available.
-On 7/21/89 a meeting attended by the Plant Manager, System Engineering, Shift Supervision and a General Electric (GE) representative with a knowledge of EHC was held to evaluate each available option and its associated risk.in order to determine an action plan. After a review of the technical information available and with the concurrence of the vendor representative the decision was made to transfer control to the stable "B" Regulator and disable the "A" Regulator by removing the "A" Steam Line
- Resonance Compensator (SLRC) card (A-42 card) from the electronic portion (see attachment) of the MT EHC Pressure Regulating System. This course of action would;
- 1) provide a stable and reliable pressure regulator, 2) eliminate the need for bias pot adjustments and 3) eliminate the potential loss of Reactor Pressure Vessel (RPV) pressure control if degradation of the "A" Regulator continued or accelerated and 4) '
utilize the least complex method of disabling the "A" Regulator, Similar card-removals had been performed at Limerick Gereerating Station and several plants of other utilities without experiencing system transients. .
,.,,,...m. _ .,..r,, _m,- #.-.. ,- __
I Z3CFeem W .. U.S. 08UCLE13 $_EIUL&To#7 COhs.48#4one l
- UCENSEE EVENT MEPORT H.ER) TEXT CONTINUATION unovio oue No. mo-oios teints:a m o
- . F AC4Li(Y esAtst $11 DOC 8LET NutAtth til LIR NUnseth fel PAGE (3)
P@achBottomAtomicPoweEStation " "Wik" man l
-Unit 2 l l 0 l5 ln l0 l0 l 2l7 l 7 8l9 -
ol1]5 -
oli ol 3 oF ol6 i mvn m
..- w.m 1
The intent was to operate on the "B" Regulator only and leave the "A" Regulator out l of service until the next shutdown. j At 2231 on 7/21/89 with Unit 2 at 79% therral power, an attempt was made to remove l the malfunctioning "A" Pressure Regulator Set from the electronic portion of the MT i EHC Pressure Regulating System. This-was to be accomplished by pulling the SLRC l circuit: card. .The individual removing the card had difficulty on the first attempt, 1 while repositioning his grip for a second attempt a slight side to side movement occurred. Before card removal, could be completed several relay actuations were heard and the Plant Manager in conjunction with the vendor representative halted 1 removal.
l Immediately, the Main Turbine Bypass and Control Valves opened, causing main steam '
line pressure to decrease to approximately 480 psig. At 850 psig main steam line pressure, a Group I Isolation occurred causing the Mein Steam Isolation Valves i (MSIVs) to close. As a result, a full reactor scram signal occurred due to MSIV 1 closure. Initially, RPV level increased due to swell as a result of RPV pressure :
decrease, then decreased due to shrink as RPV pressure increased following MSIV I
I closure. This resulted in a Primary Containment Isolation System (EIIS:JM) Group II :
and III isolation signal as level decreased below 0 inches (172 inches above the l core). Isolation of the Reactor Water Cleanup System (EIIS:CE), drywell equipment i and floor drain sumps (EIIS:WD), drywell and torus instrument nitrogen supply (EIIS:LK)._and reactor building ventilation (Ells:VA) occurred as a result. RPV :
level decrease stopped at -35 inches, then level increased as Reactor Feed Pumps {
(EIIS:P) (RFP) "A", "B" and "C" continued to provide makeup. The RFPs were tripped i before level exceeded +45 inches, level peaked at +75 inches and then decreased as :
automatic are manual Main Steam Relief Valve (MSRV) (EIIS:RV) operation reduced RPV 't inventory. Yorus (EIIS:BS) cooling was placed in operation in anticipation of the :
decay heat removal-requirements for the isolated reactor. RPV pressure increased to :
approximately 1100 psig during the transient and the following sequence occurred; I) -
an Alternate Rod Insertion (ARI) High' Pressure backup scram occurred at 2238 2) MRVs l "J" and "H" lifted once automatically at 2242 and 3) the Operator manually cycled MSRVs "A", "B", "K" and "C" once each to reduce and maintain reactor pressure at 930 -
psig to 1060 psig between the time 2242 and 2258. Both Reactor Recirculation pumps tripped as expected during a 13 KV Auxiliary Bus (EIIS:BU) Fast Transfer caused by a L MT trip initiated generator lockout'. The Reactor Recirculation (EIIS:AD) pumps were'.
Subsequently restarted to provide forced circulation. The scram and Group 11 and III '
isolations signals were reset at 2305. The High Pressure Coolant Injection (EIIS:BJ)
(HPCI) and Reactor Core Isolation Cooling (EIIS:BN) (RCIC) systems were placed in :
operation to control RPV pressure and level. HPCI was maintained in the full flow ;
test mode (Condensate Storage (EIIS:KA) Tank (CST) to CST) to remove decay heat while RC!C was utilized to control RPV level. The unit was stabilized in the Hot Shutdown t condition at 2313.
P t
.v.s. cro, i,es-no....,ooo ro :
pg. eoxu mn
=ac .:
P , p* * .en M
. ut auct4aut utatoav commerion
'J UCENSEE EVENT REPORT (LER) TEXT C3NTINUATION - movfo ove no. mo-oio.
. EXPWits: 8/31/M PACILITY esAMS (11 DOCKtttouuBERtal Lan NUMeta (4) Pact (3) '
P2ach' Bottom Atomic? Power Station' """ '-
"UE '
"Eu'aN Enit 2 Text sn ew, as m espeac un asumeant mac rean ama myn -
0 l5 l0 l0 l0 l 2(? l? Bl 9 -
olll5 -
o \1 o I4 oF 'ol6
' Cause of-the' Event i[Therootcause-ofthisevent'wasthe' unusual-' service-conditionresultingfromt malfunctioning component preceding the attcmpt to remove the malfunctioning.Preem.
l Regulator Set from-the electronic portion of the RPV Pressure Regulator System. .The=
malfunctioning component was found to be the "A" Main Steam Pressure Transducer (PT-
- h84)-(Ells:PT). which drives;the "A" Main Steam Pressure Sensor (Dwg. No.'832E840GR1)
Pressure;)(i.e.. more~open control valves). . Warm moist' air.from the moisture.(E separator area condensed in the electrical conduit resulting in; water damage to the transducer. The water camaged transducer affected the performance of the "A" pressure channel. sThis anomaly' precipitated the evaluation, decisions and'actionsi taken as addressed in the Description of Event.-
The catst of the EHC transient was determined-to be the intermittent removal of thel-
- 30. volt supply to the'"A" Pressure Regulator.and'the voltage transients that' resulted.'lnis' voltage fluctuation apparently was caused-by the slight side to. side movement that occurred during the. attempt to remove the SLRC circuit card.
. Analysis of the Event a
No. safety consequences occurred as a result of this event.
m The Reactor Protection _ System (RPS) and Primary Containment Isolation Systems (PCIS) operated properly throughout the. transient. 'The other Safety systems responded properly as described in the Description of Event. .Due to the close proximity to -j ful1 power atlthe start of.the event, the expected plant response at-100% power would not have differed significantly.
Corrective Actions '
1 The following corrective actions have been taken:
, 1. The Main Steam Pressure Sensor was returncd to the1 vendor (GE) for analysis.
A It was evaluated and found:not to have any defects. -
- 2. The "A" Pressure Regulator circuit cards with the exception of the SLRC and Operational Amplifier were replaced and calibrated.
- 3. The "A."
SLRC has.been bench tested satisfactorily withoat identifying ~any deficiencies. jl
- 4. RPV pressure has been monitored from.the Control Room since returning to power i and set p ersure has-not displayed any drifting characteristics.
1 S. This event has been reviewed by appropriate plant personnel.
11 6. The "A".and
- l -
prevent water"B" Main Steam intrusion. Pressure Transducers were sealed.at the transducer to .
1 y
- ~"
e... m . ,... m . e -
NRC fosm atla . U 8 itVCLtan s.51ULATORY COMMtBS1088 7'
- LICENSEE EVENT REPORT (LER) TEXT' CONTINUATION maovto oMe No mom EXPIRES. 8/31/2 P ACILITY NAME (1) DOCKni NUMSth GB LER NUMBER (gp , PAGE Di lPaach Bottom Atomic Power Station '" " Na '
EYaN Unit 2 - -
Ol6l0l0l0l2l7l7 8l9 -
0l1l5 -
011 0l 5 OF o l6 TEXT rX more spece As rueursef, see adutoonsIN4C Fonn JNAW 11h 5
l 7. The Preventive Maintenance.(PM) Program was enhanced to include: a) an inspection of each pressure trarr.ducer during each recalibration, and b) .i replacement of both pressure transducers every third refueling outage.
The-following corrective-action is planned:
l 1. The conduit from the moisture separator area will also be sealed.
Previous Similar Events There was one previous similar LER associated with troubleshooting the EHC System
.that resulted in a reactor scram. The following is a brief description of the event and those actions taken to prevent recurrence: .-l LER 3-87-02 addresses a low reactor water level scram caused by a main steam transient while performing testing on the EHC System. This testing was being-performed to ensure the corrective actions taken in LER-3-87-01 to eliminate EHC signal oscillations were effective. Corrective actions for this LER included having .
.a vendor representative and other personnel knowledgeable in the.EHC System present while testing. .
f This action to prevent recurrence would not have prevented this event. Personnel experienced in the operation, maintenance and testing of-the EHC. System, including the Plant Manager and the knowledgeable vendor representative were present or .
involved throughout the testing / troubleshooting and decision making process.
I f
4-
'i i
hs NRC FORF 304A 'U.S. CPO, 1988-520 589r00010 -
C43) -
Ig . y qc p g
l o ,,
2, . ,, o . a W t2 .. - ^
1i :
4, l "h~
- I ',e , ,
n
. g, ,- ,
= . . :1 g *
h 'O B ~
RPV Pressure Control 5
.g l a r ..
o-C.
o
-} .m w i.seS w e, r.n # w stac-4 a' lll T PT :g. m
-. t ~
. l
. m
'[~ 's* M *"-
Q.
kl O, y f
e 1 wst e 1,
g x/
/
_I._
. =
3r.
o.
sH 3-
. m
-ImmL 1 g
o .g E
. ~
v=
s.r
-I h-y I _ m _a .p,,,,,,, o- [
. to.d sai. .
o -
f I- N
.O:
O" z
T MSPS T Pressure Regulator -B*SLRC -
T PT u
d z-O
"'*""- ~
g'I S .[
.I k ,
L 1 3 \c , t e
=
d~
o
.I;
~
vu g / ': g
- .z.
. =
e tg gL.= o
_ g{ g
+ g; a
? Legend: ~ Attachment 1 ' _
jj .. ~ l
[ PT --Pressure Transducer * =!
h,'{j
.3 MSPS - Main Steam Pressure Sensor o 8k Ef.j
~
,3 SLRC - Ste&m Line Resonance Compensator m ,
o E .8a
- o g 5 C' ;I s
e.., -~ ~ _ - -
f 3 my. _
,r,.#.+ 1 r '-w--~