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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20046C3131993-07-28028 July 1993 LER 93-015-00:on 930630,HPCI Sys Declared Inoperable Due to Indicated Flow During Surveillance.Caused by Foreign Matl That Plugged Portion of Restricting Orifice.Restricting Orifice Removed,Cleaned & reinstalled.W/930728 Ltr ML20046B5241993-07-26026 July 1993 LER 93-007-01:on 930317,automatic Closing of RCIC Sys Turbine Steam Supply Isolation Valves Occurred Due to High Steam Flow Signal.Increased Frequency of Valve Testing from Monthly to weekly.W/930726 Ltr ML20045F8601993-06-30030 June 1993 LER 93-014-00:on 930531,automatic Scram Occurred,Resulting from Operation of Auxiliary Transformer Differential Relay During Power Ascension.Uat Phase B & C Differential Relays Left Interchanged as Investigative aid.W/930630 Ltr ML20045E2301993-06-25025 June 1993 LER 93-013-00:on 930530,RCIC Sys Declared Inoperable & 7 Day TS 3.5.D.2 LCO Entered Due to Speed Oscillations Which Occurred After Several Minutes of steady-state Operation. Caused by Actuator Failure.Actuator replaced.W/930625 Ltr ML20045E2251993-06-25025 June 1993 LER 93-012-00:on 930529,unplanned PCIS Group I Isolation Signal Occurred While Opening MSIV During Startup,Resulting in Automatic Closing of Related Valves.Caused by Licensed Operator Error.Group 1 Isolation reset.W/930625 Ltr ML20045D6941993-06-23023 June 1993 LER 93-011-00:on 930525,determined That Initial as-found Popping Pressures of Pilot Valves for Three Target Rock Main Steam Relief Valves Out of TS Tolerance.Caused by Setpoint Drift.Valves Reworked & Reassembled by Target Rock Corp ML20045D1861993-06-18018 June 1993 LER 93-010-00:on 930519,SUT Became de-energized During Planned Calibr of Turbine/Generator Relays Lockout Test While Shut Down.Caused by Personnel Error.Discussion Conducted W/Personnel Re Attention to detail.W/930618 Ltr ML20044H0441993-06-0202 June 1993 LER 93-009-00:on 930504,circuit Breaker Did Not Close During Planned Bus Transfer Due to Loose Control Circuit Wire.Lug Replaced & Reterminated on 930505 & Post Work Testing Completed W/Satisfactory results.W/930602 Ltr ML20024G7451991-04-24024 April 1991 LER 91-005-00:on 910325,diesel Generator Inoperable,Causing Loss of Ac Power to Train B Components of Safety Sys & Actuating Portions of Primary/Secondary Containment Sys. Caused by Voltage Regulator failure.W/910424 Ltr ML20029A6721991-02-25025 February 1991 LER 91-001-00:on 910125,automatic Primary Containment Isolation Control Sys Group 5 Actuation Occurred,Resulting in Closing of RCIC Turbine Steam Supply Isolation Valves. Caused by High Flow conditions.W/910225 Ltr ML20028G9731990-09-20020 September 1990 LER 89-036-01:on 891122,determined That HPCI Sys Inoperable Due to Inoperable Gland Seal Condenser Blower Motor.Caused by age-related Wear of Motor.Blower Motor Replaced & Blower Tested W/Satisfactory results.W/900920 Ltr ML20028G9121990-09-18018 September 1990 LER 89-037-01:on 891130,primary Containment/Traversing in-core Probe Ball Valve Discovered to Be Almost Full Open. Caused by Damage to Valve Stem Due to Manual Manipulation of Stem.Ball Valve & Solenoid Actuator replaced.W/900918 Ltr ML20043G3541990-06-12012 June 1990 LER 90-008-00:on 900513,automatic Scram Resulting from Load Rejection Occurred While at Full Power,Resulting in Trip of Generator Field Breakers.Caused by Fault on Offsite 345 Kv Transmission sys.Loss-of-field Relay replaced.W/900612 Ltr ML20043F8291990-06-0606 June 1990 LER 90-007-00:on 900507,discovered That Drywell to Suppression Chamber Vacuum Breaker Surveillance Not Performed Prior to Startup in 1988.Caused by Misunderstanding of requirements.W/900606 Ltr ML20042F5911990-04-30030 April 1990 LER 90-006-00:on 900330,determined That Position of Primary Containment Sys Isolation Valve Not Recorded Daily,Per Tech Specs.Review Performed to Determine If Other Problems Existed.Plant Shut Down & Review performed.W/900430 Ltr ML20012F5751990-04-0606 April 1990 LER 90-003-00:on 900311,automatic Actuation of Main Steam Sys Group 1 Portion of Primary Containment Isolation Control Sys Occurred.Caused by False High Reactor Vessel Water Level Signal.Procedure Developed Re backfill.W/900406 Ltr ML20012E9831990-03-30030 March 1990 LER 90-002-00:on 900228,determined That Max Fraction of Limiting Power Density Not Checked Daily During Reactor Power Operation,Per Tech Spec 4.1.B.On 900323,addl Tech Spec Issue Discovered.Tech Specs changed.W/900330 Ltr ML20012C4871990-03-12012 March 1990 LER 90-001-00:on 900209,24 H Limiting Condition for Operation Entered When Two RCS Instrumentation Excess Flow Check Valves Inappropriately Verified Operable During Testing.Cause Undetermined.Two Valves replaced.W/900312 Ltr ML20005G0981990-01-0808 January 1990 LER 89-038-00:on 891208,unplanned Automatic Reactor Protection Sys Scram Signal & Reactor Scram Occurred.Caused by False Low Reactor Vessel Water Level Signal.Local Level Indicators Satisfactorily calibr.W/900108 Ltr ML20005F8481990-01-0808 January 1990 LER 89-039-00:on 891209,automatic Actuation of RHR Sys Portion of Primary Containment Isolation Control Sys Occurred.Caused by Hydrodynamic Transient That Actuated Protective High Pressure switches.W/900108 Ltr ML20005E8551989-12-30030 December 1989 LER 89-037-00:on 891130,discovered That Traversing in-core Probe Ball Valve,Mfg by Consolidated Controls,Inc,Closed,In Violation of Tech Specs.Caused by Damage to Valve Stem.Ball Valve & Actuator replaced.W/891230 Ltr ML20011D1361989-12-11011 December 1989 LER 89-035-00:on 891111,inadvertent Actuation of Portion of Secondary Containment Sys During Surveillance Testing Occurred.Caused by Location of Radiation Isolation Control Sys Channel a Logic Relay.Procedure revised.W/891211 Ltr ML20005D7421989-12-0606 December 1989 LER 89-033-00:on 891106,automatic Actuation of Group 1 Portion of Primary Containment Isolation Control Sys (PCIS) Occurred.Caused by High Reactor Vessel Water Level.Manual Valve Closed & PCIS Logic Circuitry reset.W/891206 Ltr ML19351A4581989-12-0606 December 1989 LER 89-034-00:on 891108,determined That Reactor Pressure Exceeded 150 Psig on 891107 W/O Performing Surveillance Procedure 8.5.4.4, HPCI Valve Operability Test. Caused by Personnel Error.Procedure initiated.W/891206 Ltr ML19332D1421989-11-22022 November 1989 LER 89-031-00:on 891024,closing Times for Two in-series Primary Containment Isolation Valves Exceeded Acceptance Criteria During Shutdown Testing.Cause Undetermined.Closing Time Retested W/Satisfactory results.W/891122 Ltr ML19332D6231989-11-20020 November 1989 LER 89-032-00:on 891020,pneumatic Pressure Drop for Two of Four Automatic Depressurization Sys Accumulators Discovered Greater than Max Acceptable Value.Caused by Leakage from Seat of Supply Check Valves.Seat replaced.W/891120 Ltr ML19324C3181989-11-0909 November 1989 LER 88-002-01:on 880117,full Reactor Protection Sys Scram Trip Signal Occurred During Surveillance Test,Resulting in Incomplete Actuations.Caused by Procedure Inadequacy.Logic Relay Replaced & Procedure revised.W/891109 Ltr ML19324B1131989-10-21021 October 1989 LER 89-030-00:on 890926,control Room High Efficiency Air Filtration Sys Flowrate non-conservative Due to Procedure Error.Caused by Transcription Error.Procedure Revised & Test Reperformed Using Corrected procedure.W/891021 Ltr ML19325D4681989-10-16016 October 1989 LER 89-029-00:on 890914,locked High Radiation Area Door Found to Have Been Unsecured Contrary to Tech Spec 6.13.2. Caused by Failure to Adhere to Approved Procedures. Training on Access Requirements initiated.W/891016 Ltr ML19325D1861989-10-10010 October 1989 LER 89-028-00:on 890907,HPCI Sys Declared Inoperable Because Mechanical Overspeed Trip Occurred During Surveillance Test. Caused by Failure of Ramp Generator Signal Converter Module. Module Removed & Sent to Mfg for exam.W/891010 Ltr ML17277B8221984-07-20020 July 1984 LER 82-049/01X-1:on 821102,main Steam Line Safety Valve a Set W/Nitrogen at Steam Relief Setting.Caused by Personnel Error.Valve Removed & Replaced W/Properly Set Spare Valve. Procedure revised.W/840720 Ltr ML20024C3231983-06-24024 June 1983 Updated LER 82-038/03X-1:on 820817,RCIC Steam Line High Differential Pressure Alarm Received.Caused by Air in Sensing Lines to Pressure Switches.Cross Threaded Cap Found on in-line Snubber.Snubber replaced.W/830624 Ltr ML20024C2211983-06-24024 June 1983 Updated LER 80-053/03X-1:on 800814,0907 & 14,HPCI Turbine Exhaust Line Snubber 23-3-36 Determined Inoperable.Caused by Transient Hydrodynamic Shock During Startup & Shutdown. Snubbers upgraded.W/830624 Ltr ML20024B8451983-06-24024 June 1983 LER 83-031/03L-0:on 830528,HPCI Sys Made Inoperable to Repair Steam Leak on AO 2301-31.Caused by Leaking Stem Packing.Packing Replaced.Valve Tested Satisfactorily.Sys Returned to svc.W/830624 Ltr ML20024B8321983-06-24024 June 1983 LER 83-030/03L-0:on 830527,both HPCI Area Smoke Detection Sys Alarmed in Main Control Room.Caused by Valve A02301-31 Steam Leak Causing False Indications.Valve repaired.W/830624 Ltr ML20024C3101983-06-21021 June 1983 Updated LER 79-007/03X-1:on 790212,while Performing Test 8.7.4.3,valves AO-5033A,CV-5065-10,15,16 & 17 Failed to Give Proper Position Indication When Cycled & to Close within Specified Tolerance.Cause Not stated.W/830621 Ltr ML20024A8861983-06-16016 June 1983 LER 83-027/03L-0:on 830518,during Routine Insp,Folding Fire Door Found Inoperable.Caused by Operating Cable Becoming Detached from Pulley Due to Slack.Cable Reattached & Operating Mechanism adjusted.W/830616 Ltr ML20023D1691983-05-0909 May 1983 Updated LER 82-008/01X-1:on 820331,HPCI Injection Valve MO-2301-8 Failed to Open in Required Manner.Caused by Missing Wire Jumper Intended to Bypass Motor Operator Torque Switch.Jumper Installed & Valve Tested Satisfactorily ML20023D0671983-05-0909 May 1983 LER 83-021/03L-0:on 830411,during Surveillance Test 8.7.4.3, Primary Containment Isolation Valve AO-5033B Failed to Meet 10-s Closing Time Requirements.Caused by Lubricant Infiltrating Solenoid Valve Components.Components Replaced ML20023D0181983-05-0404 May 1983 LER 83-022/03L-0:on 830413,overload on Valve Motor 1301-17 Resulted in Loss of High Pressure Core Cooling Backup Capability.Caused by Oxidation in Motor End Bell & Motor Brushes,Due to Humidity Caused by Leaking Valve Packing ML20028G1591983-01-28028 January 1983 LER 83-002/01T-0:on 830116,flow Ref off-normal Alarm Received.Average Power Range Monitor Flux Scram Trip Settings Affected.Cause Under Investigation.Flow Inputs Monitored & Amplifier B Recalibr ML20028F4121983-01-17017 January 1983 LER 83-002/01X-0:on 830116,while Clearing off-normal Alarm for Flow Comparators,Average Power Range Monitor Flux Scram Trip Settings Found in Nonconservative Direction.Caused by Drifting Proportional Amplifiers.Settings Checked Daily ML20028B0421982-11-16016 November 1982 LER 82-049/01T-0:on 821102,main Steam Line Safety Valve a Set W/Nitrogen at Steam Relief Setting of 1240 Psig, Exceeding Tech Specs.Caused by Personnel Error.Set Valve Removed,Procedure Adherence Stressed & Procedure Revised ML20027C9801982-10-20020 October 1982 LER 82-039/03L-0:on 820920,timing Tests of Three Reactor Water Cleanup Isolation Valves 1201-2,-5 & -80 Not Completed within Allowed Time.Caused by Operational Constraints Requiring Testing at Reduced Power ML20027C9941982-10-20020 October 1982 LER 82-040/03L-0:on 820920,found Reactor Protection Sys Initiation Function of Turbine Stop Valve Completed 1 Day Late.Caused by Decision to Defer Test Due to MSIV D Line Isolation.Tracking Sys Implemented to Prevent Recurrences ML20052B9141982-04-23023 April 1982 LER 82-010/03L-0:on 820326,during Review of Testing Requirements,Surveillance Test 8.I.4 Standby Liquid Control Sys Found Not Verified.Caused by Breakdown in Multidiscipline Review of Results ML20052F2071982-04-14014 April 1982 Updated LER 81-060/03X-1:on 811020,determined Potential Exists to Reduce Number of Operable APRM & IRM Instrument Channels in Rod Block Trip Sys to Below Tech Spec Min ML20050A6321982-03-19019 March 1982 Updated LER 82-055/01T-1:on 810926,during Shutdown,Yarway Level Indicators Started Oscillating.Drywell Temp Was 240 F at 81-ft W/Elevation Coolant Temp 220 F.Caused by Ineffective Drywell Cooling.Ventilation Returned to Svc ML20050A4521982-03-19019 March 1982 LER 82-006/01T-0:on 820304,while Performing Maint Re SIL 352 on Pressure Adjustment Mechanism HPCI Stop Valve,Three Broken cap-screws Found.No Cause Determined.Screws Examined, Replaced & Sys Returned to Svc ML20050A4331982-03-19019 March 1982 Updated LER 81-062/01X-1:on 811116,Wyle Labs Informed Util That Two Target Rock Safety Relief Valves Had Not Passed Setpoint Tests.Caused by Setpoint Shift Due to Changes to Designed Differential Forces from Excessive Pilot Leakage 1993-07-28
[Table view] Category:RO)
MONTHYEARML20046C3131993-07-28028 July 1993 LER 93-015-00:on 930630,HPCI Sys Declared Inoperable Due to Indicated Flow During Surveillance.Caused by Foreign Matl That Plugged Portion of Restricting Orifice.Restricting Orifice Removed,Cleaned & reinstalled.W/930728 Ltr ML20046B5241993-07-26026 July 1993 LER 93-007-01:on 930317,automatic Closing of RCIC Sys Turbine Steam Supply Isolation Valves Occurred Due to High Steam Flow Signal.Increased Frequency of Valve Testing from Monthly to weekly.W/930726 Ltr ML20045F8601993-06-30030 June 1993 LER 93-014-00:on 930531,automatic Scram Occurred,Resulting from Operation of Auxiliary Transformer Differential Relay During Power Ascension.Uat Phase B & C Differential Relays Left Interchanged as Investigative aid.W/930630 Ltr ML20045E2301993-06-25025 June 1993 LER 93-013-00:on 930530,RCIC Sys Declared Inoperable & 7 Day TS 3.5.D.2 LCO Entered Due to Speed Oscillations Which Occurred After Several Minutes of steady-state Operation. Caused by Actuator Failure.Actuator replaced.W/930625 Ltr ML20045E2251993-06-25025 June 1993 LER 93-012-00:on 930529,unplanned PCIS Group I Isolation Signal Occurred While Opening MSIV During Startup,Resulting in Automatic Closing of Related Valves.Caused by Licensed Operator Error.Group 1 Isolation reset.W/930625 Ltr ML20045D6941993-06-23023 June 1993 LER 93-011-00:on 930525,determined That Initial as-found Popping Pressures of Pilot Valves for Three Target Rock Main Steam Relief Valves Out of TS Tolerance.Caused by Setpoint Drift.Valves Reworked & Reassembled by Target Rock Corp ML20045D1861993-06-18018 June 1993 LER 93-010-00:on 930519,SUT Became de-energized During Planned Calibr of Turbine/Generator Relays Lockout Test While Shut Down.Caused by Personnel Error.Discussion Conducted W/Personnel Re Attention to detail.W/930618 Ltr ML20044H0441993-06-0202 June 1993 LER 93-009-00:on 930504,circuit Breaker Did Not Close During Planned Bus Transfer Due to Loose Control Circuit Wire.Lug Replaced & Reterminated on 930505 & Post Work Testing Completed W/Satisfactory results.W/930602 Ltr ML20024G7451991-04-24024 April 1991 LER 91-005-00:on 910325,diesel Generator Inoperable,Causing Loss of Ac Power to Train B Components of Safety Sys & Actuating Portions of Primary/Secondary Containment Sys. Caused by Voltage Regulator failure.W/910424 Ltr ML20029A6721991-02-25025 February 1991 LER 91-001-00:on 910125,automatic Primary Containment Isolation Control Sys Group 5 Actuation Occurred,Resulting in Closing of RCIC Turbine Steam Supply Isolation Valves. Caused by High Flow conditions.W/910225 Ltr ML20028G9731990-09-20020 September 1990 LER 89-036-01:on 891122,determined That HPCI Sys Inoperable Due to Inoperable Gland Seal Condenser Blower Motor.Caused by age-related Wear of Motor.Blower Motor Replaced & Blower Tested W/Satisfactory results.W/900920 Ltr ML20028G9121990-09-18018 September 1990 LER 89-037-01:on 891130,primary Containment/Traversing in-core Probe Ball Valve Discovered to Be Almost Full Open. Caused by Damage to Valve Stem Due to Manual Manipulation of Stem.Ball Valve & Solenoid Actuator replaced.W/900918 Ltr ML20043G3541990-06-12012 June 1990 LER 90-008-00:on 900513,automatic Scram Resulting from Load Rejection Occurred While at Full Power,Resulting in Trip of Generator Field Breakers.Caused by Fault on Offsite 345 Kv Transmission sys.Loss-of-field Relay replaced.W/900612 Ltr ML20043F8291990-06-0606 June 1990 LER 90-007-00:on 900507,discovered That Drywell to Suppression Chamber Vacuum Breaker Surveillance Not Performed Prior to Startup in 1988.Caused by Misunderstanding of requirements.W/900606 Ltr ML20042F5911990-04-30030 April 1990 LER 90-006-00:on 900330,determined That Position of Primary Containment Sys Isolation Valve Not Recorded Daily,Per Tech Specs.Review Performed to Determine If Other Problems Existed.Plant Shut Down & Review performed.W/900430 Ltr ML20012F5751990-04-0606 April 1990 LER 90-003-00:on 900311,automatic Actuation of Main Steam Sys Group 1 Portion of Primary Containment Isolation Control Sys Occurred.Caused by False High Reactor Vessel Water Level Signal.Procedure Developed Re backfill.W/900406 Ltr ML20012E9831990-03-30030 March 1990 LER 90-002-00:on 900228,determined That Max Fraction of Limiting Power Density Not Checked Daily During Reactor Power Operation,Per Tech Spec 4.1.B.On 900323,addl Tech Spec Issue Discovered.Tech Specs changed.W/900330 Ltr ML20012C4871990-03-12012 March 1990 LER 90-001-00:on 900209,24 H Limiting Condition for Operation Entered When Two RCS Instrumentation Excess Flow Check Valves Inappropriately Verified Operable During Testing.Cause Undetermined.Two Valves replaced.W/900312 Ltr ML20005G0981990-01-0808 January 1990 LER 89-038-00:on 891208,unplanned Automatic Reactor Protection Sys Scram Signal & Reactor Scram Occurred.Caused by False Low Reactor Vessel Water Level Signal.Local Level Indicators Satisfactorily calibr.W/900108 Ltr ML20005F8481990-01-0808 January 1990 LER 89-039-00:on 891209,automatic Actuation of RHR Sys Portion of Primary Containment Isolation Control Sys Occurred.Caused by Hydrodynamic Transient That Actuated Protective High Pressure switches.W/900108 Ltr ML20005E8551989-12-30030 December 1989 LER 89-037-00:on 891130,discovered That Traversing in-core Probe Ball Valve,Mfg by Consolidated Controls,Inc,Closed,In Violation of Tech Specs.Caused by Damage to Valve Stem.Ball Valve & Actuator replaced.W/891230 Ltr ML20011D1361989-12-11011 December 1989 LER 89-035-00:on 891111,inadvertent Actuation of Portion of Secondary Containment Sys During Surveillance Testing Occurred.Caused by Location of Radiation Isolation Control Sys Channel a Logic Relay.Procedure revised.W/891211 Ltr ML20005D7421989-12-0606 December 1989 LER 89-033-00:on 891106,automatic Actuation of Group 1 Portion of Primary Containment Isolation Control Sys (PCIS) Occurred.Caused by High Reactor Vessel Water Level.Manual Valve Closed & PCIS Logic Circuitry reset.W/891206 Ltr ML19351A4581989-12-0606 December 1989 LER 89-034-00:on 891108,determined That Reactor Pressure Exceeded 150 Psig on 891107 W/O Performing Surveillance Procedure 8.5.4.4, HPCI Valve Operability Test. Caused by Personnel Error.Procedure initiated.W/891206 Ltr ML19332D1421989-11-22022 November 1989 LER 89-031-00:on 891024,closing Times for Two in-series Primary Containment Isolation Valves Exceeded Acceptance Criteria During Shutdown Testing.Cause Undetermined.Closing Time Retested W/Satisfactory results.W/891122 Ltr ML19332D6231989-11-20020 November 1989 LER 89-032-00:on 891020,pneumatic Pressure Drop for Two of Four Automatic Depressurization Sys Accumulators Discovered Greater than Max Acceptable Value.Caused by Leakage from Seat of Supply Check Valves.Seat replaced.W/891120 Ltr ML19324C3181989-11-0909 November 1989 LER 88-002-01:on 880117,full Reactor Protection Sys Scram Trip Signal Occurred During Surveillance Test,Resulting in Incomplete Actuations.Caused by Procedure Inadequacy.Logic Relay Replaced & Procedure revised.W/891109 Ltr ML19324B1131989-10-21021 October 1989 LER 89-030-00:on 890926,control Room High Efficiency Air Filtration Sys Flowrate non-conservative Due to Procedure Error.Caused by Transcription Error.Procedure Revised & Test Reperformed Using Corrected procedure.W/891021 Ltr ML19325D4681989-10-16016 October 1989 LER 89-029-00:on 890914,locked High Radiation Area Door Found to Have Been Unsecured Contrary to Tech Spec 6.13.2. Caused by Failure to Adhere to Approved Procedures. Training on Access Requirements initiated.W/891016 Ltr ML19325D1861989-10-10010 October 1989 LER 89-028-00:on 890907,HPCI Sys Declared Inoperable Because Mechanical Overspeed Trip Occurred During Surveillance Test. Caused by Failure of Ramp Generator Signal Converter Module. Module Removed & Sent to Mfg for exam.W/891010 Ltr ML17277B8221984-07-20020 July 1984 LER 82-049/01X-1:on 821102,main Steam Line Safety Valve a Set W/Nitrogen at Steam Relief Setting.Caused by Personnel Error.Valve Removed & Replaced W/Properly Set Spare Valve. Procedure revised.W/840720 Ltr ML20024C3231983-06-24024 June 1983 Updated LER 82-038/03X-1:on 820817,RCIC Steam Line High Differential Pressure Alarm Received.Caused by Air in Sensing Lines to Pressure Switches.Cross Threaded Cap Found on in-line Snubber.Snubber replaced.W/830624 Ltr ML20024C2211983-06-24024 June 1983 Updated LER 80-053/03X-1:on 800814,0907 & 14,HPCI Turbine Exhaust Line Snubber 23-3-36 Determined Inoperable.Caused by Transient Hydrodynamic Shock During Startup & Shutdown. Snubbers upgraded.W/830624 Ltr ML20024B8451983-06-24024 June 1983 LER 83-031/03L-0:on 830528,HPCI Sys Made Inoperable to Repair Steam Leak on AO 2301-31.Caused by Leaking Stem Packing.Packing Replaced.Valve Tested Satisfactorily.Sys Returned to svc.W/830624 Ltr ML20024B8321983-06-24024 June 1983 LER 83-030/03L-0:on 830527,both HPCI Area Smoke Detection Sys Alarmed in Main Control Room.Caused by Valve A02301-31 Steam Leak Causing False Indications.Valve repaired.W/830624 Ltr ML20024C3101983-06-21021 June 1983 Updated LER 79-007/03X-1:on 790212,while Performing Test 8.7.4.3,valves AO-5033A,CV-5065-10,15,16 & 17 Failed to Give Proper Position Indication When Cycled & to Close within Specified Tolerance.Cause Not stated.W/830621 Ltr ML20024A8861983-06-16016 June 1983 LER 83-027/03L-0:on 830518,during Routine Insp,Folding Fire Door Found Inoperable.Caused by Operating Cable Becoming Detached from Pulley Due to Slack.Cable Reattached & Operating Mechanism adjusted.W/830616 Ltr ML20023D1691983-05-0909 May 1983 Updated LER 82-008/01X-1:on 820331,HPCI Injection Valve MO-2301-8 Failed to Open in Required Manner.Caused by Missing Wire Jumper Intended to Bypass Motor Operator Torque Switch.Jumper Installed & Valve Tested Satisfactorily ML20023D0671983-05-0909 May 1983 LER 83-021/03L-0:on 830411,during Surveillance Test 8.7.4.3, Primary Containment Isolation Valve AO-5033B Failed to Meet 10-s Closing Time Requirements.Caused by Lubricant Infiltrating Solenoid Valve Components.Components Replaced ML20023D0181983-05-0404 May 1983 LER 83-022/03L-0:on 830413,overload on Valve Motor 1301-17 Resulted in Loss of High Pressure Core Cooling Backup Capability.Caused by Oxidation in Motor End Bell & Motor Brushes,Due to Humidity Caused by Leaking Valve Packing ML20028G1591983-01-28028 January 1983 LER 83-002/01T-0:on 830116,flow Ref off-normal Alarm Received.Average Power Range Monitor Flux Scram Trip Settings Affected.Cause Under Investigation.Flow Inputs Monitored & Amplifier B Recalibr ML20028F4121983-01-17017 January 1983 LER 83-002/01X-0:on 830116,while Clearing off-normal Alarm for Flow Comparators,Average Power Range Monitor Flux Scram Trip Settings Found in Nonconservative Direction.Caused by Drifting Proportional Amplifiers.Settings Checked Daily ML20028B0421982-11-16016 November 1982 LER 82-049/01T-0:on 821102,main Steam Line Safety Valve a Set W/Nitrogen at Steam Relief Setting of 1240 Psig, Exceeding Tech Specs.Caused by Personnel Error.Set Valve Removed,Procedure Adherence Stressed & Procedure Revised ML20027C9801982-10-20020 October 1982 LER 82-039/03L-0:on 820920,timing Tests of Three Reactor Water Cleanup Isolation Valves 1201-2,-5 & -80 Not Completed within Allowed Time.Caused by Operational Constraints Requiring Testing at Reduced Power ML20027C9941982-10-20020 October 1982 LER 82-040/03L-0:on 820920,found Reactor Protection Sys Initiation Function of Turbine Stop Valve Completed 1 Day Late.Caused by Decision to Defer Test Due to MSIV D Line Isolation.Tracking Sys Implemented to Prevent Recurrences ML20052B9141982-04-23023 April 1982 LER 82-010/03L-0:on 820326,during Review of Testing Requirements,Surveillance Test 8.I.4 Standby Liquid Control Sys Found Not Verified.Caused by Breakdown in Multidiscipline Review of Results ML20052F2071982-04-14014 April 1982 Updated LER 81-060/03X-1:on 811020,determined Potential Exists to Reduce Number of Operable APRM & IRM Instrument Channels in Rod Block Trip Sys to Below Tech Spec Min ML20050A6321982-03-19019 March 1982 Updated LER 82-055/01T-1:on 810926,during Shutdown,Yarway Level Indicators Started Oscillating.Drywell Temp Was 240 F at 81-ft W/Elevation Coolant Temp 220 F.Caused by Ineffective Drywell Cooling.Ventilation Returned to Svc ML20050A4521982-03-19019 March 1982 LER 82-006/01T-0:on 820304,while Performing Maint Re SIL 352 on Pressure Adjustment Mechanism HPCI Stop Valve,Three Broken cap-screws Found.No Cause Determined.Screws Examined, Replaced & Sys Returned to Svc ML20050A4331982-03-19019 March 1982 Updated LER 81-062/01X-1:on 811116,Wyle Labs Informed Util That Two Target Rock Safety Relief Valves Had Not Passed Setpoint Tests.Caused by Setpoint Shift Due to Changes to Designed Differential Forces from Excessive Pilot Leakage 1993-07-28
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217E3021999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Pilgrim Nuclear Station.With ML20212C2921999-09-16016 September 1999 SER Accepting Licensee Request for Relief from ASME Code Section XI Requirements as Endorsed by 10CFR50.55a for Containment Insp for Pilgrim Nuclear Power Station ML20216F3511999-09-0808 September 1999 ISI Summary Rept for Refuel Outage 12 at Pnps ML20216E6881999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Pilgrim Nuclear Power Station.With ML20210R3401999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Pilgrim Nuclear Power Station.With ML20209C4731999-07-0707 July 1999 Addendum to SE on Proposed Transfer of Operating License & Matls License from Boston Edison Co to Entergy Nuclear Generation Co ML20209H8251999-07-0101 July 1999 Provides Commission with Evaluation of & Recommendations for Improvement in Processes Used in Staff Review & Approval of Applications for Transfer of Operating Licenses of TMI-1 & Pilgrim Station ML20209E6191999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Pilgrim Nuclear Power Station.With ML20196H2451999-06-29029 June 1999 SER Denying Licensee Proposed Alternative in Relief Request PRR-13,rev 2.Staff Determined That Proposed Alternative Provides Insufficient Info to Determine Adequacy of Scope of Implementation ML20209A8901999-06-28028 June 1999 SER Accepting Licensee Proposed Alternative to Use Code Case N-573 for Remainder of 10-year Interval Pursuant to 10CFR50.55a(a)(3)(i) ML20209B9861999-06-23023 June 1999 Rev 13A to Pilgrim Nuclear Power Station COLR for Cycle 13 ML20217N9061999-06-21021 June 1999 Rept of Changes,Tests & Experiments for Period of 970422-990621 ML20195K3431999-06-15015 June 1999 Safety Evaluation Granting Licensee Request to Use Guidance of GL 90-05 to Repair Flaws in ASME Class 3 Salt Svc Water System Piping for Plant ML20195G8231999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Pnps.With ML20207E7471999-05-27027 May 1999 Safety Evaluation Granting Request Re Reduction of IGSCC Insp of Category D Welds Due to Implementation of HWC to License DPR-35 ML20206M1971999-05-11011 May 1999 SER Accepting Request for Approval to Repair Flaws in ASME Code Class 3 Salt Svc Water Piping at Plant ML20206J6611999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Pilgrim Nuclear Power Station.With ML20205L0221999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Pilgrim Nuclear Power Station.With ML20207J5471999-03-0909 March 1999 Training Simulator,1999 4-Yr Certification Rept ML20207F9401999-03-0101 March 1999 Long Term Program Semi-Annual Rept for Pilgrim Nuclear Power Station ML20207H5451999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Pilgrim Nuclear Power Station.With ML20196E2151998-12-31031 December 1998 1998 Annual Rept for Boston Edison & Securities & Exchange Commission Form 10-K Rept.With ML20206Q2741998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Pilgrim Nuclear Power Station.With ML20197J3591998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Pilgrim Nuclear Power Station.With ML20195C9951998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Pilgrim Nuclear Power Station.With ML20154K0721998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Pilgrim Nuclear Power Station.With ML20153D3901998-09-22022 September 1998 Safety Evaluation Granting 970707 Request to Use Guidance in GL 90-05 to Repair Flaws in ASME Class 3 Salt Svc Water Sys Piping for Pilgrim Nuclear Power Station ML20197C5011998-09-0404 September 1998 Rev 12C,Pages 4 & 5 to Pilgrim Nuclear Power Station Colr ML20197C5471998-08-31031 August 1998 Rev 12C to Pilgrim Nuclear Power Station Colr ML20151W8231998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Pilgrim Nuclear Power Station.With ML20237E2251998-08-26026 August 1998 Suppl & Revs to SE for Amend 173 for Pigrim Nuclear Power Station ML20237A9941998-07-31031 July 1998 Monthly Operating Rept for Pilgrim Nuclear Power Station ML20236U8201998-07-13013 July 1998 Rev 12B to Pilgrim Nuclear Power Station COLR (Cycle 12) ML20236P0151998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Pilgrim Nuclear Power Station ML20249A3741998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Pilgrim Nuclear Power Station.W/Undated Ltr ML20247H2081998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Pilgrim Nuclear Power Station ML20207B7601998-03-31031 March 1998 Final Rept, Pilgrim Nuclear Power Station Site-Specific Offsite Radiological Emergency Preparedenss Prompt Alert & Notification System Quality Assurance Verification, Prepared for FEMA ML20216G3911998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Pilgrim Nuclear Power Station ML20216J3741998-03-19019 March 1998 Safety Evaluation Accepting Licensee Request to Evaluate Elevated Tailpipe Temp on Safety Relief Valve SRV 203-3B ML20248L2241998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Pilgrim Nuclear Station ML20202G5251998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Pilgrim Nuclear Power Station ML20236M8511997-12-31031 December 1997 1997 Annual Rept for Boston Edison & Securities & Exchange Commission Form 10-K Rept ML20198L7701997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Pilgrim Nuclear Power Station ML20203D6101997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Pilgrim Nuclear Power Station ML20202D5761997-11-0808 November 1997 1997 Evaluated Exercise BECO-LTR-97-111, Monthly Operating Rept for Oct 1997 for Pilgrim Nuclear Power Station1997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Pilgrim Nuclear Power Station ML20217D6431997-10-0101 October 1997 Safety Evaluation Granting Request for Approval to Repair Flaws in Accordance W/Gl 90-05 for ASME Class 3 SSW Piping for Pilgrim ML20217H5621997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Pilgrim Nuclear Power Station ML20216J4131997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Pilgrim Nuclear Power Station ML20210J3321997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Pilgrim Nuclear Power Station 1999-09-08
[Table view] |
Text
. _ . - . _ . _ . . .._. _ _ _ _ _ _ . . _ . _ . . . _ _ _ . _ _
9 10 CFR 50.73 2
[],
BOSTON EDISON ,
, Pdgnm Nuclear Power Staten Rocky Hill Road Plymouth, Massachusetts o2360 E. T. Boulette, PhD Senior Vice President-Nuclear June 18 , 1993 '
BEco Ltr. 93- 75 U.S. Nuclear Regulatory Commission !
Attn: Document Control Desk Washington, D.C. 20555 i
Docket No._50-293 l License No. DPR-35 l
i The enclosed Licensee Event Report.(LER) 93-010-00, " Start Up Transformer Became De-energized During Testing While Shut Down", is-submitted in accordance with 10 CFR Part 50.73.
Please do not hesitate to contact me if there are' any questions regarding'this report. ,
bf hbu$>$
E. T. Boulette, PhD l
RAG /bal u
Enclosure:
LER 93-010-00 cc: Mr. Thomas T. Martin Regional Administrator, Region I
'U.S. Nuclear Regulatory. commission
( 475 Allendale Rd.
l King of Prussia, PA 19406 I
l Mr. R. B. Eaton l Div. of Reactor Projects I/II Office of NRR - USNRC One White. Flint North - Mail Stop 14D1 11555 Rockville Pike Rockville, MD 20852 Sr. NRC Resident Inspector.- Pilgrim Station Standard BEco LER Distribution 9306280072 93061s /j' I PDR - ADOCK'05000293
.S 'PDR
_ . _ .___ _ _ _ . _ _ _ _ _ _ _. __ ._. - ~ .
_ . . . . . _ _ __m . _ _ _ _
l NRC FORM 366 U.S. NUCLEAR REGULAVORY COMMISSION APPROVED BY OMB NO. 3150-0104 t s.m .
)
, EXPIRES 5/31/95
. ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) 1 lIs f"nEa % % % E E Z J f0 M "3's !
%ia*" "SSC="#"o'C A"A",
TO THE PAPERWORK REDUCTION PRG.iECT p150-0104) OFFICE l
(See reverse for number of degrts/ characters for each bloem) OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY N'AME (1) DOCKET NUMBER (2) PAGE (3)
PILGRIM NUCLEAR POWER STATION 05000 -293 1 of 7 ;
YlTLE (4) ,
I Start Up Transformer Became De-energized During Testing While Shut Down EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
I SEQUENTIAL REVISION FACluTY NAME DOCKET NUMBER ,
MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR N/A 05000 l F ACluTY NAME DOCKET NUMBER 05 19 93 93 010 00 06 18 93 N/A 05000 !
OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF to CFR S: (Check one or more)(11)
DE@) N 20.402(b) 20.405(c) X So.73(a)(2)(iv) 73.71(b)
PO R 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)
LE !
O) 000 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) ,
20.405(a)(1)(iv) 50.73(a)(2)(ii) So.73(a)(2)(viii)(B) yt;c{g 20.405fa)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) Form 366A)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (!nclude Area Code)
Robert A. Gay - Senior Compliance Engineer (508) 747-8047 ,
COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN TH!S REPORT (13) l CAUSE SYSTEM COMPONENT MANUFACTURER P CAUSE SYSTEM COMPONENT MANUFACTURER S SUPPLEMENTAL REPORT EXPECTED (14) ""'" U## #
EXPECTED YES X NO SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE)
DATE (15)
ABSTRACT (16)
On May 19, 1993, at 1328 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.05304e-4 months <br />, the Start Up Transformer (SUT) became de-energized during a planned calibration of turbine / generator relays, lockout test and associated annunciator verification while the plant was shutdown. The Start Up Transformer lockout relay was actuated instead of the Main Transformer lockout relay. The de-energization of the SVT resulted in designed responses including automatic actuations of the Reactor Protection System, Primary and Secondary Containment Isolation Control Systems and Emergency Diesel Generators. The Start Up Transformer was re-energized by 1421 hours0.0164 days <br />0.395 hours <br />0.00235 weeks <br />5.406905e-4 months <br />.
The root cause of the event was utility non-licensed Electric Lab Engineer error. The engineer mistakenly actuated the wrong relay while implementing the test procedure.
Corrective action taken included an immediate discussion with electrical maintenance personnel on attention to detail and self-checking. Preventive action planned includes review of the procedure for improvement to preclude recurrence.
The event occurred while shut down in a refueling outage. The Reactor Vessel (RV) head was installed. The reactor mode selector switch was in the SHUTDOWN position. There was no movement of a fuel assembly or fuel cask at the time of the event. This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv). This event posed no threat to the public health an:1 safety.
NRC FORM 368A l5-92)
( 1 l NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-o104 !
j .
( s-m . EXPIRES 5/31/95 ESTIMATED BURDEN PER FIEJPONSE TO COMPLY WrTH THIS LICENSEE EVENT REPORT (LER) M#lEo d'o" "R# 3 *S J L %?iN # 2 2 1 "T^
TEXT CONTINUATION E5 TORY $u"u^Ys"EENaTc"N"" ""cSsU"o">oDN TO THE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE ;
OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503 FACIUTY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR NUMB PILGRIM NUCLEAR POWER STATION 05003-293 93 --010- 00 TEXT (if more space is required, use additional copies of NRC Form 366A)(17)
BACKGROUND i The Auxiliary Power Distribution System (APDS) consists of six 4160 VAC buses. The APDS is divided into emergency service (Buses A5 and A6) and normal service (Buses Al, A2, A3, A4). Buses A5 and A6 supply power to essential loads required during normal operations and abnormal operational transients or accidents. Buses A1, A2, A3, A4 supply power to other station auxiliaries during planned operations. Power is distributed to the six 4160 VAC buses during normal operation from either the unit source (Unit Auxiliary Transformer) or the preferred offsite source (Start Up Transformer). The preferred power source is used to supply the 4160 VAC buses during normal startup and shutdown. After the main generator has been synchronized to the 345 kV transmission system, the 4160 buses are transferred from the preferred power source to the unit power source. The 4160 VAC emergency service Buses A5 and A6 can also be supplied from the standby power source PDG
'A' and 'B') or the secondary power source (Shutdown Transformer).
During conditions when neither the UAT nor the SUT are available, the emergency diesel generators are designed to power emergency Buses A5 and A6. If the diesel generators fail to pick up the emergency buses, the Shutdown Transformer is designed to automatically l power Bus A5 and A6 after a 12 second time delay. The station blackout diesel generator l may be started to power Bus A5 or Bus A6.
Non-safety related 4160 V Buses Al, A2, A3, and A4 are designated normal service system buses. These buses power loads necessary for normal plant startup, operation, and shut down. These buses can be powered only by the UAT or SUT.
4160 V buses A1, A3, and A5 are located in the Turbine Building Switchgear Room 'A'.
Buses A2, A4, and A6 are located in Switchgear Room 'B'.
Just prior to the event, plant system status was: !
l l o The reactor mode selector switch was in the REFUEL position. The control rods were fully inserted for the core locations where fuel was installed. The RV head was i installed with RV pressure at zero psig. The Reactor Vessel (RV) water temperature was approximately 91 degrees Fahrenheit. There was no movement of a fuel cask or irradiated fuel . The Residual Heat Removal (RHR) System Loop ' A' was in service in the Shutdown Cooling (SDC) mode.
l e The 345 kV switchyard ring bus was energized via transmission lines 342 and 355. Air circuit breakers 102, 103, 104 and 105 were closed. Bus B6 was being powered from Bus 81. 4160 VAC buses that were in service were energized via the SVT. The Shutdown Transformer was in standby service.
o The Emergency Diesel Generators (EDGs) 'A' and 'B' were in standby service. The APDS
- was energized by preferred offsite power via the SUT. The 23 kV distribution system l was energized.
NRC FORM 366A (b-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 o-m . EXPIRES 5/31/95
. ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) ETE"0 1 3$'S?u'J, E Jk % 5M"*^2 TEXT CONTINUATION ^"Su"'A RE CC M75 CN?[C W OC 3,1="*T =S'."r~Ka##? ""
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR R'^ NU PILGRIM NUCLEAR POWER STATION 05000-293 93 -010- 00 TEXT (tf more space is required. use additional copies oi NRC Form 366A)(17)
- The RV was vented to the Drywell via RV-203-4B flange (valve removed for ILRT).
Located at the end of this report is a Figure that depicts a simplified single line diagram of a portion of the APDS.
EVENT DESCRIPTION i On May 19, 1993, at 1328 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.05304e-4 months <br />, the Start Up Transformer became de-energized. The SUT i became de-energized because the SUT lockout relay 86X/ST was actuated instead of the Main Transformer lockout relay 86X/MT, which caused ACB 102 and ACB 103 to trip open and :
eliminate line 355 and line 342 as sources of preferred off-site power to the SUT. The ,
loss of sources of power to the SUT'resulted in the following:
- The APDS, energized by preferred off-site power via the SVT, became de-energized and '
resulted in the following:
- De-energizing / actuation of the RPS, PCIS and RBIS that included: '
- Multiple concurrent RPS scram signalc. The control rod drives remained i in the inserted position. '
The Group 2 valves, except for A0-5035 A/B that were jumpered open for ILRT, were closed and remained closed.
- The Group 3 valves MO-1001-47 and -50 that were open and jumpered for ILRT remained open. Valve M0-1001-29A, open for Residual Heat Removal / Shutdown Cooling, remained open because MO-1001-47 and -50 remained open.
- The Group 6 valves M0-1201-2, -5, and -80, closed for ILRT, remained closed.
- RBIS actuation that resulted in the automatic closing of the inboard and outboard Reactor Building /SCS ventilation supply and exhaust dampers and
, automatic start of the SGTS Trains 'A' and 'B'.
- EDGs 'A' and 'B' started automatically and re-energized Buses A5 and A6 and related AC powered load center buses, motor control centers, and distribution panel s.
NRC FORM 306A @ 92)
. _ . , , _ , _ _ _ , . . _ . , _ _ _ _ , , , _ , _ . . . _ . ~ , _ _ . _ . , _ . _ - . . . . , , . - . . . . _ , _ _ _ . , _ , .
.- - - -- -_~ - - . . . . _ . . . .-
I i
NRC FORM 366A U.S. NUCLEAR REGUI.ATORY COMMISSION APPROVED BY OMB NO. 3150-o104
] ,
(s-m . EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS
^
j LICENSEE EVENT REPORT (LER) L"E'1f" "'
a%"OJMJf0 " " "
Ar=12 '
- TEXT CONTINUATION ^ETORY c"$"ME"'W2sHYNoTO7Dc MsEfAN$
R TO THE PAPERWORK RLDUCTION PROJECT. (31504104), OFFICE )
OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. I
)i FACILITY NAME (1) - DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) ,
YEAR B M "H 4 Of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 -010- 00 i TEXT (if more space is requeed, use additional copies of NRC Form 366A)(17) 4 After initial investigation, Bus A5 was loaded on the SUT at 1405 hours0.0163 days <br />0.39 hours <br />0.00232 weeks <br />5.346025e-4 months <br />. RHR System Loop.
} 'A' was returned to service for Shut Down Cooling at 1412 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.37266e-4 months <br />. ~ Bus A6 was loaded on the '
1 SUT at 1421 hours0.0164 days <br />0.395 hours <br />0.00235 weeks <br />5.406905e-4 months <br />.
,! Problem Report 93.9260 was written to document the event. The NRC Operations Center was l j notified in accordance with 10 CFR 50.72 on May 19, 1993, at 1508 hours0.0175 days <br />0.419 hours <br />0.00249 weeks <br />5.73794e-4 months <br />.
i !
j CAUSE 1
! A critique of the event was conducted on May 19, 1993, d was attended by applicable !
] personnel including the Electric Lab engineer who was one of three individuals involved in
- performing the test. '
i j The root cause of the event was utility non-licensed Electric Lab engineer error. The j engineer mistakenly actuated SVT lockout relay 86X/ST instead 'of Main Transformer lockout
- relay 86X/MT. The error occurred at Step [12 (b)] of Procedure 3.M.3-39 (Rev. 14), y
- Attachment 1, " Calibration of Turbine / Generator Relays, Lockout Test and Associated ;
! Annunciator Verification". :
! A factor contributing to the error was that the rear board sections for the Main 4
Transformer and SUT are mirror images'of each other and are physically mounted back-to- ,
j back, with a working aisle separating the two. The manual actuation of relay 86X/ST was ,
performed at the rear of the relay panel. In addition, the lockout relays were
~
i individually labeled only as 86X at the rear of their respective panels.
)
CORRECTIVE ACTION-i Immediate corrective action taken included terminating the performance of Procedure
- 3.M.3-39 and stabilizing plant conditions in accordance'with Procedure 2.4.16 (Rev. 11),
j " Distribution Alignment Electrical System Malfunctions".
A PREVENTIVE ACTION 1 An immediate discussion was conducted with Electric Lab personnel on attention to detail i and self-checking. The discussions were completed prior to recommencing the performance j of Procedure 3.M.3-39 on May 19, 1993.
, Procedure 3.M.3-39 " Turbine / Generator Calibration of Relays, Lockout Test and Associated Annunciator Verification", will be reviewed for improvement to preclude recurrence. The focus of the review is to perform the applicable procedural steps from the front of the panel.
4 The rear of the Main Transformer and SVT board sections will have their labeling enhanced l i to make the two relays (86X/MT'and 86X/ST) more distinguishable from one another. i a (
} l NRC FORM 366A @92) i
I I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 a-m . EXPIRES 5/31/95 ESTIMATED BLh0EN PER RESPONSE TO COMPLY WITH THIS l LICENSEE EVENT REPORT (LER) 1"T5 biRo'3 0 & s LTUO O M Ts M l l TEXT CONTINUATION EEcEsI"$ leos""oE$sSs"$oNS l
$ E [o Nr E "ND E ASH G ,
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR NJ PILGRIM NUCLEAR POWER STATION 05000-293 93 --010- 00 l TEXT (If more space is required, use additional copies of NRC Form 366A)(17)
DISCIPLINARY ACTION l The Electric Lab engineer involved received a written reprimand identifying the apparent l cause of the incident and informing the engineer that any further problems would result in ;
more severe disciplinary action. 1 SAFETY CONSE0VENCES This event posed no threat to the public health and safety.
There were no component or system failures that caused or resulted from this event.
The Standby AC Power (4160 VAC) System consists of EDGs ' A' and 'B' that are self-contained and independent of the off-site power sources. The safety objective of the 1 Standby AC Power System is to provide a single failure proof source of onsite AC power l adequate for the safe shutdown of the reactor following abnormal operational transients and postulated accidents. A loss of all off-site power is described in the UFSAR Chapter
- 14. The Chapter 14 analysis bounds the analyses in the UFSAR Appendix R that includes a i loss of all off-site power to station auxiliaries. The EDGs started and provided power to Buses A5 and A6 and the related electrical system in response to the loss of power to Bus A5 and Bus A6.
The RHRS/SDC mode of operation has a power generation design b:tsis only. The SDC mode of operation functions to reduce the RV water temperature to 125 degrees Fahrenheit approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after a shutdown for refueling or servicing activities.
If the actuation had occurred when the decay heat generation rate was greater, a gradual increase in the RV water temperature could have occurred. However, alternate means for heat removal are available and described in procedure 2.4.25, " Loss of Shutdown Cooling",
including methods for feed and letdown using the Condensate System, RWCU System, and the Main Condenser.
I This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv) because the automatic actuation of EDG 'A' and 'B' was not planned.
SIMILARITY TO PREVIOUS EVENTS A review was conducted of Pilgrim Station Licensee Event Reports (LERs) submitted since January 1984. The review focused on LERs submitted in accordance with 10 CFR 50.73 (a)(2)(iv) involving automatic actuation of the Emergency Diesel Generators due to personnel error. The review identified related instances reported in LERs 50-293/84-012-00, 84-017-00 and 92-011-00.
l l
NRC FORM 366A (5 92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104
'N* EXP!RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH TH:S LICENSEE EVENT REPORT (LER) 1""0RGA;s'S!" "!!"L*LnO,~<%"=
""* " " ~ "
TEXT CONTINUATION $$u"EOnY $$$^OS80NTSHNGTC DC 0 TO THE FAPERWORK REDUCTION MnJECT (3150-0104, OFFCE OF MANAGEMENT AND BUDGET, WASHfNGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
^ '
YEAR B 7U B 6 Of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 --010- 00 l VEXT (It more space is required, use additional copies of NRC Form 366A)(17)
For LER 84-012-00, while shut down for refueling, the EDG 'B' was inadvertently started during Procedure 8.M.2-2.10.1-8 " Logic System Functional Test of the ' A' Core Spray System ;
- High Drywell Pressure Auto Initiation Trip". The event occurred when a non-licensed I&C '
Technician inadvertently actuated relays 14A-K5B and 14A-K6B simultaneously. These two
- relays are located side by side in Panel 933. This action completed the logic required to ,
start EDG 'B'. Corrective action included counselling the non-licensed I&C Technician on '
I attention to detail.
For LER 84-017-00, while shut down for refueling, a temporary loss of off-site power was experienced. Both emergency diesels automatically started and picked up the emergency j buses as required. The event occurred after a test of the turbine trip to 345 kV breaker i trip logic. Cause was determined to be utility non-licensed personnel error. The error i which occurred when two stuck breaker protection switches were activated without first !
resetting the corresponding stuck breaker lockout relays. Corrective action included l counselling of personnel to prevent recurrence.
For LER 92-011-00, while operating at 100 percent power, an unplanned actuation of a portion of the RHR System logic circuitry occurred during a functional test of the HPCI System logic circuitry. Emergency Diesel Generator 'B' started but did not load onto the respective 4160 VAC bus because the bus remained energized from the UAT. Cause of the event was utility non-licensed I&C Technician error. The error occurred when The RHR System Channel 'B' relay 10A-K10B was actuated instead of relay 10A-K88. Corrective action included counselling the responsible I&C Technician and revising Procedure 8.M.2-2.10.4-2 to preclude recurrence.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:
COMPONENTS CODES i
Circuit Breakers (ACB-102 & 103), AC 52 l Transformer (SUT) XFMR I Relay, Lccking-Out (86X/MT, 86X/ST) 86 SYSTEMS l Containment Isolation Control System (PCIS/RBIS) JM l Engineered Safety Features Actuation System (PCIS,RBIS,RPS) JE I
Emergency On-site Power Supply System (EDG ' A' & 'B') EK Medium-Voltage Power System 4.16 KV EA Plant Protection System (RPS) JC Reactor Water Cleanup (RWCU) System CE Residual Heat Removal (RHR) System B0 Standby Gas Treatment System (SGTS) BH Suitchyard System FK l
l NRC FORM 366A (5 92) l
l NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMD NO. 3150-0104 i s.en EXPlRES 5/31/95 ESTMATED BURDFN PER RESPONSE TO COMPLY W8TH THrs INFORMATON CO REQUEST: 50 0 HR$ FORWARD COMMEm, Res..LLECTION LICENSEE EVENT REPORT (LER) Nao m ~Es1. m TO we,.OsM.To~
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