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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20029E5121994-05-12012 May 1994 LER 93-008-01:on 930706,high Pressure Injection Suction Valve Determined to Be Inoperable.Caused by Lack of Engineering Review for Motor Brake Voltage Requirements. Valves Locked & Brake removed.W/940512 Ltr ML20029D7671994-05-0303 May 1994 LER 94-002-00:on 940404,performance of Surveillance Check Power Distibution Breaker Alignment & Power Availability Verification Resulted in Entry Into LCO 3.0.3.Caused by Procedure Error.Procedures revised.W/940503 Ltr ML20046C5581993-08-0404 August 1993 LER 93-008-00:on 930706,possibility of MOV W/Brakes Failing to Perform Safety Function Under Degraded Voltage Conditions Due to Lack of Engineering Review.Identified & Reviewed MOV W/Motor Brakes for Safety significance.W/930804 Ltr ML20046B4121993-07-28028 July 1993 LER 93-007-00:on 930628,noticed That Hydrogen Pressure Setpoint Found to Be Set Above 10 Psig.Caused by Personnel Error.Mod Design Package Developed to Raise Pressure Regulator setpoint.W/930728 Ltr ML20045F7741993-06-30030 June 1993 LER 93-001-01:on 930305,cooldown Exceeding Limits of TS 3.4.9.1 Experienced After Switching from SG Cooling to Dh Sys Cooling.Caused by Failure of Cv Controller.Valve Repaired & Valve Operation Instructions Revised ML20045D7741993-06-18018 June 1993 LER 93-006-00:on 930520,inadequately Secured Separation Barrier Identified That Could Affect Control Switches for safety-related Equipment on Main Control Board.Caused by Human Error.Placard Posted to Restrict area.W/930618 Ltr ML20045B4381993-06-10010 June 1993 LER 93-005-00:on 930518,notified That Control Circuit for Makeup & Purification Sys Letdown Isolation Valve Did Not Meet Electrical Isolation Criteria Due to Human Error.Mod in Development Will Be expanded.W/930610 Ltr ML20044D2101993-05-10010 May 1993 LER 93-004-00:on 930408,inappropriate Personnel Action Resulted in Degraded Class 1E Bus Voltage & Actuation of Edg.Licensee Established Addl Administrative Controls on Switchyard activities.W/930510 Ltr ML20044D1651993-05-10010 May 1993 LER 93-003-00:on 930408,determined That Insufficient Instrument Error Considered When Operating Limits for Core Flood Tank Selected.Caused by Programmatic Deficiency.Alarm Bistables Reset & Allowed Margin expanded.W/930510 Ltr ML20024H2001991-05-20020 May 1991 LER 91-003-00:on 910420,emergency Feedwater Actuation & Manual Reactor Trip Occurred.Caused by Water Instrusion Into Pump Loss of Circulating Water Pump.Plant Modification Previously initiated.W/910520 Ltr ML20024H1741991-05-17017 May 1991 LER 90-002-02:on 900216,determined That Fire Dampers May Not Be Operable Under Expected Ventilation Flow Conditions Due to Design Error.Design Basis Documents updated.W/910517 Ltr ML20029B0751991-02-28028 February 1991 LER 91-002-00:on 910129,startup Transformer Incapable of Maintaining Voltage Output Above Setpoint of Second Level Undervoltage Relays Under Es Actuation Conditions.New Offsite Power Supply installed.W/910228 Ltr ML20029A6531991-02-25025 February 1991 LER 91-001-00:on 910124,engineered Safeguards Train a HPI Recirculation Isolation Valve MUV-53 Declared Inoperable Due to Undersized Thermal Overload Elements for Motor Operator.Cause Undetermined.Elements replaced.W/910225 Ltr ML20028H7131991-01-21021 January 1991 LER 89-022-01:on 890614,erroneous Indication of Loss of Main Feedwater Pumps Occurred Resulting in Manual ESF Actuation. Temporary Loss of Supervisory Indicating Lights Expected During Bus Realignment.Procedures revised.W/910121 Ltr ML20043H5751990-06-22022 June 1990 LER 90-009-00:on 900523,discovered That Paperwork for Mod to Replace Feedwater Vent Valve Still Open 12 Months After Work Completed.Caused by Lost Blanket Work Request Re post-maint Testing.New Work Request initiated.W/900622 Ltr ML20043G6511990-06-14014 June 1990 LER 90-008-00:on 900514,determined That Fault Current Could Develop Across Engineered Safeguards 480-volt Bus Output Breakers.Caused by Failure to Perform Adequate Short Circuit Analysis.Deficient Trip Devices replaced.W/900614 Ltr ML20043C5221990-05-31031 May 1990 LER 89-035-01:on 890906,determined That Dc Powered Components Exhibited Discrepancy in Rated Voltages & Actual Voltages.Caused by Inadequate Control of Design Process.Inoperable Components replaced.W/900531 Ltr ML20043B8091990-05-23023 May 1990 LER 90-007-00:on 900112,discovered That Door Between Control Room Complex & Turbine Bldg Removed for Mod Work,Resulting in Inoperability of Both Trains of Emergency Ventilation Sys.Door Replaced & Warning Signs affixed.W/900523 Ltr ML20042G7641990-05-0909 May 1990 LER 90-006-00:on 900410,reevaluation of Design Calculations Discovered Deficient Valve Operator Installation.Sufficient Thrust Not Developed to Open or Close Valves Due to Undersized Spring Packs.Spring Packs ordered.W/900509 Ltr ML20042F9781990-05-0707 May 1990 LER 89-031-01:on 890828,480-volt Engineered Safeguards Stepdown Transformer a Faulted Causing Decay Heat Train Closed Cycle Cooling Pump a to Deenergize.Caused by Degraded Insulation.Transformer replaced.W/900507 Ltr ML20012D8091990-03-22022 March 1990 LER 90-003-00:on 900220,instrument Fluctuation Noted During Surveillance of Chilled Water Pump.Caused by Location of Flow Element within Chilled Water Sys.Relief Request Submitted & Field Problem Rept generated.W/900322 Ltr ML20012D0281990-03-19019 March 1990 LER 90-002-00:on 900216,determined That Fire Dampers May Not Be Operable Under Expected Ventilation Flow Conditions Due to Design Error.Caused by Failure of Original Design Criteria to Address Need.Documentation updated.W/900319 Ltr ML20011F1731990-02-21021 February 1990 LER 90-001-00:on 900221,RCS Leakage Calculations Indicated That Unidentified Leakage Exceeded Tech Spec Limits.Caused by Failed Packing on Block Valve.Valve Repacked During Feb 1990 Maint outage.W/900221 Ltr ML19354D9091990-01-17017 January 1990 LER 89-016-03:on 890426,administrative Problems Caused Deficiencies in Environ Qualification Program That Resulted in Plant Equipment Not Being Properly Qualified.Effort to Correct Environ Qualification Deficiencies Underway ML20006E3941990-01-0808 January 1990 LER 89-041-00:on 891208,partial,simultaneous Withdrawal of Two Control Rod Safety Groups Occurred & Safety Group 3 Control Transferred to Auxiliary Power Supply.Cause Undetermined.Select Relays replaced.W/900207 Ltr ML20005F1601990-01-0808 January 1990 LER 89-040-00:on 891208,emergency Diesel Generators a & B Actuated Due to Degraded Voltage When Condensate Pump Started.Operator Guidelines & Procedures for Starting Condensate Pumps Being revised.W/900108 Ltr ML20011D8401989-12-22022 December 1989 LER 89-030-01:on 890824,determined That Pump Discharge Pressure & Flow Less than Required.On 890826,plant Entered Hot Standby.Caused by Installation of Incorrect Impeller in Pump.Original Impeller Reinstalled Upon repair.W/891222 Ltr ML20011D2281989-12-15015 December 1989 LER 89-026-01:on 890629,emergency Diesel Generator 1B Failed post-maint Test & Could Not Be Returned to Svc within Time Allowed by Tech Specs.Caused by Rotor Not Turning.Crankcase Vapor Ejector cleaned.W/891215 Ltr ML19332E7691989-12-0707 December 1989 LER 89-039-00:on 891107,determined That Control Circuits for Two Makeup Valves Did Not Meet Separation Criteria of 10CFR50 App R.Caused by Cognitive Personnel Error.Roving Fire Watch Patrol Confirmed in effect.W/891207 Ltr ML19332F0171989-11-30030 November 1989 LER 89-038-00:on 891029,util Engineers Discovered That Administrative Controls Not in Place for Three Makeup & Purification Sys Valves.Caused by Personnel Error.Plant Procedures revised.W/891130 Ltr ML19332D1391989-11-27027 November 1989 LER 89-037-00:on 891026,determined That Instrumentation Used to Balance HPI Flow Through Four Injection Lines During Small Break LOCA Inadequate.Caused by Inadequate Review of B&W Guidelines.Flow Instrument installed.W/891127 Ltr ML19327C2611989-11-17017 November 1989 LER 89-036-00:on 891018,determined That Plant Operating Outside Design Basis Since Borated Water Storage Tank Level Transmitters Not Seismically Qualified.Caused by Personnel Error.Unqualified Transmitters replaced.W/891117 Ltr ML19325F3371989-11-10010 November 1989 LER 89-033-00:on 890908,second Level Undervoltage Relay Sys Setpoint for Engineered Safeguards Buses Not Conservative & Led to Operation Outside Plant Design Basis.Caused by Personnel Error.Setpoint changed.W/891110 Ltr ML19325F3341989-11-10010 November 1989 LER 89-035-00:on 890906,discrepancy Noted in dc-powered Component Rated Voltages & Actual Voltages Seen by Components.Caused by Inadequate Control of Design Process. All Components Replaced Prior to startup.W/891110 Ltr ML19325E7371989-10-27027 October 1989 LER 85-034-01:on 850310,valve Alarm Function for Core Flood Tank Isolation Valves Failed to Meet Acceptance Criteria.On 850809,unit Entered Mode 3 & Then Raised RCS Pressure Above 750 Psig W/O Meeting Operability surveillances.W/891027 Ltr ML19324B3921989-10-26026 October 1989 LER 88-002-02:on 880107,emergency Feedwater Actuation Occurred on Loss of Both Main Feedwater Pumps.Caused by Instrumentation & Control Technician Error.Idle Feedwater Pump Started & Actuation reset.W/891026 Ltr ML19324B3871989-10-26026 October 1989 LER 89-034-00:on 890926,two Conditions Determined to Be Outside Plant Design Basis Re Solenoid Control Valves.Caused by Cognitive Personnel Error.Test Solenoid Valve Circuits Provided W/Isolation fuses.W/891026 Ltr ML19327B2301989-10-23023 October 1989 LER 89-016-02:from 890227-0601,deficiencies Re Environ Qualification of Plant Equipment Discovered.Caused by Deficiencies in Detailed Development & Implementation of Environ Qualification Program.Program reviewed.W/891023 Ltr ML20024D2501983-07-26026 July 1983 Updated LER 83-019/03L-1:on 830406,breaker in Engineered Safeguards Motor Control Ctr 3B1 Shorted Out,Causing Various Pieces of Equipment on Train B to Be Inoperable.Caused by Personnel Error.Breaker Removed & cleaned.W/830726 Ltr ML20024D1491983-07-26026 July 1983 Updated LER 82-003/01T-0:on 820128,RCS Leakage Calculations Showed Unidentified Leakage at Rate Greater than 1 Gpm. Caused by Thermally Induced Cyclic Failure.New Valve Will Be Installed in Improved location.W/830726 Ltr ML20024D1061983-07-26026 July 1983 Updated LER 82-004/01T-1:on 820129,while Performing Visual Insp of Reactor Coolant Pump a Seal Package,Rcpb Leakage Discovered from Crack in Seal Weld.Caused by Installation Error.Seal Replaced & Procedures revised.W/830726 Ltr ML20024C0951983-06-27027 June 1983 LER 83-023/01T-0:on 830613,fire Damper FD-86,in Duct Work Between Auxiliary Bldg & Control Complex,Discovered Missing. Cause Unknown.Continuous Fire Watch Established.Evaluation underway.W/830627 Ltr ML20023E0971983-06-0101 June 1983 LER 83-021/03L-0:on 830504,three Hydraulic Snubbers Failed Functional Test During Dec Outage & 107 Snubbers Failed Test During Refuel Iv.Caused by Seal Failure & Valve Assembly Contamination.All Snubbers modified.W/830531 Ltr ML20023C5251983-05-0909 May 1983 LER 83-020/03L-0:on 830409,discovered That Surveillance Interval for Fire Detection Instrumentation in Emergency Diesel Generator & Control Rooms Exceeded by 6 Months.Caused by Procedural Inadequacy.Procedure SP-411 Will Be Revised ML20023B4091983-04-27027 April 1983 LER 83-018/01T-0:on 830413,preliminary Repts Received from B&W Indicating That 51 of 120 Upper Core Barrel Bolts Ultrasonically Tested May Be Defective.Cause Unknown. Investigation Underway.Supplemental Rept Will Be Written ML20028E0561983-01-13013 January 1983 LER 82-075/03L-0:on 821215,during Cold Shutdown,Circuit Breaker on Control Complex Ventilation Radiation Monitor RMA-5 Tripped.Caused by Broken Pump Vane Binding Pump. Pump Replaced ML20028A5321982-11-10010 November 1982 LER 82-063/03L-0:on 821011,reactor Bldg Average Air Temp Exceeded 130 F Limit.Caused by Strain on Instrument Air Line Causing Line to Split.Line Replaced on 821011 ML20027D7021982-10-29029 October 1982 LER 82-061/03L-0:on 820929,fan Damper Operator on Industrial Cooler Failed,Resulting in Reactor Bldg Average Air Temp Exceeding 130 F.Caused by Water in Instrument Air Sys Due to Personnel Failing to Close Valve FSV-250.Dampers Wired Open ML20027B8881982-09-24024 September 1982 LER 82-055/03L-0:on 820825,feedwater Ultrasonic Flow Indicator FW-313-FI Found Inoperable.Caused by Instrument Failure Due to High Ambient Temp at Instrument Cabinet Location.Instrument Repaired.Flow Transmitters Replaced ML20027B9001982-09-24024 September 1982 LER 82-056/03L-0:on 820827,during Normal Operation,Primary Containment Average Air Temp Exceeded 130 F Tech Spec Limit. Caused by Failure of Pneumatic Control Line for Industrial Coolers.Control Line Replaced & Coolers Returned to Svc 1994-05-03
[Table view] Category:RO)
MONTHYEARML20029E5121994-05-12012 May 1994 LER 93-008-01:on 930706,high Pressure Injection Suction Valve Determined to Be Inoperable.Caused by Lack of Engineering Review for Motor Brake Voltage Requirements. Valves Locked & Brake removed.W/940512 Ltr ML20029D7671994-05-0303 May 1994 LER 94-002-00:on 940404,performance of Surveillance Check Power Distibution Breaker Alignment & Power Availability Verification Resulted in Entry Into LCO 3.0.3.Caused by Procedure Error.Procedures revised.W/940503 Ltr ML20046C5581993-08-0404 August 1993 LER 93-008-00:on 930706,possibility of MOV W/Brakes Failing to Perform Safety Function Under Degraded Voltage Conditions Due to Lack of Engineering Review.Identified & Reviewed MOV W/Motor Brakes for Safety significance.W/930804 Ltr ML20046B4121993-07-28028 July 1993 LER 93-007-00:on 930628,noticed That Hydrogen Pressure Setpoint Found to Be Set Above 10 Psig.Caused by Personnel Error.Mod Design Package Developed to Raise Pressure Regulator setpoint.W/930728 Ltr ML20045F7741993-06-30030 June 1993 LER 93-001-01:on 930305,cooldown Exceeding Limits of TS 3.4.9.1 Experienced After Switching from SG Cooling to Dh Sys Cooling.Caused by Failure of Cv Controller.Valve Repaired & Valve Operation Instructions Revised ML20045D7741993-06-18018 June 1993 LER 93-006-00:on 930520,inadequately Secured Separation Barrier Identified That Could Affect Control Switches for safety-related Equipment on Main Control Board.Caused by Human Error.Placard Posted to Restrict area.W/930618 Ltr ML20045B4381993-06-10010 June 1993 LER 93-005-00:on 930518,notified That Control Circuit for Makeup & Purification Sys Letdown Isolation Valve Did Not Meet Electrical Isolation Criteria Due to Human Error.Mod in Development Will Be expanded.W/930610 Ltr ML20044D2101993-05-10010 May 1993 LER 93-004-00:on 930408,inappropriate Personnel Action Resulted in Degraded Class 1E Bus Voltage & Actuation of Edg.Licensee Established Addl Administrative Controls on Switchyard activities.W/930510 Ltr ML20044D1651993-05-10010 May 1993 LER 93-003-00:on 930408,determined That Insufficient Instrument Error Considered When Operating Limits for Core Flood Tank Selected.Caused by Programmatic Deficiency.Alarm Bistables Reset & Allowed Margin expanded.W/930510 Ltr ML20024H2001991-05-20020 May 1991 LER 91-003-00:on 910420,emergency Feedwater Actuation & Manual Reactor Trip Occurred.Caused by Water Instrusion Into Pump Loss of Circulating Water Pump.Plant Modification Previously initiated.W/910520 Ltr ML20024H1741991-05-17017 May 1991 LER 90-002-02:on 900216,determined That Fire Dampers May Not Be Operable Under Expected Ventilation Flow Conditions Due to Design Error.Design Basis Documents updated.W/910517 Ltr ML20029B0751991-02-28028 February 1991 LER 91-002-00:on 910129,startup Transformer Incapable of Maintaining Voltage Output Above Setpoint of Second Level Undervoltage Relays Under Es Actuation Conditions.New Offsite Power Supply installed.W/910228 Ltr ML20029A6531991-02-25025 February 1991 LER 91-001-00:on 910124,engineered Safeguards Train a HPI Recirculation Isolation Valve MUV-53 Declared Inoperable Due to Undersized Thermal Overload Elements for Motor Operator.Cause Undetermined.Elements replaced.W/910225 Ltr ML20028H7131991-01-21021 January 1991 LER 89-022-01:on 890614,erroneous Indication of Loss of Main Feedwater Pumps Occurred Resulting in Manual ESF Actuation. Temporary Loss of Supervisory Indicating Lights Expected During Bus Realignment.Procedures revised.W/910121 Ltr ML20043H5751990-06-22022 June 1990 LER 90-009-00:on 900523,discovered That Paperwork for Mod to Replace Feedwater Vent Valve Still Open 12 Months After Work Completed.Caused by Lost Blanket Work Request Re post-maint Testing.New Work Request initiated.W/900622 Ltr ML20043G6511990-06-14014 June 1990 LER 90-008-00:on 900514,determined That Fault Current Could Develop Across Engineered Safeguards 480-volt Bus Output Breakers.Caused by Failure to Perform Adequate Short Circuit Analysis.Deficient Trip Devices replaced.W/900614 Ltr ML20043C5221990-05-31031 May 1990 LER 89-035-01:on 890906,determined That Dc Powered Components Exhibited Discrepancy in Rated Voltages & Actual Voltages.Caused by Inadequate Control of Design Process.Inoperable Components replaced.W/900531 Ltr ML20043B8091990-05-23023 May 1990 LER 90-007-00:on 900112,discovered That Door Between Control Room Complex & Turbine Bldg Removed for Mod Work,Resulting in Inoperability of Both Trains of Emergency Ventilation Sys.Door Replaced & Warning Signs affixed.W/900523 Ltr ML20042G7641990-05-0909 May 1990 LER 90-006-00:on 900410,reevaluation of Design Calculations Discovered Deficient Valve Operator Installation.Sufficient Thrust Not Developed to Open or Close Valves Due to Undersized Spring Packs.Spring Packs ordered.W/900509 Ltr ML20042F9781990-05-0707 May 1990 LER 89-031-01:on 890828,480-volt Engineered Safeguards Stepdown Transformer a Faulted Causing Decay Heat Train Closed Cycle Cooling Pump a to Deenergize.Caused by Degraded Insulation.Transformer replaced.W/900507 Ltr ML20012D8091990-03-22022 March 1990 LER 90-003-00:on 900220,instrument Fluctuation Noted During Surveillance of Chilled Water Pump.Caused by Location of Flow Element within Chilled Water Sys.Relief Request Submitted & Field Problem Rept generated.W/900322 Ltr ML20012D0281990-03-19019 March 1990 LER 90-002-00:on 900216,determined That Fire Dampers May Not Be Operable Under Expected Ventilation Flow Conditions Due to Design Error.Caused by Failure of Original Design Criteria to Address Need.Documentation updated.W/900319 Ltr ML20011F1731990-02-21021 February 1990 LER 90-001-00:on 900221,RCS Leakage Calculations Indicated That Unidentified Leakage Exceeded Tech Spec Limits.Caused by Failed Packing on Block Valve.Valve Repacked During Feb 1990 Maint outage.W/900221 Ltr ML19354D9091990-01-17017 January 1990 LER 89-016-03:on 890426,administrative Problems Caused Deficiencies in Environ Qualification Program That Resulted in Plant Equipment Not Being Properly Qualified.Effort to Correct Environ Qualification Deficiencies Underway ML20006E3941990-01-0808 January 1990 LER 89-041-00:on 891208,partial,simultaneous Withdrawal of Two Control Rod Safety Groups Occurred & Safety Group 3 Control Transferred to Auxiliary Power Supply.Cause Undetermined.Select Relays replaced.W/900207 Ltr ML20005F1601990-01-0808 January 1990 LER 89-040-00:on 891208,emergency Diesel Generators a & B Actuated Due to Degraded Voltage When Condensate Pump Started.Operator Guidelines & Procedures for Starting Condensate Pumps Being revised.W/900108 Ltr ML20011D8401989-12-22022 December 1989 LER 89-030-01:on 890824,determined That Pump Discharge Pressure & Flow Less than Required.On 890826,plant Entered Hot Standby.Caused by Installation of Incorrect Impeller in Pump.Original Impeller Reinstalled Upon repair.W/891222 Ltr ML20011D2281989-12-15015 December 1989 LER 89-026-01:on 890629,emergency Diesel Generator 1B Failed post-maint Test & Could Not Be Returned to Svc within Time Allowed by Tech Specs.Caused by Rotor Not Turning.Crankcase Vapor Ejector cleaned.W/891215 Ltr ML19332E7691989-12-0707 December 1989 LER 89-039-00:on 891107,determined That Control Circuits for Two Makeup Valves Did Not Meet Separation Criteria of 10CFR50 App R.Caused by Cognitive Personnel Error.Roving Fire Watch Patrol Confirmed in effect.W/891207 Ltr ML19332F0171989-11-30030 November 1989 LER 89-038-00:on 891029,util Engineers Discovered That Administrative Controls Not in Place for Three Makeup & Purification Sys Valves.Caused by Personnel Error.Plant Procedures revised.W/891130 Ltr ML19332D1391989-11-27027 November 1989 LER 89-037-00:on 891026,determined That Instrumentation Used to Balance HPI Flow Through Four Injection Lines During Small Break LOCA Inadequate.Caused by Inadequate Review of B&W Guidelines.Flow Instrument installed.W/891127 Ltr ML19327C2611989-11-17017 November 1989 LER 89-036-00:on 891018,determined That Plant Operating Outside Design Basis Since Borated Water Storage Tank Level Transmitters Not Seismically Qualified.Caused by Personnel Error.Unqualified Transmitters replaced.W/891117 Ltr ML19325F3371989-11-10010 November 1989 LER 89-033-00:on 890908,second Level Undervoltage Relay Sys Setpoint for Engineered Safeguards Buses Not Conservative & Led to Operation Outside Plant Design Basis.Caused by Personnel Error.Setpoint changed.W/891110 Ltr ML19325F3341989-11-10010 November 1989 LER 89-035-00:on 890906,discrepancy Noted in dc-powered Component Rated Voltages & Actual Voltages Seen by Components.Caused by Inadequate Control of Design Process. All Components Replaced Prior to startup.W/891110 Ltr ML19325E7371989-10-27027 October 1989 LER 85-034-01:on 850310,valve Alarm Function for Core Flood Tank Isolation Valves Failed to Meet Acceptance Criteria.On 850809,unit Entered Mode 3 & Then Raised RCS Pressure Above 750 Psig W/O Meeting Operability surveillances.W/891027 Ltr ML19324B3921989-10-26026 October 1989 LER 88-002-02:on 880107,emergency Feedwater Actuation Occurred on Loss of Both Main Feedwater Pumps.Caused by Instrumentation & Control Technician Error.Idle Feedwater Pump Started & Actuation reset.W/891026 Ltr ML19324B3871989-10-26026 October 1989 LER 89-034-00:on 890926,two Conditions Determined to Be Outside Plant Design Basis Re Solenoid Control Valves.Caused by Cognitive Personnel Error.Test Solenoid Valve Circuits Provided W/Isolation fuses.W/891026 Ltr ML19327B2301989-10-23023 October 1989 LER 89-016-02:from 890227-0601,deficiencies Re Environ Qualification of Plant Equipment Discovered.Caused by Deficiencies in Detailed Development & Implementation of Environ Qualification Program.Program reviewed.W/891023 Ltr ML20024D2501983-07-26026 July 1983 Updated LER 83-019/03L-1:on 830406,breaker in Engineered Safeguards Motor Control Ctr 3B1 Shorted Out,Causing Various Pieces of Equipment on Train B to Be Inoperable.Caused by Personnel Error.Breaker Removed & cleaned.W/830726 Ltr ML20024D1491983-07-26026 July 1983 Updated LER 82-003/01T-0:on 820128,RCS Leakage Calculations Showed Unidentified Leakage at Rate Greater than 1 Gpm. Caused by Thermally Induced Cyclic Failure.New Valve Will Be Installed in Improved location.W/830726 Ltr ML20024D1061983-07-26026 July 1983 Updated LER 82-004/01T-1:on 820129,while Performing Visual Insp of Reactor Coolant Pump a Seal Package,Rcpb Leakage Discovered from Crack in Seal Weld.Caused by Installation Error.Seal Replaced & Procedures revised.W/830726 Ltr ML20024C0951983-06-27027 June 1983 LER 83-023/01T-0:on 830613,fire Damper FD-86,in Duct Work Between Auxiliary Bldg & Control Complex,Discovered Missing. Cause Unknown.Continuous Fire Watch Established.Evaluation underway.W/830627 Ltr ML20023E0971983-06-0101 June 1983 LER 83-021/03L-0:on 830504,three Hydraulic Snubbers Failed Functional Test During Dec Outage & 107 Snubbers Failed Test During Refuel Iv.Caused by Seal Failure & Valve Assembly Contamination.All Snubbers modified.W/830531 Ltr ML20023C5251983-05-0909 May 1983 LER 83-020/03L-0:on 830409,discovered That Surveillance Interval for Fire Detection Instrumentation in Emergency Diesel Generator & Control Rooms Exceeded by 6 Months.Caused by Procedural Inadequacy.Procedure SP-411 Will Be Revised ML20023B4091983-04-27027 April 1983 LER 83-018/01T-0:on 830413,preliminary Repts Received from B&W Indicating That 51 of 120 Upper Core Barrel Bolts Ultrasonically Tested May Be Defective.Cause Unknown. Investigation Underway.Supplemental Rept Will Be Written ML20028E0561983-01-13013 January 1983 LER 82-075/03L-0:on 821215,during Cold Shutdown,Circuit Breaker on Control Complex Ventilation Radiation Monitor RMA-5 Tripped.Caused by Broken Pump Vane Binding Pump. Pump Replaced ML20028A5321982-11-10010 November 1982 LER 82-063/03L-0:on 821011,reactor Bldg Average Air Temp Exceeded 130 F Limit.Caused by Strain on Instrument Air Line Causing Line to Split.Line Replaced on 821011 ML20027D7021982-10-29029 October 1982 LER 82-061/03L-0:on 820929,fan Damper Operator on Industrial Cooler Failed,Resulting in Reactor Bldg Average Air Temp Exceeding 130 F.Caused by Water in Instrument Air Sys Due to Personnel Failing to Close Valve FSV-250.Dampers Wired Open ML20027B8881982-09-24024 September 1982 LER 82-055/03L-0:on 820825,feedwater Ultrasonic Flow Indicator FW-313-FI Found Inoperable.Caused by Instrument Failure Due to High Ambient Temp at Instrument Cabinet Location.Instrument Repaired.Flow Transmitters Replaced ML20027B9001982-09-24024 September 1982 LER 82-056/03L-0:on 820827,during Normal Operation,Primary Containment Average Air Temp Exceeded 130 F Tech Spec Limit. Caused by Failure of Pneumatic Control Line for Industrial Coolers.Control Line Replaced & Coolers Returned to Svc 1994-05-03
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair 3F1099-19, Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated1999-10-13013 October 1999 Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated ML20217B0931999-10-0606 October 1999 Part 21 Rept Re Damaged Safety Grade Electrical Cabling Found in Supply on 990831.Damage Created During Cabling Process While Combining Three Conductors Just Prior to Closing.Vendor Notified of Reporting of Issue ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212E9031999-09-30030 September 1999 FPC Crystal River Unit 3 Plant Reference Simulator Four Year Simulator Certification Rept Sept 1995-Sept 1999 3F1099-02, Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20211L1321999-08-31031 August 1999 EAL Basis Document 3F0999-02, Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With ML20212C1501999-08-31031 August 1999 Non-proprietary Version of Rev 0 to Crystal River Unit 3 Enhanced Spent Fuel Storage Engineering Input to LAR Number 239 ML20211B7291999-08-16016 August 1999 Rev 2 to Cycle 11 Colr ML20210P1111999-08-0505 August 1999 SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3 ML20210U5341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Crystal River,Unit 3 ML20209F5601999-07-31031 July 1999 EAL Basis Document, for Jul 1999 3F0799-01, Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With ML20210U5411999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Crystal River,Unit 3 3F0699-07, Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With ML20210U5601999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Crystal River,Unit 3 ML20195C6271999-05-28028 May 1999 Non-proprietary Rev 0 to Addendum to Topical Rept BAW-2346P, CR-3 Plant Specific MSLB Leak Rates ML20196L2031999-05-19019 May 1999 Non-proprietary Rev 0 to BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs 3F0599-04, Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With ML20210U5631999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Crystal River,Unit 3 3F0499-04, Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With ML20204D9661999-03-31031 March 1999 Non-proprietary Rev 1,Addendum a to BAW-2342, OTSG Repair Roll Qualification Rept 3F0399-04, Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease1999-03-10010 March 1999 Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease 3F0399-03, Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With ML20203A4381999-02-0303 February 1999 Safety Evaluation Supporting EAL Changes for License DPR-72, Per 10CFR50.47(b)(4) & App E to 10CFR50 ML20206E9891998-12-31031 December 1998 Kissimmee Utility Authority 1998 Annual Rept ML20206E9021998-12-31031 December 1998 Florida Progress Corp 1998 Annual Rept ML20206E9701998-12-31031 December 1998 Ouc 1998 Annual Rept. with Financial Statements from Seminole Electric Cooperative,Inc 3F0199-05, Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With ML20206E9261998-12-31031 December 1998 Gainesville Regional Utilities 1998 Annual Rept 3F1298-13, Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With 3F1198-05, Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With ML20155F4071998-10-31031 October 1998 Rev 2 to Pressure/Temp Limits Rept ML20155J2701998-10-28028 October 1998 Second Ten-Year Insp Interval Closeout Summary Rept 3F1098-06, Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With ML20206E9461998-09-30030 September 1998 Utilities Commission City of New Smyrna Beach,Fl Comprehensive Annual Financial Rept Sept 30,1998 & 1997 ML20206E9561998-09-30030 September 1998 City of Ocala Comprehensive Annual Financial Rept for Yr Ended 980930 ML20206E9101998-09-30030 September 1998 City of Bushnell Fl Comprehensive Annual Financial Rept for Fiscal Yr Ended 980930 ML20206E9811998-09-30030 September 1998 City of Tallahassee,Fl Comprehensive Annual Financial Rept for Yr Ended 980930 ML20195E3121998-09-30030 September 1998 Comprehensive Annual Financial Rept for City of Leesburg,Fl Fiscal Yr Ended 980930 3F0998-07, Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With ML20236W6501998-07-31031 July 1998 Emergency Action Level Basis Document 3F0898-02, Monthly Operating Rept for Jul 1998 for Crystal River,Unit 11998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Crystal River,Unit 1 ML20236V8801998-07-30030 July 1998 Control Room Habitability Rept 3F0798-01, Monthly Operating Rept for June 1998 for Crystal River Unit 31998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Crystal River Unit 3 ML20236Q4611998-06-30030 June 1998 SER for Crystal River Power Station,Unit 3,individual Plant Exam (Ipe).Concludes That Plant IPE Complete Re Info Requested by GL 88-20 & IPE Results Reasonable Given Plant Design,Operation & History 3F0698-02, Monthly Operating Rept for May 1998 for Crystal River Unit 31998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Crystal River Unit 3 1999-09-30
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f (3T71 lDuring routine power operation, a nnwer sunniv c,41 rn 4.,4-<,-,,,,
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, engineered safeguards actuation. A review of the transient revealed events rennr d-able under Technical specifications 6 9 1.9.b, 6.9.1.8, and 10 CFR 20.403. The 1 i
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,,M of preparer: . K. F. Lancaster ,. (9041 795-4486 i (SEE AT"Af"W5'n SUPP7Nm EGOICLCION SM) i l
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SUPPLEMENTARY INFORMATION Rsport No.: 50-302/80-010/01T-1 Facility: Crystal River Unit #3 Rsport Date: 20 March 1980 J
Occurrence Date: 26 February 1980 Conditions Prior to Occurrence:
Flux 98.6 OTSG "A" FRLV 242 inches RC pressure 2157 psig OTSG "B" FRLV 254 inches PZR level 202 inches OTSG "A" pressure 911 psig MU cank level 71 inches OTSG "B" pressure 909 psig Th "A" 599'F Main steam pressure 894 psig j T "B" 600*F Main steam temp. 589*F h
T "A" 557'F Condenser vacuum 1.76 T "B" 556*F Generated FSi 834 Reflow"A"73x106 lbs/hr DFT level 12.7 ft.
RC Flow "B" 73 x 10" lbs/hr Feedflow"A"5x10flbs/hr Letdown flow 48 gpm Feed flow "B" 5 x 10 lbs/hr OTSG "A" level (0P) 67% Feed pressure "A" 970 psig OTSG "B" level (OP) 65% Feed pressure "B" 968 psig i
Description of Occurrence:
At 1423 during routine power operation, the plant suffered a loss of the 24 VDC non-nuclear instrumentation "X" power supply. This resulted in the loss of the "X" power supply instrumentation and initiated a plant transient condition ending in a reactor trip, turbine trip, and high pressure injection. The plant was stabilized and maintained in Mode 3 (hot standby) on natural circulation until forced Reactor Coolant System flow was initiated at 2107. The plant was subsequently taken to Mode 5 (cold shutdown) at 0755 on 29 February 1980.
A review of the plant transient revealed events contrary to the following Technical Specifications.
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IDENTIFICATION / DESCRIPTION OF OCCURRENCE 3.1.2.9 The borated water storage tank shall be operable with a contained borated water volume of between 415,200 and 449,000 gallons.
During the plant transient and as a result of the high pressure injection actuation at 1426 on 26 February 1980, the borated water storage tank volume decreased from approximately 436,914 gallons to approximately 380,000 gallons.
High pressure injection was terminated at 1550 on 26 February 1980.
I Relief for these Technical Specifications was granted as of 1715 on 26 February 1980 until the plant entered into Mode 5 (cold shutdown).
3.4.5 Each steam generator shall be operable with a water level between 18 and 360 inches.
Review of the plant conditions during the transient revealed that a reduction in feedwater caused by the failure of the "X" power supply resulted in steam generator "A" operation with a level less than 18 inches. It was determined that "A" OTSG effectively boiled dry approximately four (4) minutes into the transient at 1427 on 26 February 1980. Level was restored to greater than 18" at approxi-mately 1612 on 26 February 1980 using the auxiliary feedwater header via the bypass valves. -
3.4.6.1 The following Reactor Coolant System leakage detection systems shall be operable:
a) The containment atmosphere iodine radioactivity monitoring system.
b) The containment sump level monitoring system, and, c) the containment atmosphere gaseous radioactivity monitoring system.
3.4.6.2.d Reactor Coolant System leakage shall be limited to 10 gpm identified leakage.
The combined effort of the PORV opening on the NNI(X) 24 VDC power failure and safety valve (RCV-8) relieving to the Reactor Coolant drain tank following high pressure injection. actuation resulted in the rupture disc being blown and leakage exceeding the requirements of TS 3.6.4.2.d. Leakage rates are assumed to be consistent with the estimated HPI flow rates illustrated in Figure III-20 (at-tached). Leakage was maintained within Technical Specification limitations upon termination of HPI at 1550 on 26 February 1980.
The containment sump level monitoring system was rendered inoperable at approxi-mately 1430 on 26 February 1980 when sump level elevated off-scale high follow-ing high pressure injection actuation. The containment atmosphere radiation monitoring system was rendered inoperable upon isolation of the monitor sample line at approximately 1426 on 26 February 1980.
~
l Containment atmosphere samples were taken in accordance with post-accident sampling and analysis of Reactor Building Atmosphere Procedure EM-303. The f
containment atmosphere monitoring system was returned to service at 2115 on g 26 February 1980. The containment sump level was returned to normal on 5 March 1980. The total volume accumulation in the sump as a result of the incident was approximately 42,810 gallons.
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Idsntification/ Description 2 J.P. O'Reilly, Director of Occurrence LER 80-10-OlT-1 Relief for surveillance requirements 4.4.6.2.b and 4.4.6.2.d of Techn. cal Specification 3.4.6.2 was granted as of 0915 on 27 February 1980 to continue until the Reactor Building sump was returned to normal.
3.6.1.4 Primary containment internal pressure shall be maintained between 17.7 and 12.7 psia.
The loss of Reactor Coolant via the pressurizer relief system through the Reactor Coolant drain tank resulted in the Reactor Building pressure increasing to approximately 19 psia. RB pressure was reduced to less than the Technical Specification limit of 17.7 psia at 1550 on 26 February 1980 and was stabilized at 15 psia at 1645 on 26 February 1980.
3.6.1.5 Primary containment internal air temperature shall not exceed 130*F.
The average containment temperature exceeded 130*F and reached a maximum average temperature of 142*F during the transient. It was determined that the
{ temperature transient remained above the Technical Specification limit for approximately one (1) hour.
- 3.4.9.2 The pressurizer temperature shall be limited to a maximum heatup and cooldown of 100*F in any one (1) hour.
Prior to the transient, the pressurizer temperature was 640*F. Approximately three (3) minutes into the transient following high pressure injection
. actuation, the RCS was " solid" and was relieving through the pressurizer code safety (RCV-8). At this time the pressurizer temperature was approximately equal to incore temperature. At 1523 on 26 February 1980 one (1) hour into the transient, incore temperature was 470*F. This temperature differential corresponds to a pressurizer cooldown rate of approximately 170*F in one hour.
3.7.3.2 Two independent decay heat closed cycle cooling water loops shall be operable.
At 1930 on 26 February 1980, decay heat closed cycle cooling pump DCP-1A failed,
, rendering "A" DHCCC loop inoperable. Consequently, the components being supplied i by DCP-1A were also rendered inoperable. These include DHP-1A, DHHE-1A, MUP-1A, BSP-1A, and RWP-3A. These components are identified below in the affected Technical Specifications.
3.5.2.a,b,c. Two independent ECCS subsystems shall be operable with each subsystem comprised on one operable high pressure injection pump (MUP-1A), one operable low pressure injection pump (DHP-1A) and one operable decay heat cooler (DHHE-1A).
3.7.4.2 Two independent decay heat seawater loops shall be operable. (RWP-3A).
3.6.2.1 Two independent containment spray systems shall be operable with each spray system capable of taking suction from the BWST on a containment spray actuation signal and manually transferring suction to the containment sump. (BSP-1A).
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4 Idsntification/ Description 3 J.P.0'Reilly, Director of Occurrence LER 80-10/0lT-1 As a result of the failure of DCP-1A, a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement was entered.
Subsequent investigation revealed that the pump-motor coupling on DCP-1A had failed.
The coupling was replaced with a spare and DCP-1A was returned to service at 0948 on 28 February 1980 following satisfactory completion af ECCS Pump Operability Surveillance SP-340 Relief for Technical Specifications 3.5.2, 3.7.3.2, and 3.7.4.2 was granted as of 0915 on 27 February 1980 to continue until the plant entered Mode 5 (cold shutdown).
3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 of the Technical Specifications shall be operable with readouts displayed external to the Control Room.
A review of the transient revealed that the following remote shutdown instruments were inoperable as a result of the non-nuclear instrumentation "X" power supply failure at 1423.
MU-L4-LI Makeup tank level RC-1-LI-l Pressurizer level RC-4A-TI-2 RCS loop "A" Th RC-4B-TI-2 RCS loop "B" T h SP-1A-LI-3 OTSG "A"SU level SP-6A-PI-l OTSG "A" pressure DH-39-TI Decay heat cooler "A" outlet temp.
Power to these instruments was returned upon restoration of the NNI "X" power supply at 1444, on 26 February 1980.
3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 of the Technical Specifications shall be operable with readouts and recorders in the Control Room.
A review of the transient revealed that the following post-accident monitoring instruments were inoperable as a result of the non-nuclear instrumentation "X" power supply failure at 1423, on 26 February 1980.
Pressuri'.er level LT-1 Pressurizer level LT-2 Pressurizer level LT-3 RC-4A-TI RCS 'aop "A" Th
- RC-4B-TI RCS loop "B" T h
RC-13-FR RCS flow SP-6A-PR OTSG "A" outlet pressure SP-7A-FI FW flow "A" l SP-7B-FI FW flow "o" I
- Power to these instruments was returned upon restoration of the NNI "X" power supply at 1444 on 26 February 1980, 3.4.1 Both Reactor Coolant loops and both Reactor Coolant pumps shall be in operation.
Contrary to this specification, operation in Mode 3 (hot standby) proceeded without at least one Reactor Coelant pump or decay heat remotal pump.
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Id:ntification/ Description 4 J.P.0'Reilly of Occurrence LER 80-10-OlT-1 During the transient, following the HPI actuation from low RCS pressure, all Reactor Coolant pumps were tripped in accordance with Loss of Reactor Coolant or Reactor Coolant Pressure, Emergency Procedure, EP-106, and B&W small break guidelines. The plant was maintained on natural circulation until 2107 on 26 February 1980 when forced RCS flow was inititted.
3.4.3 A cinimum of one pressurizer code safety valve shall be operable with a lift setting of 2500 psig 1%.
Contrary to this specification, pressurizer code safety RCV-8 lifted at 2400 psig.
Reactor Coolant System pressure control was established at 1453 on 26 February 1980 using normal makeup and letdown.
3.1.1.2 The flowrate of Reactor Coolant through the Reactor Coolant System shall be 2 2700 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.
At 1453 on 26 February 1980 while re-establishing normal letdown and makeup flow to maintain Reactor Coolant system pressure control (on natural circulation) an apparent reduction in RCS boron concentration was equal to, or greater than the RCS boron following high pressure injection.
3.4.8 The specific activity of the primary coolant shall be limited to less than or equal to 1.0 microcuries per gram Dose Equivalent I-131.
At 0058 on 27 February 1980 it was discovered that the Reactor Coolant Dose Equivalent Iodine-131 was 1.45 microcuries per gra.s. The four hour sampling frequency was initited per the Action Statement re uirements. The sampling interval was terminated at 0110 on 28 February 198( -hen the DEI-131 was determined to be .755 microcuries per gram.
Reference the following Supplemental Information Data in regard to this event.
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l Analysis of the Occurrence:
The loss of electric power is an analyzed design basis event for the CR-3 FSAR. The following criteria for reactor protection are required for this event:
a) Fuel damage will not occur from an excessive power-to-flow rate.
b) Reactor Coolant system pressure will not exceed code pressure limits.
c) The resultant doses are within 10 CFR 100 limits.
The February 26, 1980 transient involved a partial loss of electrical power.
The applicable protection criteria appear to have been met throughout the event.
The sequence of events following the power supply failure in the NNI "X" cabinet is similar in nature to a loss of feedwater type transient whose consequences have been bounded for the FSAR analysis by the design basis event of a feedwater line break. The Reactor protection criteria for the feedwater line 13eak are:
a) The core shall remain intact for effective core cooling.
b) The Reactor Coolant System pressure shall not exceed code pressure limits of 110% of 2500 psig; i.e., 2750 psig.
The major difference between the February 26, 1980, event and the bounding feedwater line break event analyzed for the FSAR are itemized below:
- 1. The main feedwater valves were closed over a 25-30 second time interval rather than the conservative FSAR assumption of an instantaneous rupture of the main feedwater header at the steam generator inlet nozzles.
- 2. The combination of the opening of the PORV, the ICS demand for 103% full power and feedwater runbacks caused the Reactor trip on high pressure to occur at 10-25 seconds into the event rather than at 11.8 seconds for the FSAR event.
- 3. The opened PORV allowed a rapid depressurization to 1500 psi such that the ESFAS was actuated and two HPI pumps were started. The PORV and spray block valves were closed about five (5) minutes into the event and the RCS code safety valves actuated about 13 minutes into the event. This was much slower than 22.3 seconds for the FSAR events.
- 4. The operator initiated emergency feedwater at approximately nine (9) minutes into the event rather than the 15 minutes which was demonstrated for the FSAR as an acceptable time for operator action to prevent core or Reactor Coolant boundary damage.
- 5. Main feedwater was supplied "B" 0TSG at about four (4) minutes (approximately 1400 gpm) rather then being totally lost as analyzed in the FSAR.
- 6. The maintenance of full HPI flow for 32 minutes provided a significantly greater heat removal capacity than assumed in the FSAR where only relief through the safety valves of the expanding RCS system water is considered.
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6 Analysis of the Occurrence: (page 2)
For both events, core coverage was maintained, o fuel damage occurred and the Reactor Coolant System pressure remained within code allowable limits.
The safety evaluation criteria were met.
Although apparently initiated by a single fault, this transient involved a series of multiple, propagating abnormalities through the interaction of the
., various systems which are affected directly or indirectly by the interruption of the NNI "X" power source. These interactions led to:
a) Loss of some instrumentation and annunciator indications to the plant operator. Presentation of some confusing and invalid information b) Presentation of some invalid plant signals to the integrated control system, which acted upon them to increase Reactor power, decrease feedwater flow and increase steam flow.
c) The pressurizer PORV was opened and latched at a system pressure below its setpoint.
d) Automatic initiation of emergency feedwater did not occur, despite reduction of main feedwater.
The maintenance of plant conditions within the safety criteria required the action of automatic safeguards (Reactor trip, high pressure injection, Reactor Building isolation), safeguards features (safety relief valve), and/or operator action (start emergency feedwater) in response to items (b)(c), and (d). Item (a) reduced the operator's ability to respond.
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7 Corrective Action:
Corrective action as a result of this event was identified in our letter of 12 March 1980 to Mr. Harold Denton.
Failure Data: This is the first occurrence of this type report.
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Enclosure 1 (pg. 1)
REACTOR PO'ER HISTORY OF PRIOR FORTY-EIGHT HOURS ITEM 1 EVENT DATE: 27 FEBRUARY 1980 l
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S HOUR GMWE TURB G MWTH NI RATIO RATIO
$ ( E 711 ) (-TG56 ) sP75-3) -tF 7033 i41/h T- ME/hT
- /*FP BTU /XWH +/=FP */*FP */=FP =/*FP
.- 2 93.60 1.052 .982
- 1. 87.31 9943 88.95
,, ; 27.o. 77Ao 2877G 93750 1. 754 .755 f 3" 3 87.33 9916 88.66 93.60 1.056 .985 2 4 87.35 9927 88.70 93.70 1.056 .985 0 &T737 7733 38.6o 7730 1.055 .965 l
e 6 87.37 9933 88.66 S3.80 1.058 .9.85
,y 7 S9.59 10416 95.68 99.20 1.037 .936 5 93.-39 7t31 75-70 99-SG 1.C37 .9?S
'W 1.023 .980 97.80
- H 9 93.67 9973 95.60 5 10 97.35 9988 99.55 98.40 .988 .978
??32---99 .-67 79:t0 .794 .979
$ il 77.33 '
99.00 .993 .978 3 12 97.51 9979 99.67 13 97.52 9982 99.67 98.90 .992 .978
% 99.67 7?790 . 79; . 7M3 i e 14 F7.44 F9di E 15 47.12 9989 64.44 55.80 .866 .731 16 0.00 0 .00 0.00 0.000 0.000
% u .uu u.us u .7 0 0 0.unu g 1/ u.uu
. E 18 0.00 0 .00 0.00 0.000 0.000
.00 0.00 0.000 0.000 3
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. 4 21 0.00 0 .00 0.00 0.000 0.000
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(( AVERAGE DAILY GENERATOR GROSS 475.74 MWH(E) --
g s v .. ,n o c. w ,- A i a THERfA FC::EK . 411. 96--MitrT )
s AVERAGE DAILY TURIINE GRC55 HEAT RATE 6244 BTU /KWH y AVERAGE DAILY MUTH POWER 57.584 =/*FP v: n.casuc unAL: svuumEnA lasi. FUWed 55.663 / FF II E
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Enclosure 1 (pg 4)
FUEL BURNUP BY CORE REGION ITEM 2 EVENT DATE: February 27, 1980 I
Enclosure 1 (pg 5)
ITEM 2 The burnup was calculated at 166.494 EFPD for the three (3) enrichment regions.
REGION NUMBER OF FA BURNUP B 61 20,046 MWD /MTU C 60 15,748 MWD /MTU D 56 4,434 MWD /MTU Avg. 13,650 MWD /MTU l
l l
Enclosure 1 (pg 6)
CLEANUP FLOW HISTORY ITEM 3 EVENT DATE: February 27, 1980
Enclosure 1 (pg 7)
ITEM 3 Cleanup flow history forty-eight (48) hours prior to the first sample in which the limit was exceeded indicated a letdown flowrate of forty-five (45) gpm.
DATE/ TIME F1.0WRATE 2/25/80 @ 1420 47 gpm 2/26/80 @ 0010 100 gpm 2/26/80 0 1415 47 gpm 2/26/80 @ 2135 54 gpm 2/27/80 @ 00.18 60 gpm
Enclocurs 1 (pg 8)
DEGAS OPERATIONS ITEM 4 EVENT DATE: February 27, 1980 NO DEGAS OPERATIONS.
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l 4
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Enclosure 1 (pg 9) l l
l l
TIME DURATION WlEN DEI-131 EXCEEDED 1.0 uCi/ gram and I-131 ANALYSIS RESULTS ITEM 5 Event Date: February 27, 1980
Enclosuro 1 (pg 10) 4 l
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ITEM 5 -
l As per Technical Specification 3.4.8, the four (4) hour sampling frequency as depicted on the Table below was initiated at 0058 on 27 February 1980, and the Dose Equivalent I-131 was 1.45 microcuries per gram. The four hour sampling frequency was terminated at 0110 on 28 February 1980 when the DEI-131 was determined to be .755 micro-curies per gram. The time duration when the specific activity of primary coolant exceeded 1.0 microcuries per gram Dose Equivalent is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 12 minutes.
q DATE TIME DEI-131 ( Ci/ gram) ;
2/27/80 0058 1.45 2/27/80 0615 1.1 2/27/80 0855 .979 2/27/80 1255 .944 2/27/80 1755 1.07 2/27/80 2110 .818 2/28/80- 0110 .755 2/28/80 0815 .620 2/28/80 2020 .526 4
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