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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20046C2121993-07-30030 July 1993 LER 93-004-01:on 930301,confirmed That Channel D Axial Shape Index (Asi) Being Calculated in Reverse Since 921031-930301 Due to Drawing Discrepancies Associated W/Control Channel B. Temporary Mod 92-078 & Standing Order 0-25 Revised ML20046A8691993-07-26026 July 1993 LER 93-011-00:on 930624,experienced Reactor Trip Due to Loss of Load.Caused by Lack of Proper Job Planning,Lack of Formal Decision Making Process & Incomplete Communications.Training Will Be Provided to Operations personnel.W/930726 Ltr ML20045H2561993-07-12012 July 1993 LER 93-010-00:on 930611,1 of 14 Halon Cylinders Did Not Meet Min Pressure Acceptance Criteria Listed in Semiannual Switchgear Rooms Surveillance Test.Caused by Failure of Test to Include Necessary Steps.Cylinder recharged.W/930712 Ltr ML20045D7201993-06-22022 June 1993 LER 93-009-00:on 930524,apparent Spurious Signal from Pressurizer Level Instrumentation Caused Backup Charging Pumps to Automatically Start,Due to Deterioration of Wiring. Instrument Loop Calibration Will Be performed.W/930622 Ltr ML20045D3741993-06-21021 June 1993 LER 93-008-00:on 930520,determined That TS SR Not Satisfied for Stack Flow Indicator,Per Amend 137 Issued on 910307. Caused by Lack of Attention to Detail.Calibr & Functional Test Procedures developed.W/930621 Ltr ML20044H5261993-06-0101 June 1993 LER 93-007-00:on 930430,unplanned Emergency Generator Start & Rt Signal Occurred.Caused by Inadequate Attention to Detail,Labeling of Fuse Drawers,Caution Signs & Training. Labeling & Caution Signs upgraded.W/930601 Ltr ML20044G4941993-05-26026 May 1993 LER 93-006-00:on 930118,Halon Fire Suppression Sys for Switchgear Rooms Disabled to Allow Repair/Replacement of Halon Sys Piping.On 930427,individual Responsible for Fire Watch Not Present.Individual Relieved of Responsibilities ML20044B6711993-02-22022 February 1993 LER 93-002-00:on 930122,determined That Current SG LP Signal Block Reset Values Greater than Allowed Ts.Caused by Improper Design.Test Procedures Will Be Revised by 930917 to Specify Desired Value for Block function.W/930222 Ltr ML20024G6821991-04-19019 April 1991 LER 91-007-00:on 910320,480 Volt Circuit Breaker Coordination Outside Design Basis.Caused by Deficiencies in Original Sys Design.Breaker/Fuse Coordination Study to Be Completed & Problems Will Be corrected.W/910419 Ltr ML20029C1591991-03-21021 March 1991 LER 91-004-00:on 910212,offsite Power Low Signal Outside Design Basis.Caused by Inadequate Mod Design at Time of Performance of Original Degraded Voltage Analysis. Engineering Analysis EA-FC-91-017 performed.W/910321 Ltr ML20029C1051991-03-18018 March 1991 LER 91-002-00:on 901209,ventilation Isolation Actuation Signal Generated by High Alarm on Process Radiation Monitor RM-062.Caused by Accumulation of Noncondensible Gases in Sample Piping.Valve Packing Leak repaired.W/910318 Ltr ML20029A2981991-02-0808 February 1991 LER 91-001-00:on 910109,determined That Containment Tendon Surveillances Performed in 1981 & 1985 Did Not Reflect Guidance in Tech Specs.Caused by Inadequate Administrative Controls.Testing Program Plan implemented.W/910208 Ltr ML20029A2971991-02-0606 February 1991 LER 90-022-02:on 900907,approx 460 Fire Barrier Penetration seals,60 Fire Dampers & 6 Fire Doors Declared Nonfunctional Per NRC Info Notice 88-004 Due to Lack of Documentation. Plant Outage Required to Implement Repairs/Replacements ML20028G9171990-09-28028 September 1990 LER 90-021-00:on 900829,inadvertent Reactor Protective Sys Actuation Occurred While Operator Changed Power Source. Caused by Operator Not Following Proper Procedures.Operator counseled.W/900928 Ltr ML20044B0131990-07-12012 July 1990 LER 90-018-00:on 900612,reactor Protective Sys (RPS) Trip Units for Axial Power Distribution Determined to Be Inoperable.Caused by Procedural Deficiencies.Procedure Revised & RPS Surveillance Tests reviewed.W/900712 Ltr ML20043F6301990-06-11011 June 1990 LER 90-016-00:on 900511,accident Scenarios Identified by Which Auxiliary Feedwater Piping from Discharge of Turbine Driven Auxiliary Feedwater Pump FW-10 Can Be Overpressurized.Caused by Design deficiency.W/900611 Ltr ML20043F2441990-06-0707 June 1990 LER 90-015-00:on 900507,PORV Variable Setpoints Used for Low Pressure Overpressure Protection Determined to Be Nonconservative for PORV Opening Time.Caused by Design Deficiency.Tech Spec Amend prepared.W/900607 Ltr ML20043C0991990-05-29029 May 1990 LER 90-014-00:on 900427,investigation Revealed That Component Cooling Water Piping to Reactor Coolant Pump Seal Coolers Could Be Targets of High Energy Line Break.Safety Analysis for Operability completed.W/900529 Ltr ML20042G7211990-05-10010 May 1990 LER 90-011-00:on 900402,inadvertent Actuation of Pressurizer Pressure Low Signal Occurred While Performing Calibr Procedure.Caused by Inappropriate Action by Technician Involved.Validation of Procedures reviewed.W/900510 Ltr ML20042E6871990-04-23023 April 1990 LER 90-007-01:on 900228,determined That Several Supports Would Be Overloaded During Seismic Event on Nonsafety Related & safety-related Main Steam Piping.Caused by Design Deficiency.Piping Supports modified.W/900423 Ltr ML20042E6861990-04-23023 April 1990 LER 90-009-00:on 900316,potential Overpressurization of Auxiliary Feedwater Piping Could Have Occurred During Thermal Expansion of Process Fluid Between Closed Valved. Caused by Design deficiencies.W/900423 Ltr ML20012E7641990-03-26026 March 1990 LER 90-005-00:on 900223,determined That Spent Fuel Pool Area Charcoal Filtration Unit VA-66 Was Outside Design Basis. Caused by Insufficient Airflow Into Unit.Affected Updated SAR Analysis Will Be updated.W/900326 Ltr ML20012D0121990-03-19019 March 1990 LER 90-004-00:on 900217,lift Pressures for 6 of 10 Main Steam Safety Valves Found Outside Acceptance Criteria. Caused by Overly Restrictive Operability Criteria.Valves Recalibr & License Amend Submitted to NRC.W/900319 Ltr ML20012D0101990-03-19019 March 1990 LER 90-003-00:on 900216,determined That Auxiliary Feedwater Piping Outside Normal Stress Limits of ASME Code & Design Basis Specified in Updated Sar.Caused by Design Deficiency.Valve Operators Will Be inspected.W/900319 Ltr ML20012B6361990-03-0909 March 1990 LER 89-017-01:on 890624,internal Valve Component from Check Valve Found Lying on Pump Discharge Vane.Repair or Replacement of Valve Internals Could Not Be Accomplished within Time Requirement of Tech Spec.W/900309 Ltr ML20006E1041990-02-0909 February 1990 LER 90-001-00:on 900108,fire Barrier for Wall Between Auxiliary Bldg Rooms 26 & 34 Breached But Hourly Fire Watch Patrol Not Established.Caused by Lack of Sufficient Training for Shift Supervisors.Standing Order revised.W/900209 Ltr ML20011E2691990-02-0505 February 1990 LER 89-024-00:on 891221,determined That Containment Spray Pumps & Suction Header Piping Not Constructed for Use as Backup to LPSI Sys for Shutdown Cooling.Caused by Inadequate Review of Assumptions.Firewatch established.W/900205 Ltr ML20011E2271990-02-0101 February 1990 LER 89-021-00:on 891010,util Informed by C-E of Potential Nonconservative Setpoint in Reactor Protection Sys Thermal Margin/Low Pressure Trip Unit.Caused by Error in Incorporating Transient Setpoint analyses.W/900201 Ltr ML20005F7151990-01-10010 January 1990 LER 89-023-00:on 891211,hourly Firewatch Patrol Entered Posted High Radiation Area W/O Meeting Entry Requirements for Area.Briefings on High Radiation Entry Requirements Held for Personnel W/Assigned dosimetry.W/900110 Ltr ML19354D6381989-12-20020 December 1989 LER 89-022-00:on 890805,change to Surveillance Procedure ST-CEA-1 Became Effective Which Would Have Made Both Emergency Diesel Generators Simultaneously Inoperable During Portion of Test.Change removed.W/891220 Ltr ML19332E7431989-12-0808 December 1989 LER 88-037-01:on 881214,one of Two Supply Headers Supplying Fire Suppression Headers in Auxiliary Bldg Isolated.Caused by Lack of Procedural Guidance & Inadequate Procedural Controls.Standing Order G-58 Will Be revised.W/891208 Ltr ML19332E2681989-12-0101 December 1989 LER 89-016-02:on 890616,for Unknown Period Since 890614, Auxiliary Feedwater Pump FW-10 Operated Outside Design Basis for Certain Accident Conditions.Caused by Inoperable Speed Control Loop.Action Plan implemented.W/891201 Ltr ML19351A4541989-11-22022 November 1989 LER 89-020-00:on 891012,determined That Two of Four Component Cooling Water HXs Simultaneously Inoperable for More than 24 H.Caused by Inadequate Controls Re Return of Equipment to Svc.Standing Order revised.W/891122 Ltr ML19327B5481989-10-24024 October 1989 LER 89-019-00:on 890924,indication of High Temp for Reactor Coolant Pump RC-3A Upper Motor Thrust Bearing Received in Control Room.Caused by Damaged Cable for Bearing Resistive Temp Device.Damaged Cable replaced.W/891024 Ltr ML19325D2471989-10-13013 October 1989 LER 89-012-01:on 890502,main Feedwater Isolation Valve to Steam Generator a Found Inoperable Due to Improperly Set Torque Switch.Caused by Inadequate Program for Maint of Motor Operated Valves.Torque Switches reset.W/891013 Ltr ML20028C7711983-01-0606 January 1983 LER 82-020/03L-0:on 821207,during Main Steam Safety Valve Test,Four Main Steam Safety Valves Had Lift Setpoints Out of Tolerance.Caused by Normal Drift of Valves Over Operating Cycle.Valves Readjusted ML20028B5451982-10-28028 October 1982 LER 82-019/03L-0:on 821024,MSIVs HCV-1041A & HCV-1042A Stopped Three to Four Degrees Off Seat When Signaled to Close.Caused by Binding Between Valve Packing & Shaft. Packings Sprayed W/Penetrant Oil ML20052J0631982-04-27027 April 1982 LER 82-009/03L-0:on 820411,while Exchanging Component Cooling Water Heat Exchangers,Associated Outlet Valves HCV-490B,HCV-491B & HCV-492B Failed to Open.Cause Not stated.HCV-491B Reassembled & Tested ML20052B2361982-04-0707 April 1982 LER 82-006/03L-0:on 820323,during Surveillance Test ST-ISI- WD-1,F.1,valve HCV-506A Failed to Close Via Control Room Switch.Caused by Solenoid Valve Malfunction.Solenoid Valve Disassembled,Cleaned & Reassembled ML20052D9291982-04-0606 April 1982 LER 82-008/03L-0:on 820330,during Performance of ST-FW-1, F.2(b)(6)per Tech Spec 3.9,steam Driven Auxiliary Feedwater Pump Failed to Start.Caused by Back Pressure Trip Lever in Tripped Position.Lever Reset ML20041G1291982-02-22022 February 1982 LER 82-005/03L-0:on 820210,at 98% Power,Control Element 24 Inserted Into Core.Emergency Procedure EP-13,CEDM Malfunctions,Implemented & Power Stabilized at 88%.Caused by Erroneous Operating Instruction.Instruction Changed ML20041F7481982-02-17017 February 1982 LER 82-003/03L-0:on 820203,containment Isolation Valve Associated W/Gas Vent Header HCV-507A Failed to Close on Demand.Caused by Solenoid Valve Plunger Sticking in Energized Position.Plunger Freed ML20041F6251982-02-0505 February 1982 LER 82-004/03L-0:on 820203,small Quantity of Radioactive Gas/Particulate Released to Auxiliary Bldg During Routine Operation.Caused by Failure of Stack Gas Monitor RM-062 to Alarm at Appropriate Setpoint Due to Faulty Alarm Module ML20041B1051982-01-28028 January 1982 LER 82-002/03L-0:on 820114,at 99% Power,Lockout Relay 86B1, Containment Radiation High Signal,Failed to Actuate on Demand by Plant Radiation Monitoring Sys.Caused by Burnt Coil on Lockout Relay.Coil Replaced & Tested Satisfactorily ML20041B1171982-01-19019 January 1982 LER 82-001/03L-0:on 820111,during Normal Operation,Two Fire Barrier Penetrations Found Nonfunctional.Shift Supervisor Immediately Notified;However,Fire Watch Not Posted.Insp & Supervisor Personnel Instructed on Proper Actions ML20039B4561981-12-11011 December 1981 LER 81-011/03L-0:on 811113,containment Isolation Valves Opened & Ventilation Process Initiated W/Containment Air Monitor RM-050/051 Inoperable.Caused by Personnel Error. Valves Closed ML20010H8581981-08-27027 August 1981 LER 81-008/03L-0:on 810813,86B/CRHS (Containment Radiation High Signal) Lockout Relay Failed to Actuate When RM-062 Was Placed in Alarm,Resulting in Failure of 86B1/CRHS Relay to Actuate.Caused by Dirt in Relay Latching Mechanism ML20041F6291981-08-27027 August 1981 LER 81-008/03L-1:on 810813,containment Radiation High Signal 86B Lockout Relay Failed to Actuate When Radiation Monitor RM-062 Placed in Alarm.Caused by Bound Relay Latching Mechanism Due to Dirt & Grease.Latch Cleaned ML20010C2271981-07-0707 July 1981 LER 81-006/03L-0:on 810624,reactor Protection Sys Nuclear Power Recorder Channel B Trip Setpoints Determined to Be Nonconservative.Caused by Faulty Temp Change Power Calculation Due to Grounded Hot Leg Temp Loop ML20004B1111981-05-0606 May 1981 LER 81-005/03L-0:on 810423,dc Sequencer Timers AC-3A (Component Cooling Water Pump) & AC-102A (Raw Water Pump) Failed to Time Out within Prescribed Limit.Cause Unknown Mechanisms Satisfactorily Inspected 1993-07-30
[Table view] Category:RO)
MONTHYEARML20046C2121993-07-30030 July 1993 LER 93-004-01:on 930301,confirmed That Channel D Axial Shape Index (Asi) Being Calculated in Reverse Since 921031-930301 Due to Drawing Discrepancies Associated W/Control Channel B. Temporary Mod 92-078 & Standing Order 0-25 Revised ML20046A8691993-07-26026 July 1993 LER 93-011-00:on 930624,experienced Reactor Trip Due to Loss of Load.Caused by Lack of Proper Job Planning,Lack of Formal Decision Making Process & Incomplete Communications.Training Will Be Provided to Operations personnel.W/930726 Ltr ML20045H2561993-07-12012 July 1993 LER 93-010-00:on 930611,1 of 14 Halon Cylinders Did Not Meet Min Pressure Acceptance Criteria Listed in Semiannual Switchgear Rooms Surveillance Test.Caused by Failure of Test to Include Necessary Steps.Cylinder recharged.W/930712 Ltr ML20045D7201993-06-22022 June 1993 LER 93-009-00:on 930524,apparent Spurious Signal from Pressurizer Level Instrumentation Caused Backup Charging Pumps to Automatically Start,Due to Deterioration of Wiring. Instrument Loop Calibration Will Be performed.W/930622 Ltr ML20045D3741993-06-21021 June 1993 LER 93-008-00:on 930520,determined That TS SR Not Satisfied for Stack Flow Indicator,Per Amend 137 Issued on 910307. Caused by Lack of Attention to Detail.Calibr & Functional Test Procedures developed.W/930621 Ltr ML20044H5261993-06-0101 June 1993 LER 93-007-00:on 930430,unplanned Emergency Generator Start & Rt Signal Occurred.Caused by Inadequate Attention to Detail,Labeling of Fuse Drawers,Caution Signs & Training. Labeling & Caution Signs upgraded.W/930601 Ltr ML20044G4941993-05-26026 May 1993 LER 93-006-00:on 930118,Halon Fire Suppression Sys for Switchgear Rooms Disabled to Allow Repair/Replacement of Halon Sys Piping.On 930427,individual Responsible for Fire Watch Not Present.Individual Relieved of Responsibilities ML20044B6711993-02-22022 February 1993 LER 93-002-00:on 930122,determined That Current SG LP Signal Block Reset Values Greater than Allowed Ts.Caused by Improper Design.Test Procedures Will Be Revised by 930917 to Specify Desired Value for Block function.W/930222 Ltr ML20024G6821991-04-19019 April 1991 LER 91-007-00:on 910320,480 Volt Circuit Breaker Coordination Outside Design Basis.Caused by Deficiencies in Original Sys Design.Breaker/Fuse Coordination Study to Be Completed & Problems Will Be corrected.W/910419 Ltr ML20029C1591991-03-21021 March 1991 LER 91-004-00:on 910212,offsite Power Low Signal Outside Design Basis.Caused by Inadequate Mod Design at Time of Performance of Original Degraded Voltage Analysis. Engineering Analysis EA-FC-91-017 performed.W/910321 Ltr ML20029C1051991-03-18018 March 1991 LER 91-002-00:on 901209,ventilation Isolation Actuation Signal Generated by High Alarm on Process Radiation Monitor RM-062.Caused by Accumulation of Noncondensible Gases in Sample Piping.Valve Packing Leak repaired.W/910318 Ltr ML20029A2981991-02-0808 February 1991 LER 91-001-00:on 910109,determined That Containment Tendon Surveillances Performed in 1981 & 1985 Did Not Reflect Guidance in Tech Specs.Caused by Inadequate Administrative Controls.Testing Program Plan implemented.W/910208 Ltr ML20029A2971991-02-0606 February 1991 LER 90-022-02:on 900907,approx 460 Fire Barrier Penetration seals,60 Fire Dampers & 6 Fire Doors Declared Nonfunctional Per NRC Info Notice 88-004 Due to Lack of Documentation. Plant Outage Required to Implement Repairs/Replacements ML20028G9171990-09-28028 September 1990 LER 90-021-00:on 900829,inadvertent Reactor Protective Sys Actuation Occurred While Operator Changed Power Source. Caused by Operator Not Following Proper Procedures.Operator counseled.W/900928 Ltr ML20044B0131990-07-12012 July 1990 LER 90-018-00:on 900612,reactor Protective Sys (RPS) Trip Units for Axial Power Distribution Determined to Be Inoperable.Caused by Procedural Deficiencies.Procedure Revised & RPS Surveillance Tests reviewed.W/900712 Ltr ML20043F6301990-06-11011 June 1990 LER 90-016-00:on 900511,accident Scenarios Identified by Which Auxiliary Feedwater Piping from Discharge of Turbine Driven Auxiliary Feedwater Pump FW-10 Can Be Overpressurized.Caused by Design deficiency.W/900611 Ltr ML20043F2441990-06-0707 June 1990 LER 90-015-00:on 900507,PORV Variable Setpoints Used for Low Pressure Overpressure Protection Determined to Be Nonconservative for PORV Opening Time.Caused by Design Deficiency.Tech Spec Amend prepared.W/900607 Ltr ML20043C0991990-05-29029 May 1990 LER 90-014-00:on 900427,investigation Revealed That Component Cooling Water Piping to Reactor Coolant Pump Seal Coolers Could Be Targets of High Energy Line Break.Safety Analysis for Operability completed.W/900529 Ltr ML20042G7211990-05-10010 May 1990 LER 90-011-00:on 900402,inadvertent Actuation of Pressurizer Pressure Low Signal Occurred While Performing Calibr Procedure.Caused by Inappropriate Action by Technician Involved.Validation of Procedures reviewed.W/900510 Ltr ML20042E6871990-04-23023 April 1990 LER 90-007-01:on 900228,determined That Several Supports Would Be Overloaded During Seismic Event on Nonsafety Related & safety-related Main Steam Piping.Caused by Design Deficiency.Piping Supports modified.W/900423 Ltr ML20042E6861990-04-23023 April 1990 LER 90-009-00:on 900316,potential Overpressurization of Auxiliary Feedwater Piping Could Have Occurred During Thermal Expansion of Process Fluid Between Closed Valved. Caused by Design deficiencies.W/900423 Ltr ML20012E7641990-03-26026 March 1990 LER 90-005-00:on 900223,determined That Spent Fuel Pool Area Charcoal Filtration Unit VA-66 Was Outside Design Basis. Caused by Insufficient Airflow Into Unit.Affected Updated SAR Analysis Will Be updated.W/900326 Ltr ML20012D0121990-03-19019 March 1990 LER 90-004-00:on 900217,lift Pressures for 6 of 10 Main Steam Safety Valves Found Outside Acceptance Criteria. Caused by Overly Restrictive Operability Criteria.Valves Recalibr & License Amend Submitted to NRC.W/900319 Ltr ML20012D0101990-03-19019 March 1990 LER 90-003-00:on 900216,determined That Auxiliary Feedwater Piping Outside Normal Stress Limits of ASME Code & Design Basis Specified in Updated Sar.Caused by Design Deficiency.Valve Operators Will Be inspected.W/900319 Ltr ML20012B6361990-03-0909 March 1990 LER 89-017-01:on 890624,internal Valve Component from Check Valve Found Lying on Pump Discharge Vane.Repair or Replacement of Valve Internals Could Not Be Accomplished within Time Requirement of Tech Spec.W/900309 Ltr ML20006E1041990-02-0909 February 1990 LER 90-001-00:on 900108,fire Barrier for Wall Between Auxiliary Bldg Rooms 26 & 34 Breached But Hourly Fire Watch Patrol Not Established.Caused by Lack of Sufficient Training for Shift Supervisors.Standing Order revised.W/900209 Ltr ML20011E2691990-02-0505 February 1990 LER 89-024-00:on 891221,determined That Containment Spray Pumps & Suction Header Piping Not Constructed for Use as Backup to LPSI Sys for Shutdown Cooling.Caused by Inadequate Review of Assumptions.Firewatch established.W/900205 Ltr ML20011E2271990-02-0101 February 1990 LER 89-021-00:on 891010,util Informed by C-E of Potential Nonconservative Setpoint in Reactor Protection Sys Thermal Margin/Low Pressure Trip Unit.Caused by Error in Incorporating Transient Setpoint analyses.W/900201 Ltr ML20005F7151990-01-10010 January 1990 LER 89-023-00:on 891211,hourly Firewatch Patrol Entered Posted High Radiation Area W/O Meeting Entry Requirements for Area.Briefings on High Radiation Entry Requirements Held for Personnel W/Assigned dosimetry.W/900110 Ltr ML19354D6381989-12-20020 December 1989 LER 89-022-00:on 890805,change to Surveillance Procedure ST-CEA-1 Became Effective Which Would Have Made Both Emergency Diesel Generators Simultaneously Inoperable During Portion of Test.Change removed.W/891220 Ltr ML19332E7431989-12-0808 December 1989 LER 88-037-01:on 881214,one of Two Supply Headers Supplying Fire Suppression Headers in Auxiliary Bldg Isolated.Caused by Lack of Procedural Guidance & Inadequate Procedural Controls.Standing Order G-58 Will Be revised.W/891208 Ltr ML19332E2681989-12-0101 December 1989 LER 89-016-02:on 890616,for Unknown Period Since 890614, Auxiliary Feedwater Pump FW-10 Operated Outside Design Basis for Certain Accident Conditions.Caused by Inoperable Speed Control Loop.Action Plan implemented.W/891201 Ltr ML19351A4541989-11-22022 November 1989 LER 89-020-00:on 891012,determined That Two of Four Component Cooling Water HXs Simultaneously Inoperable for More than 24 H.Caused by Inadequate Controls Re Return of Equipment to Svc.Standing Order revised.W/891122 Ltr ML19327B5481989-10-24024 October 1989 LER 89-019-00:on 890924,indication of High Temp for Reactor Coolant Pump RC-3A Upper Motor Thrust Bearing Received in Control Room.Caused by Damaged Cable for Bearing Resistive Temp Device.Damaged Cable replaced.W/891024 Ltr ML19325D2471989-10-13013 October 1989 LER 89-012-01:on 890502,main Feedwater Isolation Valve to Steam Generator a Found Inoperable Due to Improperly Set Torque Switch.Caused by Inadequate Program for Maint of Motor Operated Valves.Torque Switches reset.W/891013 Ltr ML20028C7711983-01-0606 January 1983 LER 82-020/03L-0:on 821207,during Main Steam Safety Valve Test,Four Main Steam Safety Valves Had Lift Setpoints Out of Tolerance.Caused by Normal Drift of Valves Over Operating Cycle.Valves Readjusted ML20028B5451982-10-28028 October 1982 LER 82-019/03L-0:on 821024,MSIVs HCV-1041A & HCV-1042A Stopped Three to Four Degrees Off Seat When Signaled to Close.Caused by Binding Between Valve Packing & Shaft. Packings Sprayed W/Penetrant Oil ML20052J0631982-04-27027 April 1982 LER 82-009/03L-0:on 820411,while Exchanging Component Cooling Water Heat Exchangers,Associated Outlet Valves HCV-490B,HCV-491B & HCV-492B Failed to Open.Cause Not stated.HCV-491B Reassembled & Tested ML20052B2361982-04-0707 April 1982 LER 82-006/03L-0:on 820323,during Surveillance Test ST-ISI- WD-1,F.1,valve HCV-506A Failed to Close Via Control Room Switch.Caused by Solenoid Valve Malfunction.Solenoid Valve Disassembled,Cleaned & Reassembled ML20052D9291982-04-0606 April 1982 LER 82-008/03L-0:on 820330,during Performance of ST-FW-1, F.2(b)(6)per Tech Spec 3.9,steam Driven Auxiliary Feedwater Pump Failed to Start.Caused by Back Pressure Trip Lever in Tripped Position.Lever Reset ML20041G1291982-02-22022 February 1982 LER 82-005/03L-0:on 820210,at 98% Power,Control Element 24 Inserted Into Core.Emergency Procedure EP-13,CEDM Malfunctions,Implemented & Power Stabilized at 88%.Caused by Erroneous Operating Instruction.Instruction Changed ML20041F7481982-02-17017 February 1982 LER 82-003/03L-0:on 820203,containment Isolation Valve Associated W/Gas Vent Header HCV-507A Failed to Close on Demand.Caused by Solenoid Valve Plunger Sticking in Energized Position.Plunger Freed ML20041F6251982-02-0505 February 1982 LER 82-004/03L-0:on 820203,small Quantity of Radioactive Gas/Particulate Released to Auxiliary Bldg During Routine Operation.Caused by Failure of Stack Gas Monitor RM-062 to Alarm at Appropriate Setpoint Due to Faulty Alarm Module ML20041B1051982-01-28028 January 1982 LER 82-002/03L-0:on 820114,at 99% Power,Lockout Relay 86B1, Containment Radiation High Signal,Failed to Actuate on Demand by Plant Radiation Monitoring Sys.Caused by Burnt Coil on Lockout Relay.Coil Replaced & Tested Satisfactorily ML20041B1171982-01-19019 January 1982 LER 82-001/03L-0:on 820111,during Normal Operation,Two Fire Barrier Penetrations Found Nonfunctional.Shift Supervisor Immediately Notified;However,Fire Watch Not Posted.Insp & Supervisor Personnel Instructed on Proper Actions ML20039B4561981-12-11011 December 1981 LER 81-011/03L-0:on 811113,containment Isolation Valves Opened & Ventilation Process Initiated W/Containment Air Monitor RM-050/051 Inoperable.Caused by Personnel Error. Valves Closed ML20010H8581981-08-27027 August 1981 LER 81-008/03L-0:on 810813,86B/CRHS (Containment Radiation High Signal) Lockout Relay Failed to Actuate When RM-062 Was Placed in Alarm,Resulting in Failure of 86B1/CRHS Relay to Actuate.Caused by Dirt in Relay Latching Mechanism ML20041F6291981-08-27027 August 1981 LER 81-008/03L-1:on 810813,containment Radiation High Signal 86B Lockout Relay Failed to Actuate When Radiation Monitor RM-062 Placed in Alarm.Caused by Bound Relay Latching Mechanism Due to Dirt & Grease.Latch Cleaned ML20010C2271981-07-0707 July 1981 LER 81-006/03L-0:on 810624,reactor Protection Sys Nuclear Power Recorder Channel B Trip Setpoints Determined to Be Nonconservative.Caused by Faulty Temp Change Power Calculation Due to Grounded Hot Leg Temp Loop ML20004B1111981-05-0606 May 1981 LER 81-005/03L-0:on 810423,dc Sequencer Timers AC-3A (Component Cooling Water Pump) & AC-102A (Raw Water Pump) Failed to Time Out within Prescribed Limit.Cause Unknown Mechanisms Satisfactorily Inspected 1993-07-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217B5401999-10-0606 October 1999 Safety Evaluation Supporting Amend 193 to License DPR-40 ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data LIC-99-0096, Monthly Operating Rept for Sept 1999 for Fcs,Unit 1.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Fcs,Unit 1.With ML20211J9321999-09-0202 September 1999 Safety Evaluation Concluding That Licensee Proposed Alternatives Provide Acceptable Level of Quality & Safety. Proposed Alternatives Authorized for Remainder of Third ten- Yr ISI Interval for Fort Calhoun Station,Unit 1 LIC-99-0084, Monthly Operating Rept for Aug 1999 for Fort Calhoun Station.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Fort Calhoun Station.With ML20216E6431999-08-26026 August 1999 Rev 19 to TDB-VI, COLR for FCS Unit 1 ML20210R1961999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Fcs,Unit 1 ML20210G2181999-07-27027 July 1999 Safety Evaluation Supporting Amend 192 to License DPR-40 ML20210D9951999-07-22022 July 1999 Safety Evaluation Supporting Amend 191 to License DPR-40 ML20216E6361999-07-21021 July 1999 Rev 18 to TDB-VI, COLR for FCS Unit 1 ML20210R2081999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Fcs,Unit 1 LIC-99-0065, Monthly Operating Rept for June 1999 for Fort Calhoun Station,Unit 1.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Fort Calhoun Station,Unit 1.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20210P5461999-06-0808 June 1999 Rev 0,Vols 1-5 of Fort Calhoun Station 1999 Emergency Preparedness Exercise Manual, to Be Conducted on 990810. Pages 2-20 & 2-40 in Vol 2 & Page 4-1 in Vol 4 of Incoming Submittal Not Included ML20195B4581999-05-31031 May 1999 Rev 3 to CE NPSD-683, Development of RCS Pressure & Temp Limits Rept for Removal of P-T Limits & LTOP Requirements from Ts ML20207H7401999-05-31031 May 1999 Performance Indicators Rept for May 1999 LIC-99-0053, Monthly Operating Rept for May 1999 for Fort Calhoun Station,Unit 11999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Fort Calhoun Station,Unit 1 ML20195B4521999-05-17017 May 1999 Technical Data Book TDB-IX, RCS Pressure - Temp Limits Rept (Ptlr) ML20206L4241999-05-10010 May 1999 Safety Evaluation Supporting Corrective Actions to Ensure That Valves Are Capable of Performing Intended Safety Functions & OPPD Adequately Addressed Requested Actions Discussed in GL 95-07 ML20206M2601999-05-0606 May 1999 SER Concluding That Licensee IPEEE Complete Re Info Requested by Suppl 4 to GL 88-20 & IPEEE Results Reasonable Given FCS Design,Operation & History LIC-99-0047, Monthly Operating Rept for Apr 1999 for Fort Calhoun Station Unit 1.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Fort Calhoun Station Unit 1.With ML20195E8621999-04-30030 April 1999 Performance Indicators, for Apr 1999 ML20205Q5831999-04-15015 April 1999 Safety Evaluation Supporting Amend 190 to License DPR-40 ML20210J4331999-03-31031 March 1999 Changes,Tests, & Experiments Carried Out Without Prior Commission Approval for Period 981101-990331.With USAR Changes Other than Those Resulting from 10CFR50.59 ML20206G2641999-03-31031 March 1999 Performance Indicators Rept for Mar 1999 LIC-99-0034, Monthly Operating Rept for Mar 1999 for Fcs,Unit 1.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Fcs,Unit 1.With ML20205J8181999-02-28028 February 1999 Performance Indicators, for Feb 1999 LIC-99-0025, Monthly Operating Rept for Feb 1999 for Fort Calhoun Station,Unit 1.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Fort Calhoun Station,Unit 1.With ML20207F3291999-01-31031 January 1999 FCS Performance Indicators for Jan 1999 ML20203B0991998-12-31031 December 1998 Performance Indicators for Dec 1998 LIC-99-0026, 1998 Omaha Public Power District Annual Rept. with1998-12-31031 December 1998 1998 Omaha Public Power District Annual Rept. with LIC-99-0003, Monthly Operating Rept for Dec 1998 for Fort Calhoun Station.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Fort Calhoun Station.With ML20198S3771998-12-31031 December 1998 Safety Evaluation Supporting Amend 189 to License DPR-40 ML20198S4831998-12-31031 December 1998 Safety Evaluation Supporting Amend 188 to License DPR-40 ML20196G2251998-12-18018 December 1998 Rev 2 to EA-FC-90-082, Potential Over-Pressurization of Containment Penetration Piping Following Main Steam Line Break in Containment ML20198M3141998-11-30030 November 1998 Performance Indicators Rept for Nov 1998 LIC-98-0172, Monthly Operating Rept for Nov 1998 for Fort Calhoun Station,Unit 1.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Fort Calhoun Station,Unit 1.With LIC-98-0160, Special Rept:On 981113,MSL RM RM-064 Was Declared Inoperable Due to Leakage Past Isolation Valve HCV-922.Troubleshooting Has Indicated That Leakage Has Stopped & Cause of Leak Continues to Be Investigated1998-11-25025 November 1998 Special Rept:On 981113,MSL RM RM-064 Was Declared Inoperable Due to Leakage Past Isolation Valve HCV-922.Troubleshooting Has Indicated That Leakage Has Stopped & Cause of Leak Continues to Be Investigated ML20203B0721998-11-16016 November 1998 Rev 6 to HI-92828, Licensing Rept for Spent Fuel Storage Capacity Expansion ML20196E4981998-10-31031 October 1998 Performance Indicators Rept for Oct 1998 ML20196G2441998-10-31031 October 1998 Changes,Tests & Experiments Carried Out Without Prior Commission Approval. with USAR Changes Other than Those Resulting from 10CFR50.59 LIC-98-0154, Monthly Operating Rept for Oct 1998 for Fort Calhoun Station,Unit 1.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Fort Calhoun Station,Unit 1.With ML20154M4881998-10-19019 October 1998 Safety Evaluation Supporting Amend 186 to License DPR-40 ML20154N2411998-10-19019 October 1998 Safety Evaluation Supporting Amend 187 to License DPR-40 LIC-98-0136, Monthly Operating Rept for Sept 1998 for Fort Calhoun Station,Unit 1.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Fort Calhoun Station,Unit 1.With ML20155G4261998-09-30030 September 1998 Performance Indicators for Sept 1998 ML20154A1251998-08-31031 August 1998 Performance Indicators, Rept for Aug 1998 LIC-98-0122, Monthly Operating Rept for Aug 1998 for Fort Calhoun Station Unit 1.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Fort Calhoun Station Unit 1.With ML20238F7231998-08-17017 August 1998 Owner'S Rept for Isis ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency 1999-09-30
[Table view] |
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l March 9, 1990 l-LIC-90-112 U. S. Nuclear Regulatory Commission Attn: Document Control Desk-Mail Station P1-137 )
-Washington, DC 20555
Reference:
(1)DocketNo.50-285 (2)LicenseeEventReport89-017, July 31,1989(LIC-89-677)
Subject:
. Licensee Event Report 89-017, Revision-1 for the Fort Calhoun l Station Gentlemen:
Please find attached Licensee Event Report 89-017, Revision 1 dated March l 9,~ 1990.- The ori
- 50.73 (a)(2)(ii)(ginal report B). This was submitted revision perupcated provides:an recuirements of 10 CFR status of .
corrective actions; revised portions are noted by vertical bars in the .l margins.
If.you should have any questions, please contact me.
Sincerely, i
/wW.G. Gates u Division Manager l Nuclear Operations WGG/tcm l Attachment l
.c: R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager .
P. H. Harrell, NRC Senior Resident Inspector
.INP0 Records Center American Nuclear Insurers t
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FACILITV alAME (11 DOCEf f NueettR (26 raer cri Fort Calhoun Station Unit No. 1 o 15 l 0 l o l o l 21815 1Iorl016 TITLt les Raw Water System Outside Its Desion Basis EV9NT DAf t (St LER NUMBER t$1 REPORT DAf t (7) OTHER F ACILITill INv0Lvt0 tel
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NiME TELEPHONE NUMBER ARE A C004 D. J. Molzer, Shift Technical Advisor 4 0 12 51313 l -l 618 I 914 COMPLET8 ONE LINE FOR E ACH COMPONENT F AILURE OtsCRistD IN THit REPORT (136 C. RE POR T A LE y M C- R PORTA LE CAust $vsTEM COMPONENT MQ p p ,
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On June 24, 1989, during the removal of the pump assembly on Raw Water pump AC-10A for maintenance, an internal valve component from check valve RW-125 was found lying on the pump discharge vane. Repair or replacement of the valve internals could not be accomplished within the time requirement of the Technical Specification LCO.
On June 29, 1989 plant management decided to place the check valve body back into the system with the internals removed. A Safety Analysis for Operability was performed to justify continued operation outside the system design basis with prescribed compensatory actions. On June 30, 1989, the affected train of the Raw Water system was returned to service. Pursuantto10CFR50.72(b)(1) roximatel 1610 on June 30, 1989. This LER (ii)(B),theNRCwasnotifiedatapp(b)(2)(ii)y(B).
is submitted pursuant to 10CFR50.73 The plant was operating at 100 percent power during this period.
This event is due primarily to lack of long term )reventive maintenance and inservice testing programs for the Raw Water checc valves. Previously committed elements of the Fort Calhoun Station Safety Enhancement Program should preclude recurrence. RW-125 and corresponding check valves in the other trains of the Raw Water system have been replaced. l NRC Form att 1649)
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POftM 3ASA U.S.EUCLE3.2 att AULAThY COMMemioss 4
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TEXT CONTINUATION 2""!,",'o',"4 '
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Fort Calhoun Station Unit No. 1 o is Io Io Io l2 l 8l5 49 or 0l6 Text u . u . cr ,am m nn 011l7 -
0l1 42 3rovide cooling for the Component The Raw Water (RW))
CoolingWater(CCW systemsystem and supply is designed to baccup cooling to engineered safeguards equipment during normal and accident modes of o Raw water is supplied ;
to Fort Calhoun Station by)the MissourifourRiver.
Each of the raw water peration.
p(umps (AC-10A,B,C,andD has an air operated discharge isolation valve HCV-2850,HCV-2851,HCV-2852,andHCV-2853,respectively)thatisinterlocked to the respective ) ump's starting / stopping circuitry. The discharge isolation valve will close wien its associated pump is de-energized and will open when the pump receives a start signal. The discharge valves are equipped with air accumulatorsandhandwheelsformanualclosurecapability(handjacking). Each gumpissuppliedwithadischargecheckvalve(MissionMfg.Co. Duo-ChekStyle B)whichpreventsbackflowthoroughanon-runningpumpduringtheperiodof time its discharge isolation valve remains open. These are original equipment check valves which have been in service since plant startup in 1973.
In May 1988, the raw. water pump discharge check valves were added to the InserviceTesting(IST) program. A section was incorporated into surveillance ,
test ST-ISI-RW-1 for the quarterly testing of the check valves in the closed '
position. The surveillance provided a method in which closure of the check-valves could be verified and provided a means of trending backleakage so that potential degradation of the valves could be identified. Excessive backflow in two of the cileck valves, RW-117 ('C' pump) and RW-115 ('D' pump), was identified when the valves were first tested on July 5, 1988. Subsequently, an incident report was initiated on July 7, 1988 to document the problem and determine reportability. Management determined that the problem was not reportable since credit was taken for the discharge isolation valves providing the backflow prevention function for the check valves. However, this evaluation failed to consider the loss of a DC bus as the most limiting single
! failure. Had a loss'of a single DC bus been considered at the time, it would have been discovered that two of the discharge isolation valves would have failed open almost immediately after their respective solenoids had de-energized. Corrective actions associated with the Incident Report concentrated on upgrading the instrument air su) ply to the discharge valves.
It was felt these actions would ensure the disc 1arge valves were capable of being held closed during a' loss of instrument air and thus provide the backflow prevention function of the check valves.
To ensure the valves were capable of satisfying this functicn, a tem modification was initiated to install Critical Quality Element (CQE)poraryqualified air check valves in the supply lines to the isolation valves. To increase the reliability of the instrument air supply to the valves, permanent modification I
MR-FC-88-61 was completed by the end of the 1988 refueling outage. The I modification provided for the replacement of all instrument air tubing, fittings, and air check valves with CQE qualified components.
NRc P.-, asea m
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1 , APE aWO HE U TION t 60 II IC oF MANAGEME NT AND Su00f T. WA$HINGTON, DC 20603. +
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0l 1 01 3 or 0l6 TfXT IJaewe sysse a r M. see esissuonar Nec Fenn asE4'es itM Also, the supports and accumulators were seismically qualified and a functional test was performed on the accumulators to verify they were capable of holding the discharge valves closed for at least 30 minutes following a loss of #
instrument air. It was felt that 30 minutes would provide tl1e operator sufficient time to manually hand-jack the valves closed. The accumulators have since been functionally tested during the periodic inservice testing of the air check valves, s When excessive backleakage through the two check valves was initially identified in July 1988, an effort was started to obtain replacement parts for the check valves. Attempts to locate readily available CQE identical replacement parts for the valves were ultimately unsuccessful. In April 1989 it was then decided that the best course of action would be to initiate a modification request (MR-FC-89-53) to install four new check valves. Based on the credit taken for backflow prevention capability of the discharge isolation valves and the actions taken to enhance the operability of these isolation valves, a high priority was not placed on efforts to repair or replace the check valves.
When pump AC-10A was taken out of service for maintenance on. June 23, 1989, the plant entered a 7 day Limiting Condition for Operation (LC0) in accordance with TechnicalSpecification2.4(1)c. The Technical. Specification states, in part, that "when the river water tem)erature is greater than 60 degrees Fahrenheit, an inoperable Raw Water pump s1all be restored to operability within 7 days or the reactor shall be placed in a hot shutdown condition within 12 h~ours." The plant was operating at 100 percent power with river water temperature greater than 60 degrees F.
On June 24, 1989, during the removal of the pump assembly on Raw Water Pump
. AC-10A for maintenance, an internal component from discharge check valve RW-125 :
! was found lying on the pump discharge vane. Failure of the check valve's internals was determined to be the result of excessive wear to the hinge pin housing in the valve body, to the valve seat, and to the lugs on the valve disc plate.
1 Upon discovery of-the damaged check valve a renewed attempt was made to obtain L
replacement parts or a new check valve. Efforts to obtain cualified CQE I' replacement parts for RW-125 were again unsuccessful. In ac dition, due to the l
nature of damage to the valve's internals, it was determined that the valve was l beyond repair with available resources. Replacement of the valve was found to be a lengthy process since suitable valves require about 16 weeks manufacturing and delivery time. Repair or replacement of the check valve internals could I therefore not be accomplished within the time requirement of the Technical L Specification LCO.
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MME'.We'"WE u'"/ OCfifA 1 P APE RWO RE TI'ON J l of MANAGEMENT AND OUDGET,W ASHINGTON, DC 20603.
FACILITY NAtat Hi DJC'AtT NUMOtR (2)
LER NUGA$tR (el Pact (3)
"'" a syn Fort Calhoun Station Unit No. 1 ol5 o lo j o l 2l 8l5 8l 9 0l1 l 7 0 l1 0l 4 or O l 6 l Tm a--. . o w acr 3 a m, i
l On-June 29, 1989 plant management decided to place the check valve body I (RW-125) back into the system with its internals removed. Work was covered i under temporary modification TM-89-M-034. A Safety Analysis for Operability I (SA089-10)andsupporting10CFR50.59safetyevaluationwereperformedto I justify continued plant operation and to address safety considerations for Raw Water system operation outside its design basis. This analysis assumed loss of a DC bus. The safety evaluation confirmed that the temporary modification to the Raw Water system did not result in a unreviewed safety question, contingent on im)1ementation of adequate procedural guidance, special operator training, and tie designation of an operator on eac1 shift to perform the manual isolation of HCV-2850, if required.
On June 30, 1989, the "A" train of the raw water system was returned to service to satisfy the Technical Specification LCO. The raw water system was then in a
- condition outside of its design basis since it was not capable of performing its design function without o)erator action. The plant was operatin at 100%
L power and Mode 1 throughout t11s period. Pursuantto10CFR50.72(b)(g)(ii)(B) 1 L the NRC was notified of the plant being outside its design basis at 1610 on June 30, 1989. This LER is submitted pursuant to approximately(2)(ii)(B).
10CFR50.73(a)
L The limiting event for which the Raw Water system is analyzed is a Loss of Coolant Accident (LOCA) coincident with a loss of all off-site power and a l single failure. For all postulated DBA events with loss of off-site power and l
a single failure, with the exception of a loss of a DC bus, it can be assumed that the instrument air supply to the discharge valves is capable of holding the valves closed for et least 30 minutes following a loss of instrument air.
l The worst case single failure scenario-is postulated to be a loss of a single DC bus. It is assumed that the loss of the No. 1 DC bus would prevent the No. 1 diesel (DG-1)fromautomaticallystarting. Failure of DG-1 to start coincident with loss of off-site power would leave 'A' and 'C' Raw Water pumps without supply power. The loss of the No. 1 DC bus would also de-energize the
. solenoids which operate the instrument air pilot valves for discharge valves HCV-2850('A'RWpump)andHCV-2852('C'RWpump). Loss of power to the solenoid operated pilot valves would cause instrument air pressure to vent off the air operator, allowing the discharge valves to fail open. Thus, a reverse flowpathwouldbecreatedthroughtheopendischargevalve(HCV-2850)andthe nonfunctional check valve (RW-125), back to the river. Engineering evaluation has shown that the amount of RW flow available to the CCW heat exchangers would be insufficient to maintain a CCW temperature less than 200 degrees F during DBA conditions, considering the amount of backleakage through AC-10A plus other predicted RW flow losses. Approximately 10 minutes would be available for operator response to this event before the CCW system reached its design limit temperature of 200 degrees F. It is assumed in USAR Section 6.3.4 that at least one Containment Spray pump will start under these conditions and provide containment heat removal capability.
NRC Penn 30BA (649)
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- To ensure that appropriate operator actions were taken prior to Component Cooling Water temperature reaching its design limit during a DBA event, the following compensatory measures were implemented:
- 1. Emergency Operating Procedure E0P-20, " Functional Recovery", was revised to provide instruction to the operator for recognizing a raw water pump reverse flow condition, and direct the operator to the appropriate procedure for corrective actions, i
- 2. Abnormal Operating Procedure A0P-18, " Loss of Raw Water", was revised to provide specific guidance for manually isolating HCV-2850 and other sources of RW losses.
- 3. Operating Instruction 01-RW-1 was changed to provide guidance to the operators for stopping AC-10A.
- 4. Training was conducted on the revised procedural requirements for on-shift operations personnel. Additionally, a designated operator was assigned on g each shift to immediately perform the manual closure of HCV-2850, if required.
l-l The failure of RW-125 and subsequent operation outside design basis can be i attributed to lack of preventive maintenance and inservice testing programs for the Raw Water check valves. Contributing factors were the lack of available spare parts, inadequate evaluation of system operability following initial identification of' valve leakage in 1988, and delays in repair / replacement efforts.
An engineering evaluation was completed for the replacement of the four check valves with a similar type check valve. New valves were scheduled for installation as soon as possible after receipt, with priority given to replacement of RW-125.
On January 22, 1990 Raw Water Pump AC-10A was removed from service for replacement of check valve RW-125. Following valve replacement and successful post-maintenance testing, Pump AC-10A was returned to service and declared operable later on January 22, 1990.
On January 24, 1990 replacement of check valve RW-115 for Raw Water Pump AC-10D began. One of the two valve " flappers" was determined to be missing. Attempts to-locate the flapper in immediately adjacent piping were unsuccessful, so it was decided to remove pump AC-10D from its well in order to inspect the rotating assembly for signs of the fla)per. The last previous operation of pump AC-10D was from January 19 througi 20, 1990. From past experience, any foreign material inside this type pump caused excessive vibration level during operation. Acceptable vibration levels during operation of pump AC-10D on January 20 indicated that the missing flapper was not in the pump at that time. At 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> on January 24, pump AC-10D was removed from the pump well and the flapper from check valve RW-115 was found on top of the pump second stage. The rotating assembly was visually inspected, revealing no indications that the flapper was inside the pump during previous operation. Replacement of RW-115 was completed, pump AC-100 was reinstalled, and post-maintenance testing was successfully completed. Pump AC-10D was then declared operable.
NIC form 308A (S89)
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, ACILITY NAME Hi DOCILE T NUM0t h (2)
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Concurrent with the removal and inspection of pump AC-100, it was decided to l test check valves RW-121 and 117 for Raw Water Pumps AC-10B and AC-10C, !
respectively. The performance of existing ASME Section XI surveillance test OP-ST-RW-3004 (plus additional flow measurements) would determine if the flappers for these valves were still intact by indicating the degree of check valve backleakage. Check valve RW-121 was tested first. The Section XI criterion was not met, so RW-121 and associated pump AC-10B were promptly l declared inoperable in accordance with the station surveillance program i requirements. However, subsequent engineering evaluation of the measured d leakage determined that the check valve flappers were intact and that the affected train of the Raw Water system was capable of performing its design basis safety functicn even with the measured amount of backleakage. Based on this evaluation, RW-121 and AC-10B were declared operable.
1 On January 25, replacement of check valve RW-121 began. Concurrently, testing of check valve RW-117 was performed. The results were similar to those for RW-121, i.e., the Section XI criterion was not met but evaluation determined the integrity of the valve flappers and the capability of the affected Raw- .
Water train to perform its safety function. Replacement of check valve RW-117 was begun and by 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on January 26, 1990, both RW-121 and RW-117 had been replaced, tested, and declared operable.
After replacement of check valve RW-125 was completed, SA0-89-10 was no longer required and was cancelled by the Plant Review Committee Chairman on February 1, 1990. The surveillance testing noted above and additional monitoring by the System Engineer showed that the plant was operating safely for the duration of L SA0-89-10.
l A Station Procurement Group has been established to enhance the identification L
and procurement of equipment and parts.
I' A station check valve test program has been implemented as a result of an INP0 finding and NRC Information Notice 88-70. Im)lementation of this program was previously committed as Reference No. 43 of tie Fort Calhoun Safety Enhancement l Program (SEP).
Other completed portions of the SEP will address the identified contributing
. factors. ReferenceNo.33(developon-linemaintenanceandmodification
! schedule) and Reference No. 34 (create and staff central planning group) should assure proper planning and prioritization. ReferenceNo.62(establishinterim system engineers) will provide increased technical expertise and attention to systems important to safety.
This is the first LER associated with the Raw Water system at Fort Calhoun Station being outside its design basis. Other similar reports concerning safety systems outside design basis include LER 89-12 and LER 89-16.
Nace = Ac ,