|
---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20029E2061994-05-11011 May 1994 LER 93-025-01:on 931205,Loop a MSIVs Exceeded TS Leakage Limit.Caused by Failure of Fastner Locking Devices.Seating Surface of Main Disc in MSIV 1A Machined & Successfully Pass LLRT on 940116.W/940511 Ltr ML17352B2501993-10-0101 October 1993 LER 93-014-00:on 930902,discovered That IRM 11 & APRM Both in Bypassed Condition W/O Half Scram Being Inserted Due to Work Practices.Irm 11 Taken Out of Bypassed Condition & Individual Involved counselled.W/931001 Ltr ML17352B2511993-10-0101 October 1993 LER 93-019-00:on 930903,RHR Pump 1A Inadvertently Started Due to Personnel Error.Pump Immediately Secured & Individual Involved Counselled on Importance of self-check.W/931001 Ltr ML20046A1801993-07-21021 July 1993 LER 92-007-00:on 930624,Toxic Gas Analyzer Inoperable Due to Personnel Error Caused by Lack of Procedural Adherence. Isolated CR Ventilation,Restarted Sample Pump & Restored CR Ventilation Sys to Outside Air suction.W/930721 Ltr ML20045H5781993-07-13013 July 1993 LER 93-013-00:on 930613,internal Electrical Fault within Main Power Transformer Tripped & Caused Reactor Scram & Unexpected Group I Isolation.Caused by Turbine Stop Valve Closure.Transformer replaced.W/930713 Ltr ML20045G7291993-07-0909 July 1993 LER 93-005-00:on 930609,numerous Alarms Received in CR, Including HPCI Turbine Rupture Disc High Pressure Alarm. Caused by Ruptured Disk,Releasing Steam/Water Mixture. Test Will Be Conducted on Rupture disks.W/930709 Ltr ML20045G0951993-07-0202 July 1993 LER 93-011-00:on 930602,HPCI Declared Inoperable in Order to Perform Qcos 2300-13.Caused by MSC Stem & Stem Gear Being Scored & Worn from Age Causing Excessive Friction.C/As Include Disassembling Gearbox Every 10 yrs.W/930702 Ltr ML20045F3321993-06-30030 June 1993 LER 93-012-00:on 930601,U-2 DG Cooling Water Pump Inoperable Due to Inadequate lubrication.U-2 DG Cooling Water Pump replaced.W/930630 Ltr ML20045F3251993-06-28028 June 1993 LER 93-007-01:on 930307,B Loop MSIV Exceeded TS Leakage Limits for Containment Isolation Valves Caused by Damaged Flexitalic Bonnet Gasket.Replaced Bonnet Gasket & Valve Successfully Retested (WR Q061987).W/930622 Ltr ML20044E6171993-05-14014 May 1993 LER 93-006-00:on 930420,Unit 2 Nso Inadvertently Started Unit 1 DG When Attempting to Start Shared (1/2) Dg.Caused by Personnel Error.Unit 1 DG Shut Down & 1/2 DG Started & Loaded successfully.W/930514 Ltr ML20044C9511993-05-0707 May 1993 LER 93-009-00:on 930408,technician Discovered That Estimate of Sample Flow for U2 Reactor Bldg Vent Sampler Flow Rate Monitor Not Calculated.Caused by Personnel Error.Technician counseled.W/930428 Ltr ML20024G6811991-04-19019 April 1991 LER 91-008-00:on 910322,reactor Bldg Ventilation Isolation Occurred.Caused by Lightning Strike.Control Room Vents Reset & Toxic Gas Sample Point a Selected.Addl Trips Associated W/ Lightning Strike Immediately reset.W/910418 Ltr ML20029C1241991-03-0808 March 1991 LER 91-005-01:on 910131,1/2 B Standby Gas Treatment Sys Autostart During RPS B Power Swap Due to an Inadequate Procedure.Nso Reset 1/2 Scram & 1/2 Groups II & III isolations.W/910306 Ltr ML20029B5901991-03-0808 March 1991 LER 91-004-00:on 910211,standby Liquid Control Sys of Unit 1 & 2 Inoperable.Caused by Inadequate Mod Testing.Engineering Dept Evaluating Data & Temporary Procedure 6589 Generated. W/910308 Ltr ML20028H7841991-01-23023 January 1991 LER 90-032-00:on 901224,1/2A Diesel Fire Pump Taken out-of- Svc on 901217 & Not Returned to Svc Before Seven Day Time Allotment Expired.Caused by Mgt Deficiency.Fire Pump Successfully Tested & Returned to svc.W/910123 Ltr ML20028H6831991-01-21021 January 1991 LER 90-034-00:on 901223,high Chlorine Concentration Caused Control Room Ventilation Manual Isolation & ESF Actuation. Caused by Instrument Error Code & Misinterpretation of Analyzer Indication.Flow reduced.W/910121 Ltr ML20028G9141990-09-28028 September 1990 LER 90-018-00:on 900829,plant Outside Design Spec for Electrical Separation Criteria for Two Redundant Safety Sys. Caused by Inadequate Engineering Review During Leads installation.W/900928 Ltr ML20044B2401990-07-12012 July 1990 LER 90-012-00:on 900612,control Room Ventilation Emergency Air Filtration Unit Declared Inoperable.Caused by Heater Malfunction.Work Request initiated.W/900712 Ltr ML20044B0091990-07-11011 July 1990 LER 90-011-00:on 900611,diesel Fire Pump 1/2 a Out of Svc for Period Exceeding 7 Days Reporting Criteria,To Install New Suction Line.Caused by Required Maint Work.Repairs Completed & Pump Tested & Returned to svc.W/900711 Ltr ML20043H4831990-06-21021 June 1990 LER 90-010-00:on 900522,while Returning RWCU Sys to Svc, Nonregenerative HX High Temp Alarm Received,Challenging ESF Logic & Causing Group III Isolation.Caused by Leaking RWCU Check Valves.Work Requests issued.W/900621 Ltr ML20043H2931990-06-18018 June 1990 LER 90-009-00:on 900518,determined That Various Containment Vols Not Leak Rate Tested Due to Recent 10CFR50,App J Interpretation Re Licensing Design Criteria.Mod M4-1(2)-89-167 initiated.W/900618 Ltr ML20043H9521990-06-13013 June 1990 LER 90-016-01:on 900411,motor Control Ctr Relay 28/29-5 Setpoint Drift Occurred,Resulting in Analyzed Plant Condition.Caused by Utilizing Wrong Relay & Inadequate Review of GE Svc Info Ltr.Relay replaced.W/900613 Ltr ML20043E9121990-06-0707 June 1990 LER 90-006-00:on 900508,Unit 2 RCIC Declared Inoperable Due to Unstable Operation of RCIC Pump Flow Controller.Caused by Proportional Band of Controller Being Set to Respond to Changes in Flow Too Quickly.Flow controlled.W/900607 Ltr ML20043F2341990-06-0101 June 1990 LER 90-001-01:on 891208,seven Pathways Not Included in Type B & C Local Leak Rate Test Program.Caused by Misinterpretation of 10CFR50,App J Requirements.Required Vents & Test Taps Will Be installed.W/900601 Ltr ML20043F2161990-06-0101 June 1990 LER 89-024-01:on 891214,main Turbine Unexpectedly Tripped Following Isolation of Reactor Water Level Switch (Lits). Caused by Channel B Lits Having Been Previously Replaced W/ Switch That Operated Reverse function.W/900601 Ltr ML20042E4991990-04-17017 April 1990 LER 90-007-00:on 900318,ESF Actuation Occurred.Caused by Mgt Deficiency in Not Providing Sufficient Guidance for Review of Work Package After Scope Changed.New Work Package Preparation Procedures initiated.W/900417 Ltr ML20012C7231990-03-15015 March 1990 LER 90-004-00:on 900213,loss of Emergency Bus 23-1 Occurred Due to Shorted Conductor Cable While Performing Wiring Verification.Caused by Personnel Error & Improper Installation.Electric Power Supplies recovered.W/900315 Ltr ML20012C7171990-03-15015 March 1990 LER 90-003-00:on 900213,unit Diesel Generator Tripped on Overspeed Upon Manual Startup for Testing.Caused by Misadjustment of Diesel Generator Governor.Diesel Generator Equipment Reset & Governor adjusted.W/900315 Ltr ML20012B2931990-03-0505 March 1990 LER 90-002-00:on 900204,determined That Only Reactor Bldg Differential Pressure of 0.24-inch Water Vacuum Could Be Obtained,In Violation of Tech Spec Required 0.25 Inch.Caused by Testing Deficiency.Leak Paths sealed.W/900305 Ltr ML20012B2961990-03-0505 March 1990 LER 90-003-00:on 900205,determined That Leakage Rate for HPCI Sys Steam Exhaust Check Valve Exceeded Tech Spec Limits.Caused by Excessive Leakage.Check Valve Replaced W/ Mission Check Valve W/Carbon Steel springs.W/900305 Ltr ML20011E1821990-01-30030 January 1990 LER 90-002-00:on 900102,during Annual Water Sprinkler Sys Valve Position Insp,Discovered That Valve 2-4199-072 Not Cycled Per Tech Spec Surveillance Requirement 4.12.Caused by Procedural Deficiency.Procedure to Be revised.W/900130 Ltr ML19354D9431990-01-15015 January 1990 LER 89-025-00:on 891220,determined That Reactor Bldg Overhead Auxiliary Hook May Have Contacted Side of New Fuel Bundle on 891214,causing Extensive Damage.Caused by Personnel Error.Bundle Shipped Back to GE.W/900115 Ltr ML19354D9481990-01-15015 January 1990 LER 89-023-00:on 891219,identified Deficient Temporary Procedures Which Altered Intent of Original Procedure QAP 1100-7.Caused by Mgt Oversight in Preparation of Tech Spec Change.Temporary Procedures revised.W/900115 Ltr ML20005F5881990-01-0808 January 1990 LER 90-001-00:on 891208,seven Pathways Not Included in Type B & C Local Leak Rate Testing Program.Caused by Misinterpretation of 10CFR50,App J Testing Requirements.Mod of Sys to Be Performed as necessary.W/900108 Ltr ML20005G0231990-01-0202 January 1990 LER 88-020-01:on 880619,improper Valving Sequence Occurred Which Resulted in Various ESF Actuations.Caused by Inadequate Equipment out-of-svc Procedure.Procedure Revised & Training Lesson Plan developed.W/900102 Ltr ML20005E4531989-12-28028 December 1989 LER 89-022-00:on 891128,HPCI Declared Inoperable Following Unexpected Actuation of HPCI Pump Room Deluge Sys.Actuation Caused Dc Sys Grounds Due to Moisture Intrusion in Electrical Equipment.Equipment dried.W/891228 Ltr ML20005D9551989-12-22022 December 1989 LER 89-021-00:on 891124,HPCI Deluge Sys Functional Test Not Completed within Tech Spec Interval of 6 Months.Caused by Mgt Deficiency & Insufficient Procedure.Procedures Revised to Better Control Tech Specs.W/891222 Ltr ML19351A6391989-12-15015 December 1989 LER 89-020-00:on 891117,full Reactor Scram Occurred Due to Simultaneous Loss of Power to 1A & 1B 24/48-volt Distribution Panels During Panel Transfer.Caused by Personnel Error.Personnel Trained Re event.W/891215 Ltr ML19332E7581989-12-0505 December 1989 LER 89-019-00:on 891105,half Group II & III Isolation Occurred While Performing Valve Flow Check Surveillance. Caused by Leaky Isolation Valve on Transmitter.Surveillance Methods Will Be Reviewed & Procedures revised.W/891205 Ltr ML19332C7581989-11-20020 November 1989 LER 89-017-00:on 891022,discovered That 6-month Functional Test Procedure Qos 500-3 Re Functional Test for Reactor Protection Sys Electrical Assemblies Not Completed in Required Time.Caused by Mgt deficiency.W/891120 Ltr ML19327C0701989-11-0909 November 1989 LER 89-018-00:on 891012,notification Given of Potential Single Failure of Diesel Generator Voltage Regulator That Could Result in Loss of All But One ECCS Loop.Caused by Design Deficiency.New Procedure developed.W/891109 Ltr ML19327C0951989-11-0404 November 1989 LER 89-005-00:on 891012,reactor Scram Occurred Due to Turbine Stop Valve Closure.Caused by Personnel Error in That Work Analyst Overlooked Removal of Two Connections on Limit Switch.Maint Workers Immediately counseled.W/891104 Ltr ML19327C2631989-11-0101 November 1989 LER 87-009-01:on 870801,reactor Scram Occurred from Turbine Generator Load Mismatch Due to Generator Trip.Caused by Electrical Fault in Main Transformer C Phase Windings.Main Transformer Replaced w/spare.W/891103 Ltr ML19325E8261989-10-30030 October 1989 LER 87-012-01:on 870918,while Performing Qos 1600-1, Vacuum Breaker 2-1601-33E Remained Open After Testing.Caused by Binding in Test Cylinder Portion of Vacuum Breaker. Nuclear Work Requests completed.W/891030 Ltr ML19325E5921989-10-24024 October 1989 LER 87-017-01:on 870805,Group IV Isolation Received Which Resulted in Closure of HPCI Steam Supply Valves.Caused by Failed HPCI Steamline Differential Pressure Transmitter Due to Loss of Oil in Sensing cell.W/891024 Ltr ML19325E5971989-10-24024 October 1989 LER 88-003-01:on 880301,while Performing RCIC Monthly Test, RCIC Pump Could Only Achieve 400 Gpm Against Discharge Pressure of 500 Psig.Caused by Failed Hydraulic Actuator on Turbine Governor Valve.Actuator replaced.W/891024 Ltr ML19327B7901989-10-18018 October 1989 LER 89-016-00:on 890921,during Transfer of New Fuel,Fuel Assembly Released from Refueling Grapple & Fell Upon Spent Fuel Racks.Caused by Personnel Error & Procedural Deficiency.Fuel Handling Procedures Revised ML19327B0491989-10-16016 October 1989 LER 89-015-00:on 890916,during Transfer,Loss of Power to Off Gas Monitor a Caused Contacts to Open Giving Upscale Radiation Signal,Thus Starting Off Gas Timer.Caused by Inadequate Procedures.Off Gas Sys reset.W/891016 Ltr ML19325C5341989-10-0202 October 1989 LER 89-014-00:on 890910,determined That Combined Leakage Rate from All Penetrations & Valves Exceeded Tech Spec Limit.Root Cause Unknown.No Corrective Action Taken. Supplemental Rept Will Be issued.W/891002 Ltr ML18052B5541987-08-24024 August 1987 LER 87-015-00:on 870731,capacitor Failed on 1B Reactor Bldg & Refuel Floor Radiation Monitors Resulting in Safety Feature Actuation.Caused by Normal Aging.Capacitor replaced.W/870825 Ltr 1994-05-11
[Table view] Category:RO)
MONTHYEARML20029E2061994-05-11011 May 1994 LER 93-025-01:on 931205,Loop a MSIVs Exceeded TS Leakage Limit.Caused by Failure of Fastner Locking Devices.Seating Surface of Main Disc in MSIV 1A Machined & Successfully Pass LLRT on 940116.W/940511 Ltr ML17352B2501993-10-0101 October 1993 LER 93-014-00:on 930902,discovered That IRM 11 & APRM Both in Bypassed Condition W/O Half Scram Being Inserted Due to Work Practices.Irm 11 Taken Out of Bypassed Condition & Individual Involved counselled.W/931001 Ltr ML17352B2511993-10-0101 October 1993 LER 93-019-00:on 930903,RHR Pump 1A Inadvertently Started Due to Personnel Error.Pump Immediately Secured & Individual Involved Counselled on Importance of self-check.W/931001 Ltr ML20046A1801993-07-21021 July 1993 LER 92-007-00:on 930624,Toxic Gas Analyzer Inoperable Due to Personnel Error Caused by Lack of Procedural Adherence. Isolated CR Ventilation,Restarted Sample Pump & Restored CR Ventilation Sys to Outside Air suction.W/930721 Ltr ML20045H5781993-07-13013 July 1993 LER 93-013-00:on 930613,internal Electrical Fault within Main Power Transformer Tripped & Caused Reactor Scram & Unexpected Group I Isolation.Caused by Turbine Stop Valve Closure.Transformer replaced.W/930713 Ltr ML20045G7291993-07-0909 July 1993 LER 93-005-00:on 930609,numerous Alarms Received in CR, Including HPCI Turbine Rupture Disc High Pressure Alarm. Caused by Ruptured Disk,Releasing Steam/Water Mixture. Test Will Be Conducted on Rupture disks.W/930709 Ltr ML20045G0951993-07-0202 July 1993 LER 93-011-00:on 930602,HPCI Declared Inoperable in Order to Perform Qcos 2300-13.Caused by MSC Stem & Stem Gear Being Scored & Worn from Age Causing Excessive Friction.C/As Include Disassembling Gearbox Every 10 yrs.W/930702 Ltr ML20045F3321993-06-30030 June 1993 LER 93-012-00:on 930601,U-2 DG Cooling Water Pump Inoperable Due to Inadequate lubrication.U-2 DG Cooling Water Pump replaced.W/930630 Ltr ML20045F3251993-06-28028 June 1993 LER 93-007-01:on 930307,B Loop MSIV Exceeded TS Leakage Limits for Containment Isolation Valves Caused by Damaged Flexitalic Bonnet Gasket.Replaced Bonnet Gasket & Valve Successfully Retested (WR Q061987).W/930622 Ltr ML20044E6171993-05-14014 May 1993 LER 93-006-00:on 930420,Unit 2 Nso Inadvertently Started Unit 1 DG When Attempting to Start Shared (1/2) Dg.Caused by Personnel Error.Unit 1 DG Shut Down & 1/2 DG Started & Loaded successfully.W/930514 Ltr ML20044C9511993-05-0707 May 1993 LER 93-009-00:on 930408,technician Discovered That Estimate of Sample Flow for U2 Reactor Bldg Vent Sampler Flow Rate Monitor Not Calculated.Caused by Personnel Error.Technician counseled.W/930428 Ltr ML20024G6811991-04-19019 April 1991 LER 91-008-00:on 910322,reactor Bldg Ventilation Isolation Occurred.Caused by Lightning Strike.Control Room Vents Reset & Toxic Gas Sample Point a Selected.Addl Trips Associated W/ Lightning Strike Immediately reset.W/910418 Ltr ML20029C1241991-03-0808 March 1991 LER 91-005-01:on 910131,1/2 B Standby Gas Treatment Sys Autostart During RPS B Power Swap Due to an Inadequate Procedure.Nso Reset 1/2 Scram & 1/2 Groups II & III isolations.W/910306 Ltr ML20029B5901991-03-0808 March 1991 LER 91-004-00:on 910211,standby Liquid Control Sys of Unit 1 & 2 Inoperable.Caused by Inadequate Mod Testing.Engineering Dept Evaluating Data & Temporary Procedure 6589 Generated. W/910308 Ltr ML20028H7841991-01-23023 January 1991 LER 90-032-00:on 901224,1/2A Diesel Fire Pump Taken out-of- Svc on 901217 & Not Returned to Svc Before Seven Day Time Allotment Expired.Caused by Mgt Deficiency.Fire Pump Successfully Tested & Returned to svc.W/910123 Ltr ML20028H6831991-01-21021 January 1991 LER 90-034-00:on 901223,high Chlorine Concentration Caused Control Room Ventilation Manual Isolation & ESF Actuation. Caused by Instrument Error Code & Misinterpretation of Analyzer Indication.Flow reduced.W/910121 Ltr ML20028G9141990-09-28028 September 1990 LER 90-018-00:on 900829,plant Outside Design Spec for Electrical Separation Criteria for Two Redundant Safety Sys. Caused by Inadequate Engineering Review During Leads installation.W/900928 Ltr ML20044B2401990-07-12012 July 1990 LER 90-012-00:on 900612,control Room Ventilation Emergency Air Filtration Unit Declared Inoperable.Caused by Heater Malfunction.Work Request initiated.W/900712 Ltr ML20044B0091990-07-11011 July 1990 LER 90-011-00:on 900611,diesel Fire Pump 1/2 a Out of Svc for Period Exceeding 7 Days Reporting Criteria,To Install New Suction Line.Caused by Required Maint Work.Repairs Completed & Pump Tested & Returned to svc.W/900711 Ltr ML20043H4831990-06-21021 June 1990 LER 90-010-00:on 900522,while Returning RWCU Sys to Svc, Nonregenerative HX High Temp Alarm Received,Challenging ESF Logic & Causing Group III Isolation.Caused by Leaking RWCU Check Valves.Work Requests issued.W/900621 Ltr ML20043H2931990-06-18018 June 1990 LER 90-009-00:on 900518,determined That Various Containment Vols Not Leak Rate Tested Due to Recent 10CFR50,App J Interpretation Re Licensing Design Criteria.Mod M4-1(2)-89-167 initiated.W/900618 Ltr ML20043H9521990-06-13013 June 1990 LER 90-016-01:on 900411,motor Control Ctr Relay 28/29-5 Setpoint Drift Occurred,Resulting in Analyzed Plant Condition.Caused by Utilizing Wrong Relay & Inadequate Review of GE Svc Info Ltr.Relay replaced.W/900613 Ltr ML20043E9121990-06-0707 June 1990 LER 90-006-00:on 900508,Unit 2 RCIC Declared Inoperable Due to Unstable Operation of RCIC Pump Flow Controller.Caused by Proportional Band of Controller Being Set to Respond to Changes in Flow Too Quickly.Flow controlled.W/900607 Ltr ML20043F2341990-06-0101 June 1990 LER 90-001-01:on 891208,seven Pathways Not Included in Type B & C Local Leak Rate Test Program.Caused by Misinterpretation of 10CFR50,App J Requirements.Required Vents & Test Taps Will Be installed.W/900601 Ltr ML20043F2161990-06-0101 June 1990 LER 89-024-01:on 891214,main Turbine Unexpectedly Tripped Following Isolation of Reactor Water Level Switch (Lits). Caused by Channel B Lits Having Been Previously Replaced W/ Switch That Operated Reverse function.W/900601 Ltr ML20042E4991990-04-17017 April 1990 LER 90-007-00:on 900318,ESF Actuation Occurred.Caused by Mgt Deficiency in Not Providing Sufficient Guidance for Review of Work Package After Scope Changed.New Work Package Preparation Procedures initiated.W/900417 Ltr ML20012C7231990-03-15015 March 1990 LER 90-004-00:on 900213,loss of Emergency Bus 23-1 Occurred Due to Shorted Conductor Cable While Performing Wiring Verification.Caused by Personnel Error & Improper Installation.Electric Power Supplies recovered.W/900315 Ltr ML20012C7171990-03-15015 March 1990 LER 90-003-00:on 900213,unit Diesel Generator Tripped on Overspeed Upon Manual Startup for Testing.Caused by Misadjustment of Diesel Generator Governor.Diesel Generator Equipment Reset & Governor adjusted.W/900315 Ltr ML20012B2931990-03-0505 March 1990 LER 90-002-00:on 900204,determined That Only Reactor Bldg Differential Pressure of 0.24-inch Water Vacuum Could Be Obtained,In Violation of Tech Spec Required 0.25 Inch.Caused by Testing Deficiency.Leak Paths sealed.W/900305 Ltr ML20012B2961990-03-0505 March 1990 LER 90-003-00:on 900205,determined That Leakage Rate for HPCI Sys Steam Exhaust Check Valve Exceeded Tech Spec Limits.Caused by Excessive Leakage.Check Valve Replaced W/ Mission Check Valve W/Carbon Steel springs.W/900305 Ltr ML20011E1821990-01-30030 January 1990 LER 90-002-00:on 900102,during Annual Water Sprinkler Sys Valve Position Insp,Discovered That Valve 2-4199-072 Not Cycled Per Tech Spec Surveillance Requirement 4.12.Caused by Procedural Deficiency.Procedure to Be revised.W/900130 Ltr ML19354D9431990-01-15015 January 1990 LER 89-025-00:on 891220,determined That Reactor Bldg Overhead Auxiliary Hook May Have Contacted Side of New Fuel Bundle on 891214,causing Extensive Damage.Caused by Personnel Error.Bundle Shipped Back to GE.W/900115 Ltr ML19354D9481990-01-15015 January 1990 LER 89-023-00:on 891219,identified Deficient Temporary Procedures Which Altered Intent of Original Procedure QAP 1100-7.Caused by Mgt Oversight in Preparation of Tech Spec Change.Temporary Procedures revised.W/900115 Ltr ML20005F5881990-01-0808 January 1990 LER 90-001-00:on 891208,seven Pathways Not Included in Type B & C Local Leak Rate Testing Program.Caused by Misinterpretation of 10CFR50,App J Testing Requirements.Mod of Sys to Be Performed as necessary.W/900108 Ltr ML20005G0231990-01-0202 January 1990 LER 88-020-01:on 880619,improper Valving Sequence Occurred Which Resulted in Various ESF Actuations.Caused by Inadequate Equipment out-of-svc Procedure.Procedure Revised & Training Lesson Plan developed.W/900102 Ltr ML20005E4531989-12-28028 December 1989 LER 89-022-00:on 891128,HPCI Declared Inoperable Following Unexpected Actuation of HPCI Pump Room Deluge Sys.Actuation Caused Dc Sys Grounds Due to Moisture Intrusion in Electrical Equipment.Equipment dried.W/891228 Ltr ML20005D9551989-12-22022 December 1989 LER 89-021-00:on 891124,HPCI Deluge Sys Functional Test Not Completed within Tech Spec Interval of 6 Months.Caused by Mgt Deficiency & Insufficient Procedure.Procedures Revised to Better Control Tech Specs.W/891222 Ltr ML19351A6391989-12-15015 December 1989 LER 89-020-00:on 891117,full Reactor Scram Occurred Due to Simultaneous Loss of Power to 1A & 1B 24/48-volt Distribution Panels During Panel Transfer.Caused by Personnel Error.Personnel Trained Re event.W/891215 Ltr ML19332E7581989-12-0505 December 1989 LER 89-019-00:on 891105,half Group II & III Isolation Occurred While Performing Valve Flow Check Surveillance. Caused by Leaky Isolation Valve on Transmitter.Surveillance Methods Will Be Reviewed & Procedures revised.W/891205 Ltr ML19332C7581989-11-20020 November 1989 LER 89-017-00:on 891022,discovered That 6-month Functional Test Procedure Qos 500-3 Re Functional Test for Reactor Protection Sys Electrical Assemblies Not Completed in Required Time.Caused by Mgt deficiency.W/891120 Ltr ML19327C0701989-11-0909 November 1989 LER 89-018-00:on 891012,notification Given of Potential Single Failure of Diesel Generator Voltage Regulator That Could Result in Loss of All But One ECCS Loop.Caused by Design Deficiency.New Procedure developed.W/891109 Ltr ML19327C0951989-11-0404 November 1989 LER 89-005-00:on 891012,reactor Scram Occurred Due to Turbine Stop Valve Closure.Caused by Personnel Error in That Work Analyst Overlooked Removal of Two Connections on Limit Switch.Maint Workers Immediately counseled.W/891104 Ltr ML19327C2631989-11-0101 November 1989 LER 87-009-01:on 870801,reactor Scram Occurred from Turbine Generator Load Mismatch Due to Generator Trip.Caused by Electrical Fault in Main Transformer C Phase Windings.Main Transformer Replaced w/spare.W/891103 Ltr ML19325E8261989-10-30030 October 1989 LER 87-012-01:on 870918,while Performing Qos 1600-1, Vacuum Breaker 2-1601-33E Remained Open After Testing.Caused by Binding in Test Cylinder Portion of Vacuum Breaker. Nuclear Work Requests completed.W/891030 Ltr ML19325E5921989-10-24024 October 1989 LER 87-017-01:on 870805,Group IV Isolation Received Which Resulted in Closure of HPCI Steam Supply Valves.Caused by Failed HPCI Steamline Differential Pressure Transmitter Due to Loss of Oil in Sensing cell.W/891024 Ltr ML19325E5971989-10-24024 October 1989 LER 88-003-01:on 880301,while Performing RCIC Monthly Test, RCIC Pump Could Only Achieve 400 Gpm Against Discharge Pressure of 500 Psig.Caused by Failed Hydraulic Actuator on Turbine Governor Valve.Actuator replaced.W/891024 Ltr ML19327B7901989-10-18018 October 1989 LER 89-016-00:on 890921,during Transfer of New Fuel,Fuel Assembly Released from Refueling Grapple & Fell Upon Spent Fuel Racks.Caused by Personnel Error & Procedural Deficiency.Fuel Handling Procedures Revised ML19327B0491989-10-16016 October 1989 LER 89-015-00:on 890916,during Transfer,Loss of Power to Off Gas Monitor a Caused Contacts to Open Giving Upscale Radiation Signal,Thus Starting Off Gas Timer.Caused by Inadequate Procedures.Off Gas Sys reset.W/891016 Ltr ML19325C5341989-10-0202 October 1989 LER 89-014-00:on 890910,determined That Combined Leakage Rate from All Penetrations & Valves Exceeded Tech Spec Limit.Root Cause Unknown.No Corrective Action Taken. Supplemental Rept Will Be issued.W/891002 Ltr ML18052B5541987-08-24024 August 1987 LER 87-015-00:on 870731,capacitor Failed on 1B Reactor Bldg & Refuel Floor Radiation Monitors Resulting in Safety Feature Actuation.Caused by Normal Aging.Capacitor replaced.W/870825 Ltr 1994-05-11
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data SVP-99-204, Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 SVP-99-179, Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs SVP-99-155, Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-148, Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety SVP-99-123, Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations SVP-99-104, Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-102, Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with1999-04-30030 April 1999 Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-071, Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20205C5671999-03-19019 March 1999 Simulator Four-Yr Certification Rept ML20207D2341999-03-0101 March 1999 Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205 ML20204B1571999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.With SVP-99-021, Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With1999-01-31031 January 1999 Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With ML20205D1311998-12-31031 December 1998 1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with SVP-99-007, Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With ML20196C8391998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-030-2, Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1 SVP-98-364, Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196G1241998-11-30030 November 1998 COLR for Quad Cities Unit 1 Cycle 16 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196C8731998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-30-1, Fracture Mechanics Evaluation on Observed Indications at Two Welds in Recirculation Piping of Quad Cities,Unit 1 Station ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB SVP-98-346, Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-98-358, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With1998-10-31031 October 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With SVP-98-326, Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20153D0191998-09-18018 September 1998 Part 21 Rept Re Defect in Gap Conductance Analyses for co- Resident BWR Fuel.Initially Reported on 980917.Corrective Analyses Performed Demonstrating That Current Operating Limits Bounding from BOC to Cycle Exposure of 8 Gwd/Mtu ML20153C6771998-09-17017 September 1998 Part 21 Rept Re Defect Relative to MCPR Operating Limits as Impacted by Gap Conductance of co-resident BWR Fuel at Facilities.Operating Limit for LaSalle Unit 2 & Quad Cities Unit 2 Will Be Revised as Listed ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20151Y7261998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Quad Cities Nuclear Power Station ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20151Y7301998-07-31031 July 1998 Revised MOR for Jul 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20237A6251998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Quad Cities Nuclear Power Station,Unit 1 & 2 SVP-98-328, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With1998-07-15015 July 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With SVP-98-249, Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-98-215, Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 2 ML20247N6281998-05-19019 May 1998 Rev 2 to COLR for Quad Cities Unit 2 Cycle 15 ML20216C0561998-04-30030 April 1998 Safe Shutdown Rept for Quad Cities Station,Units 1 & 2, Vols 1 & 2.W/22 Oversize Figures SVP-98-176, Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20217D0281998-04-22022 April 1998 Part 21 Rept Re Additive Constants Used in MCPR Determination for Siemens ATRIUM-9B Fuel by Core Monitoring Sys Were Found to Be non-conservative.SPC Personnel Notified All Customers w/ATRIUM-9B Lead Test Assemblies ML20217G3951998-04-0808 April 1998 TS 3/4.8.F Snubber Functional Testing Scope Quad Cities Unit 2 TS (Safety-Related) Snubber Population 129 Snubbers SVP-98-128, Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 2 1999-09-30
[Table view] |
Text
NRC FORM 364 U.S. NUCLEAR REGULATORY COMMISSION (7.nl UPDATE' REPORT -
- PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONTROL 8 LOCK: l l l l l l lh (PLEASE PRINT CR TYPE ALL REQUIRED INFORMATioN) 1 6
@ l8l 9l L lLICENSEE 0 l ACODE l 0 l 142 [h o l 0 l 0 luCENsE 15
- l0 NUM8ER l0 l0 l- ln ln ln l@lLl1l1l1l1l@l 25 26 LICENSE TYPE JO l
Si GAi 56 l@
CON'T ITTT1 " RM l 80L ]@l410 l 5 l DOCKET 0 l 0 NUM8ERlo l 2 l6 1568 ()! 691 l 1EVENT l 2 lQATE 61719 l@l750 l 310 l 4 l 81080l@
7 8 74 REPORT DATE EVENT DESCRIPTION AND PROSA8LE CONSEQUENCES h gl While performing the Main Steam isolation Valve local leak rate test, valves I gl A0 2-203-1A, A0 2-203-2A, A0 2-203-18, and A0 2-203-2B were found to have leak rates 1 fgy-] l in excess of the 11.5 SCFH allowable limit as specified by Technical Specification i gl 4.7.A.2.i.c. The excessive leak rate did not render the MSIV's inoperable. The l 0 6 l to ta l leak rate possible through the A and B steam lines would have been 30.0 SCFH l
, f3T 1 I and 24.2 SCFH respectively had a steam line break occurred. The A0 2-203-1C, 2C, 10, l 1018: I and 2D MSIV's had satisfactory leak rate results. l E CODE sV8C E COMPONENT CODE SU8 0'E s E lOl91 l C l D l@ y@ l Bl@ l V l A l L lV lE lX l@ y@ W @ 20 7 8 9 to 11 12 13 18 19 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RO & VENT YEAR REPORT NO. COO 8 TYPE N O.
@ REP l 7l 9l 21 22 l_l 23 l 0l 2l 7l 24 26 l/l 27 l0 l3 l 28 29 y
30
[__.]
31 l1 l 32 Y N AC O O P NT ME HOURS s8 i POR 8. sVP LI 'JANL8 C RER l23Bl@l.Zl@ l35 Zl@ . l36Zl@ l Ol 0l Ol 0l (,Y_,j @ l42Y l@ W@ lCl6l6l5l@
04 JJ 40 41 43 44 47 CAUSE GESCRIPTION AND CORRECTIVE ACTIONS h ,
13 l 0 l l The A0 2-203-IA, 2A, and IB valves were found to have roughened surfaces on the main 1 yl seats. A slight amount of erosion was also found on the pilot disc of the A0 2-203-1A1
, , l valve. The stroke was extended to repair the A0 2-203-28 valve. The other three l l . valves were relapped and a new pilot disc was installed on the A0 2-203-IA valve. l
, 4 l The combined corrected leak rate for the 2-203-1A and 2A valves was 1.2 SCFH and for l 7 8 80 SJye 2-203-1B and 2B valves was 3A SCgf,ggo, l status sPOwER OTHER status (3/ oisCOvERv oisCOvERY oesCRiPT:ON @
y ] @ l0l0l0l@l '
NA l lBl@l Local Leak Rate Testing l A$TiviTY CO0 TENT RE t P ,RglE A AMOUNTOP CTIVITY @ LOCATION OP ELEAsE@
PERsO~ net exposures NUMSER TYPE DESCRIPTION 1 7 l 0l 0 l 0 l@l Z l@l _ NA l
' ' ' ,ERsONNELiNauLs "
NuveER oEsCRiPTiON@ ,
NA 7
i a 8 9 10l0lOl@l 11 12 l
80 LOSS OP OR DAMAGE TO FACILITY . )
TYPE DESCRIPTION NA 7
9 8 9
@@l 10 80 l
NLY
, 'Y8 tic:7 ,
2 0 _j@9El CRIPTION NA g lllllllllllllj l 7 E 10 64 69 80..
S. Kan 309-65h-2241 Ext. 174 o NAME OF PREPARER PHONE: $
- 1. LER NUMBER - LER/R0 79-27/03L-1 II. LICENSEE NAME: Commonwealth Edison Company i ll. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:
On November 26, 1979, local leak rate testing of the Main Steam isolation valves, procedure QTS 100-3, revealed that valves A0 2-203-l A, A0 2-203-2A, A0 2-203-IB and A0 2-203-28 had excessive leakage. The leak rates are 30.05 SCFH, 32 3 SCFH, 47 2 SCFH and 24.2 SCFH, respectively. These leak rates exceed the 11.5 SCFH allowable limit as specified by Technical Specification 4.7.A.2.i.c. All other MSIV's were leak tested satisfactorily. On November 27, 1979, the stroke of 2B valve was extended and the leak rate was found to be 8.1 SCFH, which was within the limit.
VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:
In the event of a steam line break outside the primary containment, the total leakage possible through the "A" and "B" steam lines would be 30.0 SCFH and 24.25 SCFH respectively. The excessive leakage rate did not in.any way render the MSIV's inoperable.
Upon receipt of a Group i isolation signal, these valves would have closed in the required times and performed the isolation function. Therefore, the safety implications of this occurrence are minimal.
Vll. CAUSE:
Upon disassembly of the A0 2-203-IA main steam isolation valve, an uneven surface on the main seat was observed. A slight amount of erosion was' also found on the pilot disc.
The A0 2-203-1B and 2A main steam isolation valves were also found to have roughened surfaces on the main seats.
Vill. CORRECTIVE ACTION:
Main steam isolation valve A0 2-203-1A was disassembled; a new pilot disc was installed, and the main seat was relapped.
Valves A0 2-203-1B and 2A were also relapped to eliminate the uneven surfaces on the main seats. After reassembly, valves A0 2-203-1 A, IB, and 2A were retested and found to have a i satisfactory leak rate of 1.2 SCFH for "A" steam line and 3 5 SCFH for "B" steam line. These valves are manufactured by Crane Company.
9 g&.
r
E' @o'frE REPORT- -
' PREVIOUS REPORT DATE: 11-24-J7 LICENSEE EVENT REPORT CONTROL BLOCK: l 1
l l l l l lh 6
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) lTJI)8 l9 l l LLICENSEE 7
l Q l CODE A l D l 2 l@l0 14 15 l0 l0 l-LICENSE l0 l0NvMsER l0 l- l0l0l0l@l4l1l1l1l1l@l 2$ 26 L& CENSE TYPE JG l
bl LATbd l@
CON'T '
l0lil g l L j@l 015 l0 l0 l0 l2 l6 l5 l@l 1 l 1 l 2 l 6 l 7 l 9 l@l 0 l 3 l 0 l 4 l 8 l o80l@ F4 73
? 8 60 61 DOCKET NUMBER ea8 69 EVENT DATE REPOR T O ATE EVENT DESCRIPTION AND PROSA8LE CONSEQUENCES h
[TTT1 I While performing the main steam line drain local leak rate test, valve M0 2-220-2 was l ITTTII f und to have a leak rate of 26.0 SCFH. This was in excess of the 18.36 SCFH I
[TT q l allowable limit as specified by Technical Specification 4.7.A.2.i. The in-line I i o ,5 ; ; redundant isolation valve, M0 2-220-1, was found to have an acceptable leak rate valuel o 8 l Subsequently the ability to isolate the main steam line drain was not adversely I gl affected. l I O ls t i I E COOE SU C E COMPONENT CODE SUSCODE SU E
[O_L9J 7 8 l C i o ig gg l BlQ lV l A [ L lV lE lX lQ lE lg ]g 9 to 11 12 13 18 19 20 SEOUENTI AL OCCURRENCE REPORT REVISION
,.,,, EVENT YEAR REPORT NO. CODE TYPE N O.
h "LER/RO 3u'Mg l217 22l 0l
[---j 23 l0l2l7l 24 26 l/l 27 l0 l3 l lLl 30 l_l lI l 28 29 31 J2 K N AC PLANT MET HOURS 22 58 17 FOR b 8. St.,PPL ER MAN l33 B l g ]34g lS 3
Z lg l36Zlg l0l0l0l J/
l 40 gg 48 lY [g lN lg 42 43 lC l$FACdURER[
44 l
47 CAUSE DESCRIPTION AND CCRRECTIVE ACTIONS h lil0ll A small amount of steam cutting on the valve seating surface was determined to be l gg the cause of this occurrence. The valve seating surface was lapped and a second l l
,,,,,g leak rate test was performed. The corrected leak rate was 8.45 StrH.
]
i 3 l l )
l i 4 l l 7 s 9 s0 ST % POWER OTHER STATUS Of RY DISCOVERY DESCRIPTION li Is l W@ l 0l 0l 0l@l NA l lBl@l Local Leak Rate Test l ACTIVITY CONTENT RELEASED OF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELEASE M [_Z]
7 89_
@ lZjgl 10 tt NA 44 l l 45 NA 80 l
PERSONNEL EXPOSURES NUM8ER TYPE DESCRIPTION i T NA 1010 l 01@l Z 1@l l
,iRSON~E t iN2u'!,ES NUM9ER DESCRIPTION i R NA 7 8 9 I 01010 l@l 11 12 80 l
LOSS TY OF OR DAMAGE TO FACILITY Q DESCRIPTION N./
"A 8 LZ'E]@l 7 8 9 to 60 I
NRC USE ONLY ,
IS DESCRIPTION L3_LOJ[3S@l 8 9 to NA I
68 69 I I I I I I I I I i i i li 7
30 7.
J. Kopacz 109-6 4-2241 ext. I7A NAME OF PREPARER PHON E: {
% 6
- 1. LER NUMBER: LER/R0 79-27/03L-1 II. LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station lit. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:
On November 25, 1979 Unit Two was shutdown for a refueling and maintenance outage. While local leak rate testing the main steam drain line valves on November 26, the through-seat leakage of valve M0-2-220-2 was found to be excessive. The as found leak rate was 26.0 SCFH which exceeded the Technical Specification Appendix J limit of 18.36 SCFH. A pressure decay method was used to determine this leak rate using procedure QTS 100-1. This valve is manufactured by the Crane Valve Company.
VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:
The main steam line drain valves are part of Group One isolation protection system, which is designed to minimize leakage of steam from the drywell should a main steam line rupture.
Because the two valves in series isolate the main steam line drain, the leakage from the drywell would have remained minimal.
The redundant MO 2-220-1 valve had an as-found leak rate of 1.72 SCFH. Further, these valves are usually closed during power operation and are opened during startup and shutdown to remove condensation from the main steam lines.
Vll. CAUSE:
Upon disassembly of the M0 2-220-2 valve, the seating surface showed minor signs of steam cutting.
Vill. CORRECTIVE ACTION:
The seating surface of the valve was lapped. The stem packing was also replaced as part of routine maintenance. The final combined leak rate measurement was 8.45 SCFH. This valve has had a seat repair cnce previously.
4
4 NRC FORM ESS U.S. NUCLEAR REGULATORY CoMMisSl2N
, (7 77) ' '
- LICENSEE EVENT REPORT (PLEASE PRINT OR TYPE ALL REQUIRED INFoRMATION)
CONTROL BLOCK: l I
l l l l l lh 6
['o'Til
) 8 9 I i l t LICENSEE I Q l ACODE l D l 214 l@l0 15 l0 l 0 l - lLICENSE 0 l 0NUM8ER l 0 l - l 0 l 0 l 025 l@l26 4 l LICENSE 1 l 1 ll l l1 l@l SFCAT$4 TYPE JO lg CON'T loIiI g lL l@ Q l5 l0 l0 l0 l2 l 6 l 5 ]@l 0 l 2 l 0 l 4 l 8 l 0 l@l0 l 3 l0 l 4 l 8 1080l@ 74 7 8 80 61 DOCKET NUM8ER the 69 EVENT OATE 75 REPORT DATE CVENT OESCRIPTION AND PRQ8ABLE CONSEQUENCES h lTTT1 l While performing Local Leak Rate Test, procedure QTS 100-/% the volume bounded by l
[TF51 l valves MO-2-1001-34B, MO-2-1001-363, and M0-2-1001-378 was found to leak 182 SCFH, l ITTT1 I which was in excess of the allowable limit for any one valve as specified in Section l o 5 l 4.7.A.2.i of the Technical Specifications. Further investigation revealed that the l
[ypig-} l major part of the leakage was through valve M0-2-1001-368. The upstream M0-2-1001-34B l
@l valve successfully passed the local leak rate test. l 10181 l l 7 8 9 40 C E CODE SUSC E COMPONENT CODE SUSC00'E SU E M
7 8 lC lF lg ]g l Blg lV l A lL l Vl Ej Xlg l F lg lG lg 9 to 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RQ EVENT YEAR REPORT NO. COOE TYPE N O.
@ RE g l217l229l l_l 23 l0l2l7l 24 26 jf] l 0l 3 [
28 3 lLl 30
.l _l 31 l0l J2
_ 27 N A O O P NT ME wouRS O22 *?85>MP s P3=8. u
"s'"PPL' er
' ' ~'~
MA~u"PACToa' R lJJB lglJ4 Zl@ [ Zj g 35
[36Zj@
, l0l0l0l0l 37 40 lY l@
el lY [g l Nlg 42 43 l44C l 6 l 6 l 547l@
CAUSE OESCRIPTION ANO CORRECTIVE ACTIONS h li l0 l l The valve disc was found to be slightly warped and corrosion was found on the valve l gl seat. The disc guide pin was also bent. The valve was taken apart and the disc and l
, , l seat were cleaned and remachined to provide proper fit. The guide pin was straightenecj
, 3 l and the torque switch on the limitorque operator was adjusted. The volume was l
, , l retested and the corrected leak rate was 3.0 SCFH. l 7 8 9 Bo STA S % POWE R OTHER STATUS Ott O RY DISCOVERY DESCRIPTION
@ W@ l0 l0 l0 l@l NA l W @l Local Leak Rate Test l A TIVITY CO TENT AMOUNT OF ACTIVITY LOCATION OP RELEASE s 7 1 6 8 9 W
RELEASE 10
@D OP RELE ASEygl 11 NA 44 l l 45 NA 80 l
PERSONNEL EXPO $yRES 5 NUMBER TYPE DESCRIPTION 39 y l0l0l0l@lZl@l NA l
' ' ' ,ERsON~E't' iN;u'J Es
- NUM8ER D E SCRi* TION Q l0l0l0l@l 7 8 9 11 12 NA 80 l
WPE E s'C"'?T O.? '" @
j Q l Zl@l 7 8 9 10 NA 40 l
E Lf81 iss DESCRIPTION
- NA l lllllllllllllj 7 8 9 10 68 69 80*.
T. Hafera 309-654-2241 ext. 176 o NAME OF PnEPARER PHONE: $
E
- 1. LER NUMBER: R0-79-27/04L-0
- 11. LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station Ill. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:
On February 4, 1980, while performing Local Leak Rate Test, procedure QTS 100-12, the volume bounded by valves M0-2-1001-34B, MO-2-1001-36B, and M0-2-1001-37B was found to leak in excess of the allowable limit for any one valve as specified in Section 4.7.A.2.1 of the Technical Specifications. The leak rate was found to be approximately 182 SCFH.
To determine which valves were leaking, each valve was tightened down manually using the hand wheel while the test volume remained pressurized. From this i t was found that manually tightening the M0-2-1001-368 valve reduced the leakage to an allowable limit. However, because the valve needed to be tightened excessively to reduce the leakage, it was concluded that the valve would not seat properly even under the maximum torque that could be supplied by the valv- motor. Work Request No. QO3352 was issued to repair the valve.
VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:
Because the redundant RHR "A" containment cooling loop was fully operable at all times, and the associated valves were found to have acceptable leakage, the overall safety of the system was never affected. The ability of the RHR system to provide adequate reactor and containment cooling during an accident as designed without the use of one of the cooling loops and other additional components has been documented in the FSAR in section 5.2 3 3 Because the MO-2-1001-34B valve located upstream of the M0-1001-36B valve had an apparent acceptable leak rate, proper line isolation would have occurred if necessary.
V11. CAUSE:
The cause of the failure of the MO-2-1001-36B valve has been designated as equipment failure. The valve disc was found to be slightly warped and corrosion was found on the valve seat.
The disc guide pin was also bent. These factors contributed to'the inability of the valve to sest properly. The M0 1001-36B valve is a 14 inch, 390 pound, motor operated gl6be type valve that is manufactured by Crane Company.
Vill. CORRECTIVE ACTION:
The valve was taken apart-and the valve disc and seat were cleaned and remachined to provide a proper fit. The guide pin was removed., straightened and rewelded into place in the valve. The torque switch on the limitorque operator was adjusted and the valve was retested. The final leak rate was found to be 3 0 SCFH which is less than the allowable 18.36 SCFH leak rate in the Technical Specifications.
There have been two previous occurrences in 1976 and 1977 related to this voume, but were both related to the M0-1-1001-37A valve.
4 I
e p
6 9
L_:
5"n*i . M E. REPORT- .
' PREVIOUS REPORT DATE 12-24-79 LICENSEE EVENT REPORT CONTROL 8 LOCK: l i
l l l l l lh (PLEASE PRINT 09 TYPE ALL REQUIRED INFORMATIONI Io t i l i I l ' l 4 l ^ I D l 2 lgl 0 l0 l0 l -l 0 l 0 l 0 l - l 0 l 0 l 025lgl 4 l 1 l 1 l 1 l 1 lgl57 CAT 14 s.lCENst NUMeER 26 LICENSE TYPE JC l
68 lg 7 8 9 LICENSEE CODE 15 CON'T a
lol,l ,g l L gl J0 l 5 l0 l0 l0 l2 l6 l5 lgl 1 l 169 l 2 l 6 l 7 l 9 l@l 0ISl 3 lREPORT 74 0l4 l 8 l080l@
OATE 7 8 60 64 OOCKET NUMBER t,d EVENT QATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h
[TTT1l While performing the RCIC turbine steam supply valve local leak rate test, valve l ITTTIl MO 2-1301-16 was found to have a leak rate of 76.7 SCFH. This was in excess of the l ITT~i l i 18 36 SCFH allowable limit as specified by Technical Specification 4.7. A.2.i. [
gl Redundant in-line valve, M0 2-1301-17, was found to have a leak rate of 5.6 SCFH, l
[g 7;;-) l which was within acceptable leakage limits. Thus, the steam line. leakage was minimal.l ITITl I l 10181l l 7 8 9 80
- CE'E "^EoY C sis 8CooE CouPoNENT CooE su'8CI6E sENNE -
g 7 8 lClEl@ ] @ ] @ l Vl Al LlV lE lX l@ ]@ W @
9 10 11 12 13 is 19 20 SEQUENTIAL OCCURRENCE REPORT REVislON EVENT YEAR REPORT NO. CODE TYPE NO.
h y*LER/RO u'Mg [---J l0 l2l7l l/l l 0l 31 lLl [---J l1 l
. l21 71 22 9l 23 24 26 27 28 29 30 J1 32 J
.N AC ON ON Pt NT MET HOURS 22 s8 IT FOH 8. sVP Li R MANUFACTURER I Al@l zlg 33 34 l6z lg .l36 zig
.3 l
40
[NJg lY lg lN lg lC l6 l6 l5 l@
44 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h i o l The valve body seat was found to be out of alignment such that the valve seating l 3 i l surfaces did not match up. When lapping of the valve seats did not prove successful, l t 2 l the valve was replaced. The newly installed M0 2-1301-16 valve was tested and the l l l
W, 3 l . corrected leak rate was 0.9 SCFH. -
l l
l U_Li] I I !
7 8 9 80 l
's*TANIs* sPowER oTHERsTATus h o"isERv' olsCoVERY DESCRIPTION i s
]@ l 0] 0l 0l@l_" NA l lBl@l** Local Leak Rate Test l
' ' ^^ **
AEnvirY Co'0 TENT RELEASED OF RELEASE AMOUNT QF ACTIVITY LOCATION OF RELEASE 1 6 ] @ l zlgl NA l l NA l 7 8 9 10 tt 44 45 80 PERSONNEL EXPosVREs NUM8ER TYPE DESCRIPTION "A
E 10101 Ol@LZJ@l
,ERso EL imuLs NUVSER DESCRIPTION NA 7
i R 8 9 Iof01Ol@l il 12 80 l
Wi'E
C'"@
1 9 glEs'C7^?To'N NA l l 7 8 9 10 80
- ss DESCRIPTION 2 o [_jhl NA l llllllltllllljI I 8 9 10 68 69 80* 5 1 NAME OF PREPARER
. enwald PHONE:
NRC FCCM 366 U. S. NUCLEAR REGULATORY COMMISSl!N (7 77)
- REPORT-PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONTROL SLOCX: l 1
l l l l l lh 6
(PLEASE PRINT OR TYPE ALL REOUIRED INFORMATION) l'o IIl l l l L l 0 l A l 0 l 2 l@l 0 l0 l 0l -l 0l 0 ] O l - l 0 l 0 l260 l@l 26 4LICEr l 15E;TYPg l1l1l@l JO l
57 CAT SS l@
7 8 9 LICEN34E CODE 14 15 LsCENSE NUM8ER CON'T IOI11 3 U,$ l L@l 0 l 5 !0 l0 10 l2 l6 15 l@l 1121015 l 7 l 9 l@l 74 075 l 3 l 0 l 4 l 8 l 080l@
7 8 80 61 COCKET NUMSER 68 69 EVENT OATE REPORT DATE EVENT DESCRIPTION AND PROBA8LE CONSEQUENCES h
< ITTTI I A local leak rate test of the drywell-suooression chamber vent exhaust valves was l ITTTll performed in accordance with procedure QTS 100-30. The as-found leak rate from the l gl volume bounded by valves A0 2-1601-23, 24, 60, 61, 62, and 63 was found to be 27.0 I iO ;s; ; SCFH. This was in excess of the 18.36 SCFH allowable limit for any one valve as I lo is i i specified by Technical Specification 4.7.A.2.i. I f0TIl l I, 10181 I l 7 8 9 88 E ODE S 8C E COMPONENT CODE su8 o'E SU E g lSlDl@ ] @ IB l@ l V l A l L lV lE lX l@ l Bl@ l Ll@ 20
, 7 8 9 10 11 12 13 18 19 SEOUENTIAL OCCURRENCE REPORT REul5ICN LE R/R0 EVENT YEAR REPORT NO. CODE TYPE .
@ A4P l 7l 9l
_ 21 22 l _._ l 23 l0l2l7l 24 26 jfl 27 l0 l3 l 28 29 lLl 3d jl 31 l
32 l
TKr AC 0 ON PL NT ME HOURS ~ S8 IT FO 8. S PPLIE MANUPA RER
. 33l 8 lgl34Zl@ l3S Zl@ l36 Zl@ l0l0l0l0l l~Y l@ l42Yl@ lN l@ lP l3 l4 l0 47l@
37 40 41 4 4 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h li l0 l l The valve disc on the A0 2-1601-23 valve was found to be off-center. The disc was I i
gl re-centered. In addition, the valve seats on A0 2-1601-61 and 62 were lapOed. After I i ,12 I I these repairs the volume bounded by valves A0 2-1601-23, 24, 60, 61, 62, and 63 was I gl retested and the combined corrected leak rate was 18.0 SCFH. l i 4 l l 7 8 9 80
$A % POWER OTHER STATUS IS RY Ol500VERY DESCRIPTION g ] @ l0l0l0lgl NA l lBlgl Local Leak Rate Test l A TlVITV CO TENT RELEASED OF RELEASE AMOUNT OF ACTivtTY LOCATION OF RELEASE N4 NA ITTil8 9W @ IZ l@l 7 10 13 44 l l 45 80 l
PERSONNEL EXPO $URES NUMBER TYPE DESCRIPTION Q l 0] 0 l 0 l@l Z j@l NA l PERSONNE L INJUPIES NUMBER DESCRIPTION y l0l0l0l@l NA l 7 8 9 11 12 80
' "^
WS* PE V^C ' ' '' " @
,1 O~
"A ITT18 L2J@l 1 9 10 80 I
D
- ISS E L38l ESCRIPTION NA l lllllllllll((~
7 8 9 10 68 69 80 Z NAME OF PREPARER PHONE:
- l. LER NUMBER: LER/RO-79-27/03L-1
- 11. _ LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station Ill. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:
On December 5, 1979, a local leak rate test of valves A0 2-1601-23, 24, 60, 61, 62, and 63, revealed an excessive leak rate of 27 SCFH. Technical Specification 4.7.A.2.! requires a leak rate for any one isolation valve of less than 18.36 SCFH.
These valves are located on the drywell-suppression chamber vent system.
VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:
During testing of this drywell-suppressi-a chamber vent volume, it was determined that the major tior of the leakage was attributable to the A0 2-1601-23 valve. Since A0 2-1601-63 and A0 2-1601-24 did not exhibit signs of excessive leakage, and are the second isolation valves in the vent lines, the total leakage through.these vent lines was less than 18.36 SCFH, and safe plant operation was not affected by this occurrence.
Vll. CAUSE:
The cause of this occurrence was component failure. When disassembled the A0 2-1601-23 valve disc was fcond to be off-center. The A0 2-1601-23 valve is an 18 inch butterfly valve manufactured by the Henry Pratt Co. The A0 2-1601-61 and 62 valves are 2 inch globe valves manufactured by the Crane Company.
Vill. CORRECTIVE ACTION:
The disc for A0 2-1601-23 was recentered. The valve seats on A0 2-1601-61 and 62 were lapped. After these repairs, a successful local leak rate test was performed. The measured leak rate was 18.0 SCFH.
4
" #" "' #"" * "U" **' "
""I[0DITEREPORT- -
' PREVIOUS REPORT DATE 12-24-79 LICENSEE EVENT REPORT (PLEASE PRINT OR TYPE ALL REQUIRED INFoRMATION)
CONTROt. BLOCK: l 1
l l l l l lh 6
l0 l1l8 ll9 l L UCENSEE 7
l Q l ACODE l0 l 214l@l150 l 0 l 0 l - uCENSE l 0 l 0NUM8ERl O l - l 0 l 0 l250 l@l 26 4LICENSE l I l I TYPE ll l1JOl@l67 CATl 58 l@
CON'T "EPO
-lOlil 7 8
,, R l60L@l 0l 5l 0lOOCKET 61 0 ] 0l 2l 615 l@l NUMg gR 68 1
69 l 1 EVENT l 2 lOATE6 l 719 l@l013 74 75 l0 l 4 l 8 l080l@
REPORT OATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h ITTTl I While performing local leak rate testing on the A feedwater line. the volumes I
- O ,3, j required to test check valves CV 2-220-SdA and CV 2-220-62A could not be pressurized. l
[T77l l On November 27. 1979 leak rate tests were performed on the B feedwater line with l ITTTl I resultant leakages of 0 SCFH for check valve CV 2-220-588 and 406.8 SCFH for check l lO ls 1 l valve CV 2-220-62B. (Technical Specification 4.7. A.2.1) l IO l 71 l I F5171 [__ f 7 8 9 80 E CODE SUSC E COMPONENT CODE SUSCOO'E SU E FOTi'l 7 8 IC IH l@ W@ l B l@ IV llaA lL IV lE IX l @ l C l @ [_0_l 9 to 11 12 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LE R/RO EVENT YE A R REPOR T NO. CODE TYPE N O.
@ ,afg u l217l 9l l-l l0l2l7l 24
_1/l l 01 3 l 28
[L_l l-,-j l1l 32 22 23 26 27 23 .30 31 KN AC O ONP NT M HOURS 22 S8 iT FOR1 8. SUPPLIE MANUFACTURER I22 Bl@l24 Zl@ d@
35 lasZ l@ l 0 l 0 l .0 l 0 l 2i 7 40
[ Nljg 4i ,
l Y lg 42 l42N l@ lC 44 l6 l6 15l@
47 "CAUSE DESCRIPTION AND CORRECTIVE ACTIONS i O l Corrosion on the valve seats and worn Kairaz 0-rings were found to be the cause. I i : l The valves were cleaned, reworked, and new Kalraz 0-rings were installed. T5e valves I
, 2 l were retested and the corrected leak rates were 7.8 SCFH for CV 2-220-58A, 0.0 SCFH l gl for CV 2-220-62A, and 14.9 SCFH for CV 2-220-62B. l i 4 l l 7 8 9 80 ST $ % POWER OTHER STATUS ISCO RY OtSCOVERY OESCRIPTION y W g l0l0l0lgl NA l lB[gl Local Leak Rate Test l l ACTIVITY CO TENT RELE ASED OP RELEASE AMOUNT OF ACTIVITY LOCATION OF RELEASE ,
7 i 6 8 9 l Zl@ to Ql 1:
NA 44 l l 45 NA 80 l l PERSONNEL EXPOSURES NUM8ER TYPE CESCRIPTION LLLII I I o l@l Z l@l NA l
,ERSON~El'iN; UKES
~uM s A DESCRi,TiO~@
- g l0l0l0l@l NA l 7 8 9 11 12 80
' ^"^
'#5S PE PT ON ' " " @ l 8 I Z l@LSC 7 8 9 l
10 NA 80 I ;
1 NRC USE ONLY ,
RIPR ,
2 O L3@Ol NA l lllllllllllllj 7 8 9 10 68 69 80 5 l J. Hoeller 309-654-2241 ext 171 o !
NAME OF PREPARER PHONE: $ l I
NRC FORM && U.S. NUCLEAR REGULATORY COMMISSION n.n.: . UPDATE
- REPORT-j- PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONT.70L 8 LOCK: l 1
l l l l l lh 6
(PLEASE P*.lNT CR TYPE ALL REQUIRED INFoRMATsON) l I68l@
l
[ 10111
? 8 9 l l lUCENSEE L l QCOOEl A l 0 l 2 l@15 0luCENSE 14 0l 0l -l 0l 0l 0 l - l 0 l 0LICENSE NuMeER 45 26 l 0 l@l TYPE JJ4 l 57 1CAT l 1 l 1 l@
CON'T 101i1 "*'
S uRg l L l@l 0 l 5 l0 l0 l0 l2 l6 l 5 l@l 1 l 210 l 3 l 719 l@l 0 l 310 69 74 l 4 l 81080 l@
REPORT DATE 7 6 60 61 GoCKET NUMBER 68 EVENT DATE 7%
EVENT DESCRIPTION ANO PROBABLE CONSEQUENCES h 1032; l While performing local leak rate testing, the containment purge volume bounded by )
yl valves A0 2-1601-21, 22, 55, and 56 was found to have a leak rate of 135.1 SCFH. Thist
- g o 3, g l was in excess of the 18.36 SCFH allowable limit for any one valve as specified by J o 3 g Technical Speci fica tion 4.7. A.2. i . The safety implications of this event were l gl minimal since valves A0 2-160-22, 55, and 56 were within acceptable leakage limits. l FaTi1 1 The volume would have isolated properly when needed. l I
Lo_L8.] I 88 y a9 E oOE SUSC E COMPONENT CODE SU8 CODE SU E ITTTl 7 8 l S l A l@ d@ l Bl@ l V l A l L lV l E lX l@ l Bl@ 20W @
9 10 13 12 13 14 19
,,, SEQUENTIAL OCCU R R ENCE REPORT REVISION EVENT YE AR REPORT NO. CCOE TYPE N O.
LER
@ ,ag/RO l 7l 9l [---J l 0l 2l 7l l/l l 01 31
- 8
[_L 30_,
J l-l 31 l 1l 32 21 22 23 24 26 27 29 A N A ope P T MET HOURS 22 38 iT POR 8. SUPPLIE MAN A TURER (Ejg[34,_,
V Z
,jg lJS Zlg lM Zl@ l0]0l0l 37 4 l gg 48 l Y* g 42 l N lg 43 l4 C l6 l6 l547l@
CAUSE DESCRIPTION AND CORRECTIVE ACTIONS gl Most of tl e leakage was determined to be through valve A0 2-1601-21. The valve disc j gl was found to be slightly out of alignment with the seat. The valve shaft was I
, , l adjusted to provide correct alignment and a second leak rate test .was performed on l
, 3 l the volume. The corrected leak rate was found to be 14.5 SCFH. l l
U, L4,,,,1 1 I 7 8 9 80 SYA % POWER OTHER STATUS I C RY DtSCOVERY CESCRIPTION l NA Local Leak Rate Test l i 5 l H l@ l 0 l 0 l 0 l@l l lBl@l l ACTivlTY CO TENT (
RELEAsEc OP RELe Ast AMOUNT OP ACTIVITY LOCATION 08 RELEASE l 1 6 NA l l NA l 7 8 9 g l Zto [gl11 44 45 83 PERSONNEL EXPOSURES esuM8ER TYPE DESCRIPTION l.1,,,lf,] l 0l 0 l 0 l@l Z j@l NA l
' ' ' PERSONNE'L ' iN;UWiES " "
NUMBER OE5CRIPTION NA i R l0l0l0lgl l 7 8 9 11 12 80 LOSS 08 OR DAMAGE TO PACIL*TY Q TYPE DESCRIPTION %./
9 NA l 7 8 9 l Zl@l10 80 DESCRIPTION 2 O ISS381 L
NA l lllllllllllll}
7 8 9 10 68 69 80 7.
D. Wykoff 309-654-2241 ext. 180o NAME OF PREPARER PHONE: i l