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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20046B1481993-07-30030 July 1993 LER 93-012-00:on 930702,reactor Core Isolation Cooling Sys Declared Inoperable Due to Associated Bus Voltage Dropping Below TS Limits.Sent Operator to Cycle Timer Which Caused Affected Contact to reclose.W/930730 Ltr ML20045D9331993-07-0202 July 1993 LER 93-004-00:on 930604,unexpected CRD Low Charging Water Header Scram Received Followed by Charging Water Header A2/B2 Alarm.Caused by Crud or Foreign Matl Passing Through Suction Filter.Filters Cleaned & reused.W/930702 ML20045D7501993-06-23023 June 1993 LER 93-003-00:on 930524,Div 1 ECCS Initiation Signal Received & LPCS Pump,Lpci Pump 2A & EDG Unit 0 Automatically Started.Caused by Personnel Error.Pumps Secured & Event Documented in Personnel file.W/930623 Ltr ML20044E4161993-05-28028 May 1993 LER 92-009-01:on 920923,spurious Auto Start of CR Ventilation Emergency make-up Train Occurred Due to High Radiation Spike.Radiation Monitor Circuit modified.W/930528 Ltr ML20044E4191993-05-21021 May 1993 LER 93-011-00:on 930423,manual Scram Initiated.Caused by Disconnected Linkage on Valve Positioner on Heater Drain Valve Due to Loose Jam Nut.Tailgate Session Will Be Held W/ Instrument Maint Dept Re Jam nuts.W/930521 Ltr ML20044D5161993-05-15015 May 1993 LER 92-007-01:on 920613,high Radiation Spike Received from CR Ventilation Process Radiation Monitor,Initiating Emergency Makeup Train B.Caused by Normal Variations in Radiation Readings.Spike modified.W/930515 Ltr ML20044D5571993-05-14014 May 1993 LER 93-010-00:on 930414,DG Cooling Water Pump Automatically Tripped on Magnetic Overload.Caused by Inexperienced Trainee in Operation of Control Lever.Lesson Plans & Training Programs for Operators to Be reviewed.W/930514 Ltr ML20044C9801993-05-0707 May 1993 LER 91-010-01:on 910719 & 0805,CR a Ventilation Emergency Makeup Fan auto-started on Spurious Trip of CR Air Intake Process Radiation Monitor.Caused by Normal Variations in Background Radiation.Supply Board replaced.W/930507 Ltr ML20044B6161993-02-25025 February 1993 LER 93-002-00:on 930126,Unit 1 Manual Scram Due to a SRV Being Stuck Open Due to Duct Tape Being Over Actuators Air Valve Manifold Exhaust Port.Maint Procedures That Involve Cleanliness reviewed.W/930225 Ltr ML20024G9771991-05-10010 May 1991 LER 91-005-00:on 910410,determined That Tech Spec Required Surveillance of Suppression Chamber Oxygen Sampling Missed. Caused by Inadequate Review of Tech Spec Change.Drywell & Suppression Chamber Checked for oxygen.W/910510 Ltr ML20044A3851990-06-25025 June 1990 LER 90-008-00:on 900525,Tech Spec Hourly Fire Watch Missed Due to Miscommunications Between Security Personnel & Radiation Protection Personnel.Fire Watch re-established & Memo issued.W/900625 Ltr ML20043F1721990-06-0505 June 1990 LER 90-009-00:on 900510,RWCU Outboard Suction Isolation Valve 2G33-F004 Auto Closed Which Tripped RWCU Pump B. Caused by Procedure Deficiency.Procedure LTS-500-209 Will Be revised.W/900605 Ltr ML20043F1281990-06-0101 June 1990 LER 90-009-00:on 900511,apparent Ruptured Diaphragm Found on Pressure Differential Switch in RCIC Steam Line.Caused by Torn Diaphragm Inside Switch.Replacement Switch Installed, Calibr & Functionally Tested satisfactorily.W/900608 Ltr ML20043B5681990-05-23023 May 1990 LER 89-027-01:on 891113,primary Containment Isolation Sys Group 1 Isolation Occurred During Surveillance Testing. Caused by Burnt Out Window Light Bulbs on Alarm Window. Light Bulbs Replaced & Jumpers installed.W/900523 Ltr ML20043A7721990-05-18018 May 1990 LER 90-007-00:on 900421,reactor Protection Sys Bus a Transfer & Reactor Recirculation Hydraulic Power Unit a Inboard Isolation Valves Closed,Causing Partial Group II isolation.Out-of-svc Procedure revised.W/900518 Ltr ML20043D0591990-05-18018 May 1990 LER 90-008-00:on 900502,ESF Actuation of Control Room B Emergency Ctk Ventilation Makeup Fan Occurred.Caused by Procedure deficiency.LTS-800-205 & Similar Procedures Will Be revised.W/900601 Ltr ML20042H0211990-05-10010 May 1990 LER 90-006-00:on 900412,loop a Primary Containment Chilled Water Sys Inboard Isolation Valves & Reactor Bldg Closed Cooling Water Sys Inboard Isolation Valve Went Closed.Caused by Inadequate Procedure.Procedures revised.W/900510 Ltr ML20042F9001990-05-0404 May 1990 LER 90-005-00:on 900411,experienced Loss of Dc Power to Portion of Div I Primary Containment Isolation Sys Logic Which Resulted in Isolation Signal & Actuation.Caused by Failure to Update Drawings & procedures.W/900504 Ltr ML20042F1941990-04-30030 April 1990 LER 89-025-01:on 891101,sys Auxiliary Transformer Feed to Bus 142Y Tripped Open When Door Containing Undervoltage Relays Closed.Caused by Misalignment of Door.Door Repaired. W/900430 Ltr ML20012D5341990-03-16016 March 1990 LER 90-003-01:on 900201,RCIC Isolation Signal Occurred During Warmup.Caused by Spurious High Steam Flow Signal.Rcic Sys Piping Integrity Verified & Isolation Logic Reset.W/ 900316 Ltr ML20012D5371990-03-16016 March 1990 LER 90-002-01:on 900129,oil in Diesel Generator Governor 1A Found to Be Low & Could Not Be Seen in Sight Glass.Caused by Slow Leak from Compensation Needle Valve Plug.Proper Amount of Oil Added & Operability Test performed.W/900316 Ltr ML20012C4931990-03-15015 March 1990 LER 90-004-00:on 900213,control Room B HVAC Intake Radiation Monitor Lost Power Causing auto-start of Emergency make-up Train B.Caused by Blown Fuses.Fuses Replaced.Logic Revs for Radiation Monitors Will Be installed.W/900315 Ltr ML20012B6501990-03-0909 March 1990 LER 90-002-00:on 900209,Operating Surveillance LOS-TG-W1 Determined to Have Exceeded Required Testing Interval,Per Tech Spec 3/4.7.10.Caused by Personnel Error.Personnel Counseled & Ref Procedures Will Be revised.W/900309 Ltr ML20012B5521990-03-0808 March 1990 LER 90-001-00:on 900206,full Reactor Scram Occurred During Instrument Surveillance Testing.Caused by Actuation of APRM E Trip Circuitry.Shutdown Margin Revised & Caution Card Placed on Control Room Bench board.W/900308 Ltr ML20011F8251990-03-0202 March 1990 LER 90-003-00:on 900201,RCIC Received Div 2 Isolation on RCIC High Steam Line Flow.Caused by Spurious High Steam Flow Signal Generated When Steam/Water Mixture Admitted to RCIC Steam Line.Isolation Logic reset.W/900302 Ltr ML20011F5551990-02-28028 February 1990 LER 90-002-00:on 900129,after Filling Diesel Generator 1A Governor W/Oil,Generator Started & Declared Inoperable. Caused by Slow Leak Coming from Compensation Needle Valve Plug.Oil Added & Plug Washer replaced.W/900228 Ltr ML20006F5951990-02-21021 February 1990 LER 90-001-00:on 900122,RWCU Sys Received Div 1 Leakage Detection Ambient Temp High Isolation Signal,Causing Trips of RWCU Pumps a & C.Caused by Broken Thermocouple Input Lead.Lead Wire Reconnected & Isolation reset.W/900221 Ltr ML20011F4921990-02-16016 February 1990 LER 89-013-01:on 890907,Group I Isolation Received During Performance of Instrument Surveillance LIS-MS-401.Caused by Depressurization of Main Steam Line Low Pressure Switch. Surveillance Revised to Split Into Two parts.W/900216 Ltr ML20006E4501990-02-15015 February 1990 LER 89-010-01:on 890715 & 17,voltage Oscillations Noted on Div II Battery Charger,Resulting in Inoperability of HPCS Sys.Caused by Failure of Charger in High Voltage Shutdown Relay.Charger Energized & Relay replaced.W/900215 Ltr ML20006E4451990-02-14014 February 1990 LER 89-009-01:on 890619,diaphragm Leak Discovered in RCIC Steam Line High Flow Isolation Switch.Caused by 1 & 1/2-inch Tear in Diaphragm.Pressure Differential Switch Replaced & Calibr.Reported Per NRC Bulletin 86-002.W/900214 Ltr ML20006E3981990-02-14014 February 1990 LER 89-008-01:on 890228,reactor Vessel Low Water Level 2 Switch Found W/Setpoint in Excess of Reject Limit.Caused by Setpoint Drift.Channel Placed in Tripped Condition & All Static-O-Ring Pressure Switches to Be replaced.W/900214 Ltr ML20006E3951990-02-14014 February 1990 LER 89-010-01:on 890303,automatic Depressurization Sys Permissive Switch Found W/Setpoint in Excess of Reject Limit.Caused by Setpoint Drift.All Static-O-Ring Reactor Vessel Level Switches Will Be replaced.W/900214 Ltr ML19354D8361990-01-15015 January 1990 LER 89-018-00:on 891216,plant 250-volt Battery & RCIC Sys Declared Inoperable Due to Low Battery Electrolyte Temps. Caused by Failure of Div I Switchgear Heat Removal Sys Damper Actuators.Air Intake Dampers closed.W/900115 Ltr ML20042D3921990-01-0404 January 1990 LER 89-011-01:on 890826,spurious Reactor Protection Sys Actuation Occurred.Definite Cause of Trip Not Determined. Brief Disturbance in Reactor Protection Sys Allowed Some Contactors to Trip.Procedure revised.W/900104 Ltr ML20005E2741989-12-22022 December 1989 LER 89-028-00:on 891204,RHR Shutdown Cooling Suction Header Outboard Isolation Valve Automatically Isolated.Caused by Miscommunication Between Technician & Station Operator.Task Force Developed to Review event.W/891222 Ltr ML19351A6301989-12-15015 December 1989 LER 89-006-01:on 890214,reactor Vessel Low Water Level 3 Switch Setpoint Found Out of Tolerance.Caused by Setpoint Drift.Level Switch to Be Replaced by Analog Trip Sys During First Quarter 1990.W/891215 Ltr ML20011D1341989-12-14014 December 1989 LER 89-017-00:on 891117,flow Switch FS-2E22-N006 Found W/ Setpoint Out of Tolerance Above Reject Limit.Caused by Setpoint Drift.Work Request Written to Replace Flow Switch. W/891214 Ltr ML19351A6751989-12-12012 December 1989 LER 89-027-00:on 891113,primary Containment Isolation Sys Group I Isolation Occurred While Performing Instrument Surveillance.Caused by Loss of Power to Leak Detection Sys Logic.Isolation reset.W/891213 Ltr ML19332F0081989-12-0808 December 1989 LER 89-016-00:on 891109,RWCU Isolation Occurred While Instrument Surveillance on Ventilation Differential Temp Isolation Functional Test in Progress.Caused by Faulty Thermocouple.Thermocouple repaired.W/891208 Ltr ML20005D6621989-12-0606 December 1989 LER 89-026-00:on 891106,inadvertent Primary Containment Isolation Actuation Occurred While Clearing out-of-svc. Caused by Inadequate Logic Setup During Mod Installation. Trip Status & Output Switches repositioned.W/891206 Ltr ML19332F2431989-12-0101 December 1989 LER 89-025-00:on 891101,sys Auxiliary Transformer Feed to Bus 142Y Tripped Open When Equipment Operator Closed Door Containing Relays.Caused by Misalignment of Door.Isolations Reset,Bus Energized & Circuit Logic tested.W/891201 Ltr ML19332E6581989-11-29029 November 1989 LER 89-015-00:on 891030,shift Control Room Engineer Noted That Quarterly Standby Liquid Control Operating Surveillance LOS-SC-Q1 Was Past Critical Date.Caused by Clerical Data Entry Error.Missed Surveillance performed.W/891129 Ltr ML19332D5181989-11-22022 November 1989 LER 89-018-01:on 890515,RCIC Received Div I & Div II Isolation on RCIC High Steam Line Flow.Caused by Spurious High Steam Flow Signal When Steam Added to RCIC Steam Line. Special Test Initiated.Isolation Logic reset.W/891122 Ltr ML19327C2601989-11-17017 November 1989 LER 89-014-00:on 891020,primary Containment Isolation Sys Group 4 Isolation Occurred Causing Isolation Dampers to Close.Caused by Opening of Div 2 125-volt Dc Breaker.Power Supply replaced.W/891117 Ltr ML19325F3411989-11-13013 November 1989 LER 89-024-00:on 891013 & 30,unsealed Openings in Main Control Room Floor & Main Control Room West Wall Discovered. Caused by Wide Gap Between Structural Beam & Cable Tray. Openings Sealed & Fire Watch established.W/891113 Ltr ML19324C1751989-11-0808 November 1989 LER 89-012-01:on 890309,diaphragm Leak Found in Pressure Differential Switch 1E31-N013BB.Caused by Tear Found in Diaphragm.Replacement Switch Installed,Calibr & Functionally Tested satisfactorily.W/891108 Ltr ML20024E9121983-08-26026 August 1983 LER 83-093/03L-0:on 830801,discovered full-in Indication on Core Display for Control Rod 34-47 Inoperable.Caused by Failed Switching Transistor on Data Memory Board 19.Data Memory Board replaced.W/830826 Ltr ML20024B8381983-07-0505 July 1983 LER 83-060/03L-0:on 830606,following Turbine Trip & Scram Due to High Vibration on Main Turbine,Reactor Recirculation Breaker 4B Closed.Cause Undetermined.Auxiliary Contacts cleaned.W/830705 Ltr ML20024B0721983-06-16016 June 1983 Updated LER 83-012/03X-1:on 830210,w/reactor in Cold Shutdown,Setpoints for Switches 1E31-N612A/B Found Above Tech Spec Limits.Cause Unknown.Switches calibr.W/830616 Ltr ML20024A6961983-06-14014 June 1983 LER 83-051/03L-0:on 830516,reactor Protection Sys Trip Channel H2 Limit Switch Failed to Operate Properly.Cause Unknown.Limit Switch readjusted.W/830614 Ltr 1993-07-30
[Table view] Category:RO)
MONTHYEARML20046B1481993-07-30030 July 1993 LER 93-012-00:on 930702,reactor Core Isolation Cooling Sys Declared Inoperable Due to Associated Bus Voltage Dropping Below TS Limits.Sent Operator to Cycle Timer Which Caused Affected Contact to reclose.W/930730 Ltr ML20045D9331993-07-0202 July 1993 LER 93-004-00:on 930604,unexpected CRD Low Charging Water Header Scram Received Followed by Charging Water Header A2/B2 Alarm.Caused by Crud or Foreign Matl Passing Through Suction Filter.Filters Cleaned & reused.W/930702 ML20045D7501993-06-23023 June 1993 LER 93-003-00:on 930524,Div 1 ECCS Initiation Signal Received & LPCS Pump,Lpci Pump 2A & EDG Unit 0 Automatically Started.Caused by Personnel Error.Pumps Secured & Event Documented in Personnel file.W/930623 Ltr ML20044E4161993-05-28028 May 1993 LER 92-009-01:on 920923,spurious Auto Start of CR Ventilation Emergency make-up Train Occurred Due to High Radiation Spike.Radiation Monitor Circuit modified.W/930528 Ltr ML20044E4191993-05-21021 May 1993 LER 93-011-00:on 930423,manual Scram Initiated.Caused by Disconnected Linkage on Valve Positioner on Heater Drain Valve Due to Loose Jam Nut.Tailgate Session Will Be Held W/ Instrument Maint Dept Re Jam nuts.W/930521 Ltr ML20044D5161993-05-15015 May 1993 LER 92-007-01:on 920613,high Radiation Spike Received from CR Ventilation Process Radiation Monitor,Initiating Emergency Makeup Train B.Caused by Normal Variations in Radiation Readings.Spike modified.W/930515 Ltr ML20044D5571993-05-14014 May 1993 LER 93-010-00:on 930414,DG Cooling Water Pump Automatically Tripped on Magnetic Overload.Caused by Inexperienced Trainee in Operation of Control Lever.Lesson Plans & Training Programs for Operators to Be reviewed.W/930514 Ltr ML20044C9801993-05-0707 May 1993 LER 91-010-01:on 910719 & 0805,CR a Ventilation Emergency Makeup Fan auto-started on Spurious Trip of CR Air Intake Process Radiation Monitor.Caused by Normal Variations in Background Radiation.Supply Board replaced.W/930507 Ltr ML20044B6161993-02-25025 February 1993 LER 93-002-00:on 930126,Unit 1 Manual Scram Due to a SRV Being Stuck Open Due to Duct Tape Being Over Actuators Air Valve Manifold Exhaust Port.Maint Procedures That Involve Cleanliness reviewed.W/930225 Ltr ML20024G9771991-05-10010 May 1991 LER 91-005-00:on 910410,determined That Tech Spec Required Surveillance of Suppression Chamber Oxygen Sampling Missed. Caused by Inadequate Review of Tech Spec Change.Drywell & Suppression Chamber Checked for oxygen.W/910510 Ltr ML20044A3851990-06-25025 June 1990 LER 90-008-00:on 900525,Tech Spec Hourly Fire Watch Missed Due to Miscommunications Between Security Personnel & Radiation Protection Personnel.Fire Watch re-established & Memo issued.W/900625 Ltr ML20043F1721990-06-0505 June 1990 LER 90-009-00:on 900510,RWCU Outboard Suction Isolation Valve 2G33-F004 Auto Closed Which Tripped RWCU Pump B. Caused by Procedure Deficiency.Procedure LTS-500-209 Will Be revised.W/900605 Ltr ML20043F1281990-06-0101 June 1990 LER 90-009-00:on 900511,apparent Ruptured Diaphragm Found on Pressure Differential Switch in RCIC Steam Line.Caused by Torn Diaphragm Inside Switch.Replacement Switch Installed, Calibr & Functionally Tested satisfactorily.W/900608 Ltr ML20043B5681990-05-23023 May 1990 LER 89-027-01:on 891113,primary Containment Isolation Sys Group 1 Isolation Occurred During Surveillance Testing. Caused by Burnt Out Window Light Bulbs on Alarm Window. Light Bulbs Replaced & Jumpers installed.W/900523 Ltr ML20043A7721990-05-18018 May 1990 LER 90-007-00:on 900421,reactor Protection Sys Bus a Transfer & Reactor Recirculation Hydraulic Power Unit a Inboard Isolation Valves Closed,Causing Partial Group II isolation.Out-of-svc Procedure revised.W/900518 Ltr ML20043D0591990-05-18018 May 1990 LER 90-008-00:on 900502,ESF Actuation of Control Room B Emergency Ctk Ventilation Makeup Fan Occurred.Caused by Procedure deficiency.LTS-800-205 & Similar Procedures Will Be revised.W/900601 Ltr ML20042H0211990-05-10010 May 1990 LER 90-006-00:on 900412,loop a Primary Containment Chilled Water Sys Inboard Isolation Valves & Reactor Bldg Closed Cooling Water Sys Inboard Isolation Valve Went Closed.Caused by Inadequate Procedure.Procedures revised.W/900510 Ltr ML20042F9001990-05-0404 May 1990 LER 90-005-00:on 900411,experienced Loss of Dc Power to Portion of Div I Primary Containment Isolation Sys Logic Which Resulted in Isolation Signal & Actuation.Caused by Failure to Update Drawings & procedures.W/900504 Ltr ML20042F1941990-04-30030 April 1990 LER 89-025-01:on 891101,sys Auxiliary Transformer Feed to Bus 142Y Tripped Open When Door Containing Undervoltage Relays Closed.Caused by Misalignment of Door.Door Repaired. W/900430 Ltr ML20012D5341990-03-16016 March 1990 LER 90-003-01:on 900201,RCIC Isolation Signal Occurred During Warmup.Caused by Spurious High Steam Flow Signal.Rcic Sys Piping Integrity Verified & Isolation Logic Reset.W/ 900316 Ltr ML20012D5371990-03-16016 March 1990 LER 90-002-01:on 900129,oil in Diesel Generator Governor 1A Found to Be Low & Could Not Be Seen in Sight Glass.Caused by Slow Leak from Compensation Needle Valve Plug.Proper Amount of Oil Added & Operability Test performed.W/900316 Ltr ML20012C4931990-03-15015 March 1990 LER 90-004-00:on 900213,control Room B HVAC Intake Radiation Monitor Lost Power Causing auto-start of Emergency make-up Train B.Caused by Blown Fuses.Fuses Replaced.Logic Revs for Radiation Monitors Will Be installed.W/900315 Ltr ML20012B6501990-03-0909 March 1990 LER 90-002-00:on 900209,Operating Surveillance LOS-TG-W1 Determined to Have Exceeded Required Testing Interval,Per Tech Spec 3/4.7.10.Caused by Personnel Error.Personnel Counseled & Ref Procedures Will Be revised.W/900309 Ltr ML20012B5521990-03-0808 March 1990 LER 90-001-00:on 900206,full Reactor Scram Occurred During Instrument Surveillance Testing.Caused by Actuation of APRM E Trip Circuitry.Shutdown Margin Revised & Caution Card Placed on Control Room Bench board.W/900308 Ltr ML20011F8251990-03-0202 March 1990 LER 90-003-00:on 900201,RCIC Received Div 2 Isolation on RCIC High Steam Line Flow.Caused by Spurious High Steam Flow Signal Generated When Steam/Water Mixture Admitted to RCIC Steam Line.Isolation Logic reset.W/900302 Ltr ML20011F5551990-02-28028 February 1990 LER 90-002-00:on 900129,after Filling Diesel Generator 1A Governor W/Oil,Generator Started & Declared Inoperable. Caused by Slow Leak Coming from Compensation Needle Valve Plug.Oil Added & Plug Washer replaced.W/900228 Ltr ML20006F5951990-02-21021 February 1990 LER 90-001-00:on 900122,RWCU Sys Received Div 1 Leakage Detection Ambient Temp High Isolation Signal,Causing Trips of RWCU Pumps a & C.Caused by Broken Thermocouple Input Lead.Lead Wire Reconnected & Isolation reset.W/900221 Ltr ML20011F4921990-02-16016 February 1990 LER 89-013-01:on 890907,Group I Isolation Received During Performance of Instrument Surveillance LIS-MS-401.Caused by Depressurization of Main Steam Line Low Pressure Switch. Surveillance Revised to Split Into Two parts.W/900216 Ltr ML20006E4501990-02-15015 February 1990 LER 89-010-01:on 890715 & 17,voltage Oscillations Noted on Div II Battery Charger,Resulting in Inoperability of HPCS Sys.Caused by Failure of Charger in High Voltage Shutdown Relay.Charger Energized & Relay replaced.W/900215 Ltr ML20006E4451990-02-14014 February 1990 LER 89-009-01:on 890619,diaphragm Leak Discovered in RCIC Steam Line High Flow Isolation Switch.Caused by 1 & 1/2-inch Tear in Diaphragm.Pressure Differential Switch Replaced & Calibr.Reported Per NRC Bulletin 86-002.W/900214 Ltr ML20006E3981990-02-14014 February 1990 LER 89-008-01:on 890228,reactor Vessel Low Water Level 2 Switch Found W/Setpoint in Excess of Reject Limit.Caused by Setpoint Drift.Channel Placed in Tripped Condition & All Static-O-Ring Pressure Switches to Be replaced.W/900214 Ltr ML20006E3951990-02-14014 February 1990 LER 89-010-01:on 890303,automatic Depressurization Sys Permissive Switch Found W/Setpoint in Excess of Reject Limit.Caused by Setpoint Drift.All Static-O-Ring Reactor Vessel Level Switches Will Be replaced.W/900214 Ltr ML19354D8361990-01-15015 January 1990 LER 89-018-00:on 891216,plant 250-volt Battery & RCIC Sys Declared Inoperable Due to Low Battery Electrolyte Temps. Caused by Failure of Div I Switchgear Heat Removal Sys Damper Actuators.Air Intake Dampers closed.W/900115 Ltr ML20042D3921990-01-0404 January 1990 LER 89-011-01:on 890826,spurious Reactor Protection Sys Actuation Occurred.Definite Cause of Trip Not Determined. Brief Disturbance in Reactor Protection Sys Allowed Some Contactors to Trip.Procedure revised.W/900104 Ltr ML20005E2741989-12-22022 December 1989 LER 89-028-00:on 891204,RHR Shutdown Cooling Suction Header Outboard Isolation Valve Automatically Isolated.Caused by Miscommunication Between Technician & Station Operator.Task Force Developed to Review event.W/891222 Ltr ML19351A6301989-12-15015 December 1989 LER 89-006-01:on 890214,reactor Vessel Low Water Level 3 Switch Setpoint Found Out of Tolerance.Caused by Setpoint Drift.Level Switch to Be Replaced by Analog Trip Sys During First Quarter 1990.W/891215 Ltr ML20011D1341989-12-14014 December 1989 LER 89-017-00:on 891117,flow Switch FS-2E22-N006 Found W/ Setpoint Out of Tolerance Above Reject Limit.Caused by Setpoint Drift.Work Request Written to Replace Flow Switch. W/891214 Ltr ML19351A6751989-12-12012 December 1989 LER 89-027-00:on 891113,primary Containment Isolation Sys Group I Isolation Occurred While Performing Instrument Surveillance.Caused by Loss of Power to Leak Detection Sys Logic.Isolation reset.W/891213 Ltr ML19332F0081989-12-0808 December 1989 LER 89-016-00:on 891109,RWCU Isolation Occurred While Instrument Surveillance on Ventilation Differential Temp Isolation Functional Test in Progress.Caused by Faulty Thermocouple.Thermocouple repaired.W/891208 Ltr ML20005D6621989-12-0606 December 1989 LER 89-026-00:on 891106,inadvertent Primary Containment Isolation Actuation Occurred While Clearing out-of-svc. Caused by Inadequate Logic Setup During Mod Installation. Trip Status & Output Switches repositioned.W/891206 Ltr ML19332F2431989-12-0101 December 1989 LER 89-025-00:on 891101,sys Auxiliary Transformer Feed to Bus 142Y Tripped Open When Equipment Operator Closed Door Containing Relays.Caused by Misalignment of Door.Isolations Reset,Bus Energized & Circuit Logic tested.W/891201 Ltr ML19332E6581989-11-29029 November 1989 LER 89-015-00:on 891030,shift Control Room Engineer Noted That Quarterly Standby Liquid Control Operating Surveillance LOS-SC-Q1 Was Past Critical Date.Caused by Clerical Data Entry Error.Missed Surveillance performed.W/891129 Ltr ML19332D5181989-11-22022 November 1989 LER 89-018-01:on 890515,RCIC Received Div I & Div II Isolation on RCIC High Steam Line Flow.Caused by Spurious High Steam Flow Signal When Steam Added to RCIC Steam Line. Special Test Initiated.Isolation Logic reset.W/891122 Ltr ML19327C2601989-11-17017 November 1989 LER 89-014-00:on 891020,primary Containment Isolation Sys Group 4 Isolation Occurred Causing Isolation Dampers to Close.Caused by Opening of Div 2 125-volt Dc Breaker.Power Supply replaced.W/891117 Ltr ML19325F3411989-11-13013 November 1989 LER 89-024-00:on 891013 & 30,unsealed Openings in Main Control Room Floor & Main Control Room West Wall Discovered. Caused by Wide Gap Between Structural Beam & Cable Tray. Openings Sealed & Fire Watch established.W/891113 Ltr ML19324C1751989-11-0808 November 1989 LER 89-012-01:on 890309,diaphragm Leak Found in Pressure Differential Switch 1E31-N013BB.Caused by Tear Found in Diaphragm.Replacement Switch Installed,Calibr & Functionally Tested satisfactorily.W/891108 Ltr ML20024E9121983-08-26026 August 1983 LER 83-093/03L-0:on 830801,discovered full-in Indication on Core Display for Control Rod 34-47 Inoperable.Caused by Failed Switching Transistor on Data Memory Board 19.Data Memory Board replaced.W/830826 Ltr ML20024B8381983-07-0505 July 1983 LER 83-060/03L-0:on 830606,following Turbine Trip & Scram Due to High Vibration on Main Turbine,Reactor Recirculation Breaker 4B Closed.Cause Undetermined.Auxiliary Contacts cleaned.W/830705 Ltr ML20024B0721983-06-16016 June 1983 Updated LER 83-012/03X-1:on 830210,w/reactor in Cold Shutdown,Setpoints for Switches 1E31-N612A/B Found Above Tech Spec Limits.Cause Unknown.Switches calibr.W/830616 Ltr ML20024A6961983-06-14014 June 1983 LER 83-051/03L-0:on 830516,reactor Protection Sys Trip Channel H2 Limit Switch Failed to Operate Properly.Cause Unknown.Limit Switch readjusted.W/830614 Ltr 1993-07-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217C9121999-10-12012 October 1999 SER Input Authorizing Licensee Proposed Request to Modify Definition of Core Alteration in Section 1.0 of TS & Update Sections 3/4.1,3.4.3 & 3/4.9 to Reflect Proposed Definition Change ML20217F9091999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for LaSalle County Stations,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 ML20212C4501999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for LaSalle County Station,Units 1 & 2.With ML20210R0671999-07-31031 July 1999 Monthly Operating Repts for July 1999 for LaSalle County Station,Units 1 & 2.With ML20210C1681999-07-0909 July 1999 Seventh Refueling Outage ASME Section XI Summary Rept ML20209H1501999-06-30030 June 1999 Monthly Operating Repts for June 1999 for LaSalle County Station,Units 1 & 2.With ML20195J7871999-05-31031 May 1999 Monthly Operating Repts for May 1999 for LaSalle County Station,Units 1 & 2.With ML20209E1431999-05-31031 May 1999 Cycle 8 COLR, for May 1999 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 ML20206N2071999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for LaSalle County Station,Units 1 & 2.With ML20205L8421999-03-31031 March 1999 Rev 2 to EMF-96-125, LaSalle Unit 2 Cycle 8 Reload Analysis ML20205L8301999-03-31031 March 1999 Administrative Technical Requirements App B (Amend 26) LaSalle Unit 2 Cycle 8 COLR & Reload Transient Analysis Results, for Mar 1999 ML20205R2721999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for LaSalle County Station,Units 1 & 2.With ML20205L8391999-03-22022 March 1999 Rev 2 to 960103, Neutronics Licensing Rept for LaSalle Unit 2,Cycle 8 ML20204C8141999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for LaSalle County Station,Units 1 & 2.With ML20199E4601998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for LaSalle County Station,Units 1 & 2.With ML20207C7371998-12-31031 December 1998 Annual Rept for LaSalle County Station for Jan 1998 Through Dec 1998 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with ML20198B3801998-12-14014 December 1998 SER Accepting one-time Request for Relief from Certain Provisions of Section XI of ASME Boiler & Pressure Vessel Code,Per 10CFR50.55a for Certain Plant Safety/Relief Valves ML20206N2261998-12-0909 December 1998 LER 98-S03-00:on 981116,protected Area Was Entered Without Current Authorization for Unescorted Access Due to Programmatic Deficiency Error.Changed Badge Control Process ML20197K0981998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for LaSalle County Station,Unts 1 & 2.With ML20196B1441998-11-23023 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Bindings 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 ML20195D3191998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for LaSalle County Station.With ML20154H6781998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for LaSalle County Nuclear Power Station,Units 1 & 2 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 ML20153C7621998-09-18018 September 1998 Safety Evaluation Acceping NRC Bulletin 95-002, Unexpected Clogging of RHR Pump Strainer While Operating in Suppression Pool Cooling Mode 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 ML20151W0241998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for LaSalle County Station.With ML20237E2921998-08-21021 August 1998 Special Rept:On 980811,channel 5 of Lpms Became Inoperable. Caused by Channel Failed pre-amplifier Located Inside Primary Containment at Inboard Side of Electrical Penetration E-19.Initiated Repairs of Channel ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20237B4861998-07-31031 July 1998 Monthly Operating Repts for July 1998 for LaSalle County Nuclear Power Station Units 1 & 2 ML20236V7701998-07-31031 July 1998 Revised LaSalle Unit 1 Cycle 8 COLR & Reload Transient Analysis Results ML20236P8231998-07-14014 July 1998 Special Rept:From 980614-17,various Fire Rated Assemblies Were Inoperable for Period Greater than Seven Days.Caused by Test Equipment Being Routed Through Fire Doors.Established Fire Watches & on 980619 Assemblies Were Declared Operable ML20236N6751998-07-0909 July 1998 Part 21 & Deficiency Rept Re Notification of Potential Safety Hazard from Breakage of Cast Iron Suction Heads in Apkd Type Pumps.Caused by Migration of Suction Head Journal Sleeve Along Lower End of Pump Shaft.Will Inspect Pumps ML20236L8041998-07-0606 July 1998 Safety Evaluation Granting Licensee 980304 Request for Second 10-yr Interval Pump & Valve IST Program Plan,Rev 2, Including Changes to 2 ASME Boiler & Pressure Vessel Code Relief Requests Previously Submitted in Rev 1 ML20236P3611998-06-30030 June 1998 Monthly Operating Repts for June 1998 for LaSalle County Nuclear Power Station,Units 1 & 2 ML20249C4891998-06-22022 June 1998 Special Rept:On 980522,Fire Detection Zone 1-31 Was Noted out-of-service for More than 14 Days.Detection Sys Was Taken out-of-service on 980508 to Prevent False Alarms During Hot Work Activities.Sys Was Returned to Operable Status 980528 ML20248M3101998-05-31031 May 1998 Monthly Operating Repts for May 1998 for LaSalle County Nuclear Power Station,Units 1 & 2 ML20236V7771998-05-31031 May 1998 Rev 1 to 24A5180, Supplemental Reload Licensing Rept for LaSalle County Station Unit 1 Reload 7 Cycle 8 ML20217Q7041998-05-0404 May 1998 Safety Evaluation Accepting Util Request to Leave Leak Chase Channels Plugged During Performance of Containment ILRT ML20247M4491998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for LaSalle County Station ML20216F4941998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for LaSalle County Station,Units 1 & 2 ML20217N6581998-03-30030 March 1998 Special Rept on Fire Detection,Deluge Sys & Fire Rated Assemblies During Period of 980303-25.Established Fire Watches Until Affected Equipment Is Returned to Operable Status ML20216D9511998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for LaSalle County Station,Units 1 & 2 ML20247M4631998-02-28028 February 1998 Rev Monthly Operating Rept for Feb 1998 for LaSalle County Station ML20203D7241998-02-20020 February 1998 Special Rept:On 980118,Fire Detection Zones 1-18 & 2-18 Taken out-of-svc to Prevent False Alarms During Hot Work Activities on Auxiliary Electric Equipment Room Ventilation Sys.Fire Watches Will Remain in Place ML20202G9851998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for LaSalle County Station,Units 1 & 2 ML20199K1651998-01-23023 January 1998 Rev 65h to Topical Rept CE-1-A, Comm Ed QA Tr 1999-09-30
[Table view] |
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. h m E.PtpqT ;A rg sc EVENT OE5cRimCN AND PRCEASti 00HSECUENCES h m :On 7/23/82 Commonwealth Edison notified NRC Realon ill and NRR personnel of a potential i
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r i;.t lWri tten response of this potential 10CFR50I55E) will be issued. ;
I f i3 y.gAdditional_.tes.ts will be conducted on a full size LaSaile Countv valve as described l
l , ; j. ;i n .let ter. f. rom .L. DelGeorge to H. Denton, dated 7/26/82.
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. - 1. LER NUMBER: 82-065/olT-0
- 11. LASALLE COUNTY STATION: UNIT I 111. DOCKET NUMBER: 050 373 IV. EVENT DESCRIPTION On July 23,1982 at 1300 hours Commonwealth. Edison Nuclear Licensing notified NRC Region 111 and NRR personnel of a potenial 10CFR50 (55E) assesment. Recent design reviews of the " Surpression Chamber to Dry-well Vacuum Breakers" has led to the tentative conclusion that the installed GPE 30" Vacuum Dreakers may exceed the LaSalle County specific load defini tion as defined by the Mark 11 Owners Group for the case of a double guillotine failure on a recirculation loop. The LaSalle County Drywell 1 Surpression Chamber Vacuum Breakers Valves are ex-pected to undergo a single open-shut cycle as a result of the Surpression Chamber pressurization associated with pool swell in a Post-Loca condition..
V. PROBABLE CONSEQUENCES OF THE OCCURANCE:
Commonwealth Edison has reviewed and presented to the NRC Staff results of extensive analyses and confirmatory tests performed by Mark I con-tainment owners having GPE 18" and 24" vacuum breaker valves of similar design. The applicability of these results to the LaSalle County valves was reviewed with the . Mark 1 Owners' consultant who Judged the results indicative of the capability for the. LaSalle County Vacuum Breaker valves to operate adequately and perform their intended safety function.
VI. CAUSE The potenial problem of Drywell/Surpression Chamber Vaccum Breaker valve failure was reviewed with the NRR Staff on July 23, 1982. Further discussion is provided in the attached letter from L.O. DelGeorge to H.R.
Denton dated July 26, 1982.
Vll. CORRECTIVE ACTION Vritten response of this potenial 10CFR50 (55E) will be issused to document this Drywell/Surpression Chamber Vacuum Breaker Valve Post-Loca chugging phe aomena. Additional tests will be conducted on a full size LaSalle County valve as described in the above letter.
Prepared by: D. Vinterhof f o$c gr.9<-s2 s' 4
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LSCS Jul y 2 6, 1982 Messrs: C. Reed /N. A. Kershaw Q.A. Engineer - SNED J. S. Abel D. J. Scott (NC Only)
H. E. Bliss V. I. Schlosser - RIII Only J. D. Bowers B.R. Shelto n/ T . E. Watts L. J. Burke - RIII Only W. J. Shewski T. C. Cihla r/ G. E. Peterson B . B . Stephenson/D . L. ShambJ in E. E. Fitrpatrick W. L. Stiede D. P. Galle H. K . Stol t/W . L. Eck K. L. Graesser (NC Only) (Bull. , Circ . , & I. No t. Only)
J. F. Gudac' (NC Only) R. J. Tamminga (ISI Only)
R. H. Holyoak G. P. Wagner J. H. Hughes P. P. Steptoe-IL&B (Letters Only)
R. E. Jo rtberg H. R . Pe f fer - G.E.
N. J. Kalivianakis M. A. Bowidowicz - S&L A. W. Kleinrath/A. D.'Rossin G. Wright - State o f Illinois R. Kyrouac (NRC/ CECO Ltrs only)
D. E. Lindvall G. F. Owsley - Exxon J. J. Maley W. R. Bird - Consumer s Powe r Co .
T. E. Quaka - RIII Only R. E. Querio - RIII Only NL Distribution R. Ralph In the judgement of the Nuclear Licensing . Administrator, the attached document contains the fol* lowing commitments to the NRC or requirements from the NRC. .
Identification o f At tached Document: LSCS1/2 - Drywell/Wetwell Vacuum Breaker Assessmen t; L. O. DelGeorge letter to H. R. Denton dated July 26, 1982. '
NRC Commitment or Recuirement:
Responsible '
Due Date Commitment or Recuirement Edison Depa rtme n t 8/13/82 Provide NL A schedule for conduct of vacuum PE/B. Shelton breaker test-inclucing expected date o f bounding test run.
11/1/82 Provide NLA evaluation report on results PE/B. Shelton of vacuum breaker test for transmittal to NRC. Note: NLA should be on distribution for all preliminary drafts of this report. -
NOTE: (1) Proj. Eng. is expected to coordinate with LSCS operations / construction to make a valve availaole for testing.
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4 (2) Proj. Eng. is expected to initiate a final design basis evaluation o f the subject valves including plant specific load definition, possible modifications and replacement of the U2 valve used for the test.
When it is determined by the responsible department that a due date will not be met, the Nuclear Licensing Administrator should be notified immediately.
L. O. De1 George 82-29 d
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Ac::ress Reply to- Post Othee Box 767
,, Chicago lhinois 60690 July 26, 1962 Mr . Ha r ol d R . Denton, Director Dffice o f Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Units 1 and 2 Drywell/Wetwell Vacuum Breake r Assessment NRC Docket Nos. 50-373/374
Dear Mr. Denton:
The purpose o f this submittal is to document the
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Commonwealth Edison assessment of the drywell/wetwell vacuum breakers 1982. The which was reviewed with your staff on July 22 and 23, subject valves have been reviewed through the Mark II Containment Owners Group and a generic. bounding load definition established. The methodology associated with this review was discussed with your s ta f f in May,1982. We are unaware, at this time,methodology this of any dutstanding has not questions although a report documenting yet been submitted. Relative to the bounding load definition, it should be recognized that the LaSalle County valves, due to their location outside the wetwell, do not experience cycling (open and shut) as a result of the post-LOC A cnugging phenomena. The LaSalle County valves are expected to undergo a single open-shut cycle as a result of the wetwell pressurization associated with pool swell. The pipe break size leading to the maximum swell induced dif ferential pressure (dP) is ciscussec in At tachment I o f this letter.
having established tha t a pool swell inducedd P across the vacuum breakers must be assesse0 to assure that the valve safety function will be fulfilleo, Commonwealth Edison has performed what is juoged a bounding assessment u-sing bounding generic angular velocities associated with the maximumd P developed through the Mark II Owners Group. The conservatisms in this load definition, both witn respect to the generic methodology and plant specific applicaton to LaSalle County are discussed in At tachment II. Of primary importance is the fact that the LaSalle County valves are outsice tne wedell and are in a pipe loop which also contains two isclation valves. Tnis configuration is likely to have a sicnificant effect on the LaSalle County specified P and thereby, tne LaSalle County specific load definition. These numbers , and other ' plant-specific adjustments to the generic load denfition are now being deve1Yp[ed"a'nq Sill be incorporateo into the final design casis evaluation.
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_2 Notwithstanding the expected reduction in the plant specific load definition, a bounding evaluation of the vacuum breakers has been performed. This evaluation, which was reviewed with the staff on July 22, 1982 indicated that the LaSalle County vacuum breakers would fulfill their safety function.
--~ ; n uncertainties associated with this evaluation, i.e. theAs a result o f applicability of the closing impact results on the opening impact
" loads on the valve body and internals; Commonwealth Edison reviewed and prsented to the NRC staf f the results of extensive analyses and confirmatory tests performed by Mark I containment owners having vacuum breakers of similar design (18" and 24" GPE valves) . The applicability of these results to the LaSalle County valves was reviewed with the Mark I owners' consultant, and those results were judged indicative o f the capability of the LaSalle County valve.
The results of this detailed assessment, presented to the staff on July 23, 1982, are docuemented in At tachment III. O f particular interest is the fact that this assessment clearly _ indicates that the; closing impact bounds the opening impact case, the major expressed staff concern.
It is o'n the basis of the simplified analysis of the LaSalle County vacuum breakers, supported by the applicable Mark I program detailed analyses anc test of similar valves that Commonwealth Eoison.has concluded that the LaSalle County vacuum breakers function. are adequate and w111 ful' fill 1 heir intended safety This conclusion is further supported by the fact that the
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- r. bounding load ' definition is of exteremely low probability due to its dependence on a double guillotine failure of the largest diameter pipe (s) ( At tachment I); and due to the expected reduction in the LaSalle specific load definition When appropriate refinements are made t o reflect actual LaSalle County parameters ( At tachment II) .
Notwithstanding our belief in the present adequacy of these valves, Commonwealth Edison will conduct a confirmatory test on a full size LaSalle County valve using the bounding generic load definition. This test will be completed and an evaluation presented to the NRC by November 1, 1982. The NRC Staf f will be given advance notice of the cate of the bounding test run to allow for their observa tion o f the' tes t if desired. It is fully expected that this test of an actual LaSalle County valve to what is judged a very conservative load cefinition will resolve all present uncertainties.
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If you have any questions concerning this submittal, please contact this office. .
Very truly yours, L . O . DelGeo rg e cc: NRC Resident Inspector - LSCS O
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Reference:
LSCS OAR Vow = 221,500 ft3 Vww = 166,400 f t3 Ventsub = 12.83 f t Avent = 295.2 f t2 98 vents)
Vpool = 142,160 ft Pool depth = 26.5 ft.
Apool = 5069 ft2 (i.e . Voool - A vents)
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- 2. For LaSalle County at F = .06 ABreax 1.37 ft2 for max. uplift pressure
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,$, SEE FIG. l- 4 , PALLET STOP MAGNET L ATCt i STuo LOCATION YACUOM BREMER W/LL (4 Pl.CS)
P!GUllE 1-2 U .
GPl! 18" VALVF. Cl!OHliTRY c
* (GPli TYl'F. LD-240-208) -
g O .
- v. . w_ . : . . . . . = . - = -
l l
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0I5-67-018 1-10 Mu.i.ec!h t
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^
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CIS-67-01B 1-e nutec 1
.. . ... . .m,-,,.-.--,- -.:------..._.-.--.
. 'l i
i it h
F !
PALLET SE AL _
F PALLET SF.AL RET AINER M ]i\\
l-I- ~ STANLESS STEEL '
WELD BUILDUP pAtLE r i
,\\\
'd
~ SEAT Rt NG . SA-54 I
(fl//////nN-i w
/gi I il ,
. i 1 i , i -
r iI ,'il tvl: ,s
- q \
PALLET STOP STUO h m.4 u:Es -
SEE FIO. 1-2 FIGURE l-4 GPE IS" VALVE SEAT DETAILS t:; .- __ ..
cts-c -ois 1-9 nuteCh
~
~%, w ww - - -,. ,T -C
~ . . .,mr -
_ - ,. .. o-n ,- ~ n.- ...- , , _ _ . - - . - - - .-----.a-w---->
l l
l PA1.LET HlNSE - HINGE PlVCT SHAFT l TBMXET l I
I f// ~^ 2
+-dk y l\\ \\(
[N\ x\ H % L.-.e/ rA i I LNNN NNNNNNNNNNN z r N i M
'-~
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cl a .- ...}
lY /A i .P)voT stock t HINGE Anu SECTIObd A-A
' ' VA1.VE
-- SE;47 x .
x'
- pivot t-PwoT BLOCK
\ SoLT (V:Y o ,
di siwcn i
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, HINGE ARM
-e-m-
STUD (Vi)
W4.LE"T PAL 1 ET HI N G E.
PALLET BFJ CKET HINGE SR AC.K,c i sTuo(.W)
FIGURE l-7 i
GPE 24" VALVE HINGE GEOMETRY
\
l E~.'..- . , _ . - .
1-12 nUteC:~1 j n-s7-oss i
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c 7:-:hLEsS STEF-Wr D St M'- O U P FIGURE 1-6 C?E 24" VALVE SEAT GEOMETRY C .- ,n.
cts-a -ci s 1-11 nutOCh
..._- : n...- - . _..-
-- . w . -- - - ~ .: w:. . . . . . . . -
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VALVE ASSDel.Y AND 50RCE SYSTEM .
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- J GEN-67-011 e 01'* - ( 91) 3380d 3MnSS3Md DSIC (3 nil 03HSUG) 00 a co t h 00 0 t co g- .
co a- ?
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-..;. J..a r-;:..:: a L m.a u:=- . .
GEN-67-011 (a) UP IMPACT 6
6 7 -
7
~
8 kN
.r \
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- - \
S GAP ELEMENT 8
/T 1 2 3 4
/HH/' *
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(b) D0'h3 IM?ACT ._
I 2 GAS.GT MATERIAI. GAP ELEMENTS - 8, 9, 10, 12 I egi
' ' !]
d METAL TO METAL CONTACT l .
(DISC WITH STUDS) i 3 GAP EME - 11 ..
I j '10,' 4) , ,
-h^
4 I n
.- /-bCEO
; 5 Q+'_
2 n .,,
r-i &
FIGURE 4-1 EEAM IMPACT MODELS USED c..
~~~
IN'GPE' ANALYSIS
. .s Mutech
~ ~ - - . - -
. . . . . - . . . 2 m .x a .. a m .. a _. , ;g g3; 3; GEN-67-011 UP = UP-IMPACT (OPENING) DOWN = DOWN-IMPAC (CLOSING) 2
~ -
~
$ IUP .
k.
e.
(u
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5
- \ y
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f, a -
- s. l, l .:.
F *'
.' ...,- t 1:i ;!
l , ; . ~C 3. S ll l
4- ,
j FIGURE 4-2 3 GFE DISC AND VALVE EODY STIFFNESS AND l
DYNAMIC MASS DETER.MINATION MODELS c:. .- .-. -.
~
4-7 nutech
......-....;-. . . - w za .
w= - :. .- s-w k
! \
s GEN-67-011 1
.L
- T -
a.
A e 6 ~e @ -
~ 1. 5. . ' ,
1.25 1.00 f
P G f \
n .
as .
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z U
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[N 3 z
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\ pg .
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x ,v - x p"
o 0-o \ /
h
.25 s <
s ( -
> .50 g
.75 ,
- 1
2 3 4 ,
O
~
TIFE (MSEC)
A'
GPE UP-IMPACT ANALYSIS DISC (3EAM) NODES DISPLACIMENT HISTORY e .- -
3-8 nuteCh
.. . . . -.L __ .. . - , . .- - - - . . . . ::.-- :,: ..- . c.a . . . .
4 GEN-67-011 ,
3 . 6 '- ,
i i
.(
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,. 1
/
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c *S 17 r.: d N J - % ~
4 i
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I' J
i 4 0 -
i I l 7 8 9 10
_4s, 4 5 6 2 3 O 1 -
TIME (MSEC) 4 ..e
- i. J .c -
/. 2 .
1 l
FIGURE 4-4 GPE DOWN-IMPACT ANAI.YSIS DISC ($21M) TODES DISPLACEMENT HISTORY nutech
>-= _ .
, :::,; - -- - - - - - - =
t-
_.,==,g--c.;~22.,: ;
.l o
'h i
,h ,
4S i l* . ,!
I 25 Ks1 I
^
r.
i !
' ! 50 Ksr l 12s nst
- 00 K51 ,
u o
- 's
\
r
/
Y
' \
O KSt 1 \ { ~
^ \ 50 KSt #
{
i i
FIGURE B.3-5 !.
O g GPF. PA!l.ET DOWN-IFIFACT STRESS CONTOUR Pl.0T OF
' l H(D flAXI.tlli.tl STRl!SS AT TOP SURFACF. . .
IO ,
7 :
i i
.. _. 6 4 .._ . . . . ,
e Max Principal Stress Top Surface O M: Principal Stress Mid Surface
- Min Principal Stress Bottom Surf ace 150 -
100 - -
r 50 -
n m C -
0 y . i 6 3
- r. .-
12 2 4 6 8 10 w Distance Frc= Centerline (Inches) m l
)
1
-1C0
-150 -
FIGURE 3.3-6 GPE PALLET D3WN-IMPACT PRINCIPAL STPlSS
~ DISTRIBUTION ALONG SECTION FRO *. PALI.ET fESTErTO ' IMPACT POINT (SECT 7.ON A- A)
CES.57_cig 3.23 nutec1
.I.
- s. .
~
i . ., -
A- -
125(SI y 100 KSI ci, -
w 75 KSI- , ,-i 50 KSI' is
% 25 KSI 1 __
~ .
(
p %
/ *
! f k- -
0 KSI
% N i
' J .
* - t
- g.
' [ 5 KSI . .. .
~
l l 25 KSI . .
l .
.s. a i .
- t ..
c)
. A_ .
p1 w
(
.r.a.,*. 7 '
- PIGURE 5-6 e
- . .o ,
u U , t..-
CPP. DISC UP-!MPACT STRESS CONTOUR l'I.0T OF f, ,
, C ..
~
e l ed MAXIMUM STRUSS AT TOP SURFACE g ,
O 1
- J
. (,
__.t i
, .l g za
/ .
jk
,24 n< nili n g Str-ss (tx)-
N $ :
S 20 i
16 3
- p R 12 -
I i v4 n " ,
V i kl g ~
'V
.. ~ -
nn (F2
/ )
Dir cct : tres 0
M ,.~
O l
, .a _ _ _
Benling stre s (M ')
s ,
t .
-12 1.00 1.25 1.50 1.75 2.00 2.25 2.50 0 .25 .50 .75 J
1 Time (MS!!C) !
FIGilRE 11.3-11 DiiTAll,ED lilNGE MODEl, D0lfN-DIPACT STRESS RESULTS l
3 l MAXIMUM STilESS IN ll!NGE EAR C ?
c+
(D .
O or .
p
l I
- 't
_s *
. 2 t -
s A 5 I
=/, :
?..*2 \~ :
@Q .* / : ,*, ,\ ,\,e\" I I I l
.
8' -
- e
,\s h I
i
- /~
\
/ @
t@ h
.=
FIGURE 3.3-7 ,
CRITICAL STFlSS PLANES FOR COMPUTING PALL.T .
STRENGTH RATIOS FOR D0hW-IMPACT r; .- _ . . .
- 3. n nutech Cn-o w .
. - .. .- ..-.;..-...... . . . . .. . . .:a.. .... .
+
~@
e.o a n ..
e.g n ,
= ,
g v
n w w "\T
- .! is ,i , ,
e i n
** i a m se y ,,' -l t .wl u p .
. j se .
-2.. *
- n n ., !
, .i = t,
~~
" u
-=.- . ,
M M g
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4.1, en
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4.4 4.A is.4 4AJ 4.s 4.9 880.'@'
- C.Pt me.a.at segactt C.'14
.@ . e,-
FIGURE 7-1 ' . .. . -
-~~~
CRITICAL STRESS PLANES FOR CO".?ITING
- DISC STRENGTH RATIOS FOR UP-IMPACT -
c .- .
7-3 DUteCh
P GEN-67-011
,t TABLE 7-1 Disc /?alle Strenr. h Ratios
- SHORT TERM 3 ..
FSTF FORCING FUNCTION FORCING FUNCTION.
l J DOWN-1MPACT DOWN-IMPACT
.J- UP-L'CJACT 4~ " (19.59 RAD /SEC) (8.04 RAD /SEC) e- .
(16.02 RAD /SEC)
.' Disc Section' .36 .15
'-- .20 (1) * .
.62 .L .25
'. .36 _...-
I (2) .52 .21
.24 (3) .32 .13
.20 (4) .14 .06
. 19 (5) .
NOTE: .
ten forcing 5metion.
- 1. No up-i= pac predicted with shcr:
l : ,
t*
J J . ,
1 .
4 .
h *%.
* . ==.
,. .:.J
)
e i
I
.a e
t*; . . -n..
G 7-5 nutech,
- , _ - - . . -- -- - - , _ _ _ _ _ . , - - _ . . ._.,,_v ,-...---,y-. s-- - - -
l I
4 .
sj> . . . ll
9.'
=
- s. _
T Aftl.E 7-2 [
, (,, '
Illny,e Component StrenP,th Rattos 8
e b ' '
Sil0RT TERM
'
- FORCitlG FUNCTION FSTF FORCIllG FUtiCTIOri DOWN-IHPACT (19.59 RAD /SEC) DOWN-IMPACT (8.04 RAD /SEC) f U P- I ttP ACT (16.02 RAD /SEC)
STRESS CAPACITY STREllGTil STRESS CAPACITY STREt4GTil CAPACITY STREttGTit (KSI) RATIO
; TYPE OF STHESS RATIO (KSI)_ JL RATIO g{
JKy( Jg_
cottPotlEt1T STRESS 0.26 110.4 0.63 28.6 110.4 78.02 110.4 0.71 69.7 0.1 89.7 IllllCE ARM DENDittC 0.2 89.7 DIRECT 0.4 89.7
'.' 49.9 108.5 0.47 o.
108.5 0.71 .121.6 108.5 1.l4 46.6 tiltlGE SilAPT BF.tIDillG 73.2 '10.1 46.6 4.1 SilEAR ' 9.5 46.6 84,8 1.7 84.8 T.I 84.8 4.1 .
DIRECT 8.9 110.4 0.09 0.21 21.8 110.4 0.22 DEt4Dit1G 22.6 110.4 89.7 3.6 89.7 litt4GE EARS 4.5 89.7 8.7 ,
DIRECT 0.87 28.2 84.8 0.36 64.8 1.18 68.6 84.8 46.6 tilllGE ARH DIRECT 75.5 .
46.6 6.3 36.2 46.3 15 5 STtil)$ SilEAR ,
fl0TE:
* - llo up-1:npact predicted.wlth short terra forcing function. n 6
vO C. .
0 1
(D O ,
7 ~
es t
y a ._ _ _ . - . - . a -- .r m . a ~ -- --n=-. --
_ w w w ;
9.6 STRENGTiiS FATIOS C:=pe nen: strengths for the vacuu: breakers were computed u:ilizing the Von-Mises =axi.=u= energy distortion criterion to
- cycluate the ulti= ate strength capacities. According to this crit erion , failure occurs when the strain energy reaches the ulti= ate capacity of the componen: section. Accordingly, the interaction equation for general loading is: ,
( ( _)2 , ( u_)2 , (_ u )2 , ( u
)2y 1/2 - s:rength Ratio u
where:
M, S,T, V- co=pu,ted bending =o=ent, nor=al force, torque, and she ar force * ~
Mu = ulti=a:e bending = omen: assu=ing Sy at the neutral surface and S u a: the outer fib er Nu = ultimate =e=brane capability assuming S u through the section Tu - ulti= ate torque capability assu=ing a shear s:ress equal :o .555 u through the section ,
vu - ul:Lcate sheer capabili:y assu=ing a shear stress eque.1 :o .555 u through the section I0r all co=ponents, ulti=a:e capacities for each = ode of loading Vere developed based on the =aterial properbies fro Tables 2-3
'nd 2-0 =u'_:iplied by she S:: air Rate Factor.
05 - 5 7_ g -- 2-24 ,
n a i + n m,
TABLE C.3-1 STRENGTd RATIOS FOR G?E 24" VALVE .
ULTIMATE YIELD TENSTir STRENGTd MATERIAL STRENGTA STRENGTd RATIO COMPONENT s
FALLET SA240 TYPE 316 22.33 ksi 75 ksi .27 (SECTION 3-3)
SA240 Tf?E316 22.35 ksi 75 ksi .32 EINCE 3?.ACKET FALLET HINGE '
SA320 BS 30 ksi 75 ksi .76 3 RACKET STUD i
t
!VOT BLOCK BOLT 30 ksi 75 ksi .35 1
SA193[B8M' _
?IVOT BLOCK SA479 TYPE 316 : 30 ksi 75 ksi .13 SA479 TYPE 316 30 ksi 75 ksi .28 -
EINGE FIVOT SF.AT-
.a 1
SA320 BSM 30 ksi 75 ksi .503' i f. NGE A:LM STUD I: 75 ksi .06 l H:5GE ARM SA479 TYPE 316 30 ksi r;_ . .- .
- ,. ,7 - 01 S
*-o =
C.26 L
Flu teC 1
j.
^
UNMODIF18 TEST PROGRAM I PURPOSE OF TESTS SHOW LINEARITY OF ANALYSIS CONFIRM THE ACCURACY OF ANALYTICAL RESULTS WITH A SIMPLE DROP TEST 1DENTIFY ANALYTICAL CONSERVATISMS 8 TEST SEIUP 4 TEST RESULTS e
i 1
C .- -.~ - --
17 -
5/2/s2 nutggcb
, - - ,,-w-y- ,---
--iy ,- ,,,y. -
p ,m,y,% --.- . _ - -_ - ,.,.---. _ _ . .-,-, ,_--,. . . ,ye ,, ,,rm.-- , - ,-- -.,----,,--+-m _
c - , j 1
~
1 l
i TEST SETUP I SIMPLE DROP TEST e 10 * , 20* , 30* , 40* , 50, 54*
8 TEST INSTRUMENTATION s STRAIN GAUGES e ' VELOCITY PROXIMITY PRO 3E (USED -
TO MEASURE IM?ACT VELOCITIES) .
e IliCLIN0 METER O
9 I
e U.; . .- , . _ _ .
~~ MUI Q 5/2/52 _ _ _ _ _ __ _ _ _ --
. _ . . . _ . _ _ .. ..______.______.m_.- ..:_e__ _ ..e _ - . u _
.. 4 m-
=
C U
w
=
. _=
H 8
8 a
a W
A >
3 m
C e
k w e m z
& O '
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s' D a -
0 z
e o
~
w x b W
5 .
z 1
= .
m
= -
2
, o
< s. a l
> w z c:
- < ar>
H u-d w e E z 5 x, = =
- > m - -
a e: a $ U O C
c g o l > 2
= w
< m 2
5
~ s = a b w w =
w W g $
C " ~
t>e r f -g1._ m vXC JCO
- we=
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s -
a a
2 e m. _.
I C .-
nut _eCh 5/2/E2 -_ - . - - - .
m 1
i l
aeea
- e s
e . >-
e ~ o m
" o i A w
I
~
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m o m
, / e a E f A
~'
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;g "-
w
= -
/ O m v ,.
_ z= u
/ . <
. n, z m
- =
S =
. ? - m N_ &
m m 4
F O-w w _
.a
\\ s s
a
< ~
=
< c.
a
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=
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- u e O
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a n
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rsAxtrtoes s rtESS .^.R.EA nie s,,s,1 -
l is Himt EAn is litu G E pia
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hsAWit11AM Sr*4rss \
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i F
,LOCATI0li 0F SilAFT AND lilNGE STRAI N GAllGES 1A, 28, AND 2A, 2n !
3 C
eu -
-(9 .
IO '
IT '
(
.m m. _ : .c.. . z. * ~ ~ '- ...., _ _ _
~
Max Principsi stress Tap Isrface
-h- Test te. sulks I
w- ,
- I -
,a - -
n- .
g
#* i i i to 1:
. tista.=cs firs Car 2871188 (I2:1883
- 1. STRESSES ALONG SECTION A-A .
- 2. IM?ACT VELOCITY OF D.0 RADS /SEC. -
l
. cn r?. ~ :cw.:m=
ra:x::rn nn:: ::nI:nT:: 5 n:x:
- t=: , :::: w.1 : :n c n:n at=::= un I
{
W..-
1 4
n..
' n u t. _e C_n
= u' n_ _ c _ - - - - - - - . , _ . , - . . , _ _ _ _ _ _ _ _ - - - - ,rm-- -- -_
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