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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR05000010/LER-1997-001-06, :on 970210,120 Volt Circuit Breaker Supplying Alternate Iodine & Particulate Sampling Sys Tripped.Caused by Mgt Deficiency.Disconnected Auxiliary Loads,Restricted Use of Outlets & Verified Continued Sys Operation1997-03-12012 March 1997
- on 970210,120 Volt Circuit Breaker Supplying Alternate Iodine & Particulate Sampling Sys Tripped.Caused by Mgt Deficiency.Disconnected Auxiliary Loads,Restricted Use of Outlets & Verified Continued Sys Operation
05000237/LER-1996-003-01, :on 960219,EP DGA-12 Failed to Provide Cro Action to Allow CR Air Filtration Unit to Operate as required.DGA-12 Revised to Give Guidance on Restoration of CR HVAC in Pressurization Mode1996-03-14014 March 1996
- on 960219,EP DGA-12 Failed to Provide Cro Action to Allow CR Air Filtration Unit to Operate as required.DGA-12 Revised to Give Guidance on Restoration of CR HVAC in Pressurization Mode
05000249/LER-1992-004-02, :on 920122,calculations Indicated Min Required 4 Kv Safety Bus Voltages for Buses 33-1 & 34-1 to Be 3832 Volts & 3792 Volts,Respectively.Caused by Inadequate Design Control.Mods Designed & Installed1992-02-21021 February 1992
- on 920122,calculations Indicated Min Required 4 Kv Safety Bus Voltages for Buses 33-1 & 34-1 to Be 3832 Volts & 3792 Volts,Respectively.Caused by Inadequate Design Control.Mods Designed & Installed
05000237/LER-1990-007-09, :on 900802,unplanned Primary Containment Group V Isolation Occurred.Cause Not Identified.Relay Circuitry & Sys Parameters Investigated Prior to Resetting Isolation Signal & Visual Observation Made1990-08-28028 August 1990
- on 900802,unplanned Primary Containment Group V Isolation Occurred.Cause Not Identified.Relay Circuitry & Sys Parameters Investigated Prior to Resetting Isolation Signal & Visual Observation Made
05000237/LER-1989-032, :on 891210,as-left Type B & C Local Leak Rate Test Data Exceeded Tech Spec Limit of 488.452 Std Cubic Ft/H.Caused by Degradation of Telephone & Electrical Conduit Penetration Seal.Conduit Cut & Capped1989-12-29029 December 1989
- on 891210,as-left Type B & C Local Leak Rate Test Data Exceeded Tech Spec Limit of 488.452 Std Cubic Ft/H.Caused by Degradation of Telephone & Electrical Conduit Penetration Seal.Conduit Cut & Capped
05000237/LER-1989-001-05, :on 890121,unanticipated Reactor Scram & Group II Primary Containment Isolation Occurred Due to Loss of Reactor Protection Sys Motor Generator Sets.Caused by Procedural Deficiency.Procedure Reviewed1989-02-21021 February 1989
- on 890121,unanticipated Reactor Scram & Group II Primary Containment Isolation Occurred Due to Loss of Reactor Protection Sys Motor Generator Sets.Caused by Procedural Deficiency.Procedure Reviewed
05000249/LER-1988-009-03, :on 880505,Groups II & III Primary Containment Isolation Occurred During Shutdown for Scheduled Refueling Outage.Caused by Mgt Deficiency for Failure to Implement Corrective Actions from Previous Event1988-06-0303 June 1988
- on 880505,Groups II & III Primary Containment Isolation Occurred During Shutdown for Scheduled Refueling Outage.Caused by Mgt Deficiency for Failure to Implement Corrective Actions from Previous Event
05000237/LER-1988-007-06, :on 880504,discovered That CRD Scram Testing Surveillance Interval Exceeded.Caused by Scram Testing Procedure Deficiency.Strip Chart Event Recorder Wired, Utilizing Wire plug-in Jacks1988-06-0101 June 1988
- on 880504,discovered That CRD Scram Testing Surveillance Interval Exceeded.Caused by Scram Testing Procedure Deficiency.Strip Chart Event Recorder Wired, Utilizing Wire plug-in Jacks
05000237/LER-1988-006-07, :on 880429,primary Nitrogen Inerting Makeup Sys Line Kinked & Cracked.Caused by Personnel Error & Rigging Procedure Deficiency.Rigging Procedure Improvement Initiated & Addl Rigging Operations Training Planned1988-05-25025 May 1988
- on 880429,primary Nitrogen Inerting Makeup Sys Line Kinked & Cracked.Caused by Personnel Error & Rigging Procedure Deficiency.Rigging Procedure Improvement Initiated & Addl Rigging Operations Training Planned
05000237/LER-1988-008-04, :on 880506,reactor Bldg Ventilation Sys Automatically Tripped & Standby Gas Treatment Train Auto Started.Caused by Error in Chemistry Procedure 2000-5. Radiation Source Removed from Monitor1988-05-25025 May 1988
- on 880506,reactor Bldg Ventilation Sys Automatically Tripped & Standby Gas Treatment Train Auto Started.Caused by Error in Chemistry Procedure 2000-5. Radiation Source Removed from Monitor
05000237/LER-1988-005-06, :on 880426,reactor Bldg Ventilation Sys (RBVS) Tripped,Initiating Standby Gas Treatment a Sys Train.Caused by Spurious Instantaneous Spike on Unit 2 Rbvs.Calibr Check of Unit 2 RBVS Performed Satisfactorily1988-05-17017 May 1988
- on 880426,reactor Bldg Ventilation Sys (RBVS) Tripped,Initiating Standby Gas Treatment a Sys Train.Caused by Spurious Instantaneous Spike on Unit 2 Rbvs.Calibr Check of Unit 2 RBVS Performed Satisfactorily
05000249/LER-1988-010-01, :on 880424,main Steam Safety Valve 3-203-4D Opened at Pressure in Excess of Tech Spec 4.6.E Required Setpoint.Caused by Setpoint Drift.Valve Overhauling,Setting & Retesting Prior to Installation Planned1988-05-13013 May 1988
- on 880424,main Steam Safety Valve 3-203-4D Opened at Pressure in Excess of Tech Spec 4.6.E Required Setpoint.Caused by Setpoint Drift.Valve Overhauling,Setting & Retesting Prior to Installation Planned
05000249/LER-1988-011, :on 880412,Group II Primary Containment Isolation Occurred.Caused by Procedural Inadequacy.Power Restored to Monitors & Isolation Reset.Procedural Change Being Investigated1988-05-10010 May 1988
- on 880412,Group II Primary Containment Isolation Occurred.Caused by Procedural Inadequacy.Power Restored to Monitors & Isolation Reset.Procedural Change Being Investigated
05000249/LER-1988-007-04, :on 880413,standby Liquid Control Relief Valves Failed to Open During Performance.Caused by Blockage as Result of Sodium Pentaborate Solution.Procedure Will Be Revised to Indicate Testing of Relief Valves1988-05-0606 May 1988
- on 880413,standby Liquid Control Relief Valves Failed to Open During Performance.Caused by Blockage as Result of Sodium Pentaborate Solution.Procedure Will Be Revised to Indicate Testing of Relief Valves
05000249/LER-1988-004-07, :on 880404,atmosphere Containment & Dilution Purge Check Valve Leaked Exceeding Tech Spec Limit.Cause of Leakage Undetermined.Cause Will Be Determined & Valve Will Be Repaired & Retested1988-04-28028 April 1988
- on 880404,atmosphere Containment & Dilution Purge Check Valve Leaked Exceeding Tech Spec Limit.Cause of Leakage Undetermined.Cause Will Be Determined & Valve Will Be Repaired & Retested
05000237/LER-1988-001, :on 880329,diesel Generator Air Start Piping (Dgasp) Exceeded Code Stress Analysis Allowables Specified in Fsar.Caused by Original Design Deficiency.Mod of Dgasp on Unit 2 & Unit 3 Planned1988-04-25025 April 1988
- on 880329,diesel Generator Air Start Piping (Dgasp) Exceeded Code Stress Analysis Allowables Specified in Fsar.Caused by Original Design Deficiency.Mod of Dgasp on Unit 2 & Unit 3 Planned
05000237/LER-1988-001-12, :on 880329,diesel Generator Air Start Piping Found Outside FSAR Design Criteria.Caused by Original Design Deficiency.Mods Initiated to Modify Piping Sys & Procedure DAP 5-1, Plant Mod Program, Revised1988-04-25025 April 1988
- on 880329,diesel Generator Air Start Piping Found Outside FSAR Design Criteria.Caused by Original Design Deficiency.Mods Initiated to Modify Piping Sys & Procedure DAP 5-1, Plant Mod Program, Revised
05000249/LER-1988-005-03, :on 880326,HPCI Sys Made Inoperable to Facilitate Preplanned Preventive Maint Testing.Caused by Mgt Decision to Perform Preventive Maint Testing of HPCI Turbine Overspeed Trip Sys.Hpci Tripped on 5,015 Rpm1988-04-18018 April 1988
- on 880326,HPCI Sys Made Inoperable to Facilitate Preplanned Preventive Maint Testing.Caused by Mgt Decision to Perform Preventive Maint Testing of HPCI Turbine Overspeed Trip Sys.Hpci Tripped on 5,015 Rpm
05000249/LER-1988-003-06, :on 880323,flued Head Anchor (FHA) Supports Found in Excess of FSAR Design Criteria.Caused by Design & Const Deficiencies.Repairs to FHA to Be Performed Under Mod M12-3-88-201988-04-18018 April 1988
- on 880323,flued Head Anchor (FHA) Supports Found in Excess of FSAR Design Criteria.Caused by Design & Const Deficiencies.Repairs to FHA to Be Performed Under Mod M12-3-88-20
05000237/LER-1988-004-02, :on 880221,inoperable CRD E-9 Found Electrically Armed Due to Personnel Error.Crd Disarmed & out-of-svc Card Hung on Hydraulic Control Unit.Event Discussed W/Personnel at Weekly Meeting1988-03-15015 March 1988
- on 880221,inoperable CRD E-9 Found Electrically Armed Due to Personnel Error.Crd Disarmed & out-of-svc Card Hung on Hydraulic Control Unit.Event Discussed W/Personnel at Weekly Meeting
05000237/LER-1988-002-04, :on 880201,identified Deficiency in Weekly APRM Surveillance Procedure Operational Surveillance,Aprm Rod Block & Scram Function Test.Caused by Procedural Deficiency. Procedure to Be Revised to Ensure Weekly Tests1988-02-29029 February 1988
- on 880201,identified Deficiency in Weekly APRM Surveillance Procedure Operational Surveillance,Aprm Rod Block & Scram Function Test.Caused by Procedural Deficiency. Procedure to Be Revised to Ensure Weekly Tests
05000249/LER-1988-002-01, :on 880113,emergency Diesel Generator Automatically Started W/No Initiation Conditions Indicated in Control Room.Caused by Mgt Deficiency Re Work Activity on Inservice Equipment.Station Policy Issued1988-02-0404 February 1988
- on 880113,emergency Diesel Generator Automatically Started W/No Initiation Conditions Indicated in Control Room.Caused by Mgt Deficiency Re Work Activity on Inservice Equipment.Station Policy Issued
05000249/LER-1988-001, :on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted1988-01-29029 January 1988
- on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted
05000237/LER-1987-010, :on 870317,discovered Embedment Plate for Support Located on Core Spray Sys a Piping Degraded.Six Addl Discrepancies Found During Review of Sys.Caused by Improper Hold Down Clamp Spacings.Repairs Completed1988-01-27027 January 1988
- on 870317,discovered Embedment Plate for Support Located on Core Spray Sys a Piping Degraded.Six Addl Discrepancies Found During Review of Sys.Caused by Improper Hold Down Clamp Spacings.Repairs Completed
05000237/LER-1987-035-02, :on 871219,while Responding to APRM 4 hi-hi Alarm,Operator Noted Number of Operable APRM Downscale Trips in Reactor Protection Sys Insufficient.Caused by Personnel Error.Procedure Changed & Operator Counseled1988-01-0808 January 1988
- on 871219,while Responding to APRM 4 hi-hi Alarm,Operator Noted Number of Operable APRM Downscale Trips in Reactor Protection Sys Insufficient.Caused by Personnel Error.Procedure Changed & Operator Counseled
05000249/LER-1983-002-01, /01T-0:on 830105,leakage from RWCU Test Tap Valve 3-1299-13 Occurred.Caused by Seat Damage of Valve.Valve Replaced During 1983 Refueling Outage1987-12-17017 December 1987 /01T-0:on 830105,leakage from RWCU Test Tap Valve 3-1299-13 Occurred.Caused by Seat Damage of Valve.Valve Replaced During 1983 Refueling Outage 05000237/LER-1982-058-03, /03L-1:on 821226,small Leak Found on RWCU Line 2-1201B Between 2-1201-135B & 2-1201-98B Valves.Caused by IGSCC at heat-affected Zone Near Weld.Degraded Pipe Section Replaced Per Work Package D247801987-12-15015 December 1987 /03L-1:on 821226,small Leak Found on RWCU Line 2-1201B Between 2-1201-135B & 2-1201-98B Valves.Caused by IGSCC at heat-affected Zone Near Weld.Degraded Pipe Section Replaced Per Work Package D24780 05000249/LER-1986-018-01, :on 861014,spurious Group V Containment Isolation Occurred.Caused by Differential Pressure Spikes &/Or Noise Generated by Annubar Flow Instrument.Mod to Isolation Circuitry Implemented1987-12-11011 December 1987
- on 861014,spurious Group V Containment Isolation Occurred.Caused by Differential Pressure Spikes &/Or Noise Generated by Annubar Flow Instrument.Mod to Isolation Circuitry Implemented
05000249/LER-1986-020, :on 861113,spurious Group V Containment Isolation Occurred.Caused by Differential Pressure Spikes &/Or Noise Generated by Annubar Flow Instrument.Mod to Isolation Circuitry Implemented1987-12-11011 December 1987
- on 861113,spurious Group V Containment Isolation Occurred.Caused by Differential Pressure Spikes &/Or Noise Generated by Annubar Flow Instrument.Mod to Isolation Circuitry Implemented
05000249/LER-1987-005, :on 870316,primary Containment Drywell Structural Steel Failed FSAR Design Requirements.Caused by Inadequate Verification of as-built Condition W/Design Prints During Const.Repair & Insp Planned1987-12-0909 December 1987
- on 870316,primary Containment Drywell Structural Steel Failed FSAR Design Requirements.Caused by Inadequate Verification of as-built Condition W/Design Prints During Const.Repair & Insp Planned
05000237/LER-1987-033-02, :on 871110,reactor Bldg Ventilation Sys Sample Holder Contained No Particulate Sample Filter.Caused by Personnel Error.New Procedure Written & Sample Filter Replaced1987-11-30030 November 1987
- on 871110,reactor Bldg Ventilation Sys Sample Holder Contained No Particulate Sample Filter.Caused by Personnel Error.New Procedure Written & Sample Filter Replaced
05000237/LER-1987-030-02, :on 871013 & 16,main Steam Valve Setpoints Found Outside Tech Spec 4.6.E Limits.Caused by Mishandling of Valves During Transport & Setpoint Drift.Valves Will Be Overhauled,Retested,Reset & Procedure Improved1987-11-13013 November 1987
- on 871013 & 16,main Steam Valve Setpoints Found Outside Tech Spec 4.6.E Limits.Caused by Mishandling of Valves During Transport & Setpoint Drift.Valves Will Be Overhauled,Retested,Reset & Procedure Improved
05000237/LER-1987-010-01, :on 870317,core Spray Sys a Analytical Piping Stresses Exceeded FSAR Design Requirements.Caused by Design & Const Errors.Sys Analysis Model Revised to Address as-found Condition1987-11-12012 November 1987
- on 870317,core Spray Sys a Analytical Piping Stresses Exceeded FSAR Design Requirements.Caused by Design & Const Errors.Sys Analysis Model Revised to Address as-found Condition
05000237/LER-1987-019, :on 860811,reactor Scram Occurred from Main Turbine Trip on High Water Level.Caused by Personnel Error. Event Will Be Reviewed W/All Operators & Action Item Record 12-87-10 Issued to BWR Engineering Dept1987-11-0707 November 1987
- on 860811,reactor Scram Occurred from Main Turbine Trip on High Water Level.Caused by Personnel Error. Event Will Be Reviewed W/All Operators & Action Item Record 12-87-10 Issued to BWR Engineering Dept
05000237/LER-1987-001-02, :on 870102,ultrasonic Testing Showed Through Wall Indications on Weld PD5-D20.On 870121.circumferential Indication Through Wall of RWCU Sys 8 Weld 8-12 Discovered. Caused by Igscc.Evaluation Performed & Welds Repaired1987-11-0707 November 1987
- on 870102,ultrasonic Testing Showed Through Wall Indications on Weld PD5-D20.On 870121.circumferential Indication Through Wall of RWCU Sys 8 Weld 8-12 Discovered. Caused by Igscc.Evaluation Performed & Welds Repaired
05000237/LER-1987-032-02, :on 871020,primary Containment Group I Isolation & Reactor Scram Occurred.Caused by Vibration of Main Steam Line Low Pressure Switch.Isolators Installed to Make Switches Less Susceptible to Vibration1987-11-0505 November 1987
- on 871020,primary Containment Group I Isolation & Reactor Scram Occurred.Caused by Vibration of Main Steam Line Low Pressure Switch.Isolators Installed to Make Switches Less Susceptible to Vibration
05000237/LER-1987-034-01, :on 871016,discovered That Nonconservative Core Thermal Power Calculation Due to Inadequate Calibr Procedure.Rosemount Feedwater Flow Transmitters Recalibr & Listed Procedures to Be Revised1987-11-0404 November 1987
- on 871016,discovered That Nonconservative Core Thermal Power Calculation Due to Inadequate Calibr Procedure.Rosemount Feedwater Flow Transmitters Recalibr & Listed Procedures to Be Revised
05000237/LER-1987-022-02, :on 870818,APRM 4 Generated Spurious hi-hi Signal Which Tripped Reactor Protection Sys Channel.Caused by Administrative Deficiency.Operating Order & Procedure Changes Alerting Operators Issued1987-09-15015 September 1987
- on 870818,APRM 4 Generated Spurious hi-hi Signal Which Tripped Reactor Protection Sys Channel.Caused by Administrative Deficiency.Operating Order & Procedure Changes Alerting Operators Issued
05000249/LER-1984-019-02, :on 841024,determined That Two Flow Makeup Boxes Used for Local Leak Rate Testing Out of Calibr & Should Have Been Added to as-found Type B & C Leakage Rate. Caused by Improper Calibr Due to Box Design1987-08-0303 August 1987
- on 841024,determined That Two Flow Makeup Boxes Used for Local Leak Rate Testing Out of Calibr & Should Have Been Added to as-found Type B & C Leakage Rate. Caused by Improper Calibr Due to Box Design
05000249/LER-1987-008, :on 870321,reactor Scrammed on Low Reactor Water Level.Caused by Procedural Inadequacy in Conjunction W/Component Failure.Procedure Revised to Add Provision for Opening Feedwater Pump Discharge Valves1987-04-29029 April 1987
- on 870321,reactor Scrammed on Low Reactor Water Level.Caused by Procedural Inadequacy in Conjunction W/Component Failure.Procedure Revised to Add Provision for Opening Feedwater Pump Discharge Valves
05000237/LER-1987-006-02, :on 870228,during Refueling,Reactor Scram & Group II Isolation Occurred.Caused by Trip at Bus 28 of Main Feed Breaker of Motor Control Ctr (MCC) 28-2.Reactor Protection Sys Bus 2B Transferred to MCC 25-21987-03-30030 March 1987
- on 870228,during Refueling,Reactor Scram & Group II Isolation Occurred.Caused by Trip at Bus 28 of Main Feed Breaker of Motor Control Ctr (MCC) 28-2.Reactor Protection Sys Bus 2B Transferred to MCC 25-2
05000237/LER-1987-007-02, :on 870227,drywell Structural Steel Radial Beam Discovered Outside Design Requirements.Caused by Reinstallation Deficiency.Beam Repair Completed on 870314, Per Engineering Change Notice D875-091987-03-25025 March 1987
- on 870227,drywell Structural Steel Radial Beam Discovered Outside Design Requirements.Caused by Reinstallation Deficiency.Beam Repair Completed on 870314, Per Engineering Change Notice D875-09
05000249/LER-1987-002-09, :on 870225,while Unit Operating at 87% Power, HPCI Sys Discovered Inoperable.Caused by Oil Pressure Regulation Valve Failure.Similar Event Occurred on 870313. Limits of PS4 Reset to 60 Psig Increasing1987-03-25025 March 1987
- on 870225,while Unit Operating at 87% Power, HPCI Sys Discovered Inoperable.Caused by Oil Pressure Regulation Valve Failure.Similar Event Occurred on 870313. Limits of PS4 Reset to 60 Psig Increasing
05000249/LER-1987-003-06, :on 870227,shutdown Cooling HX Outlet Temp Found Greater than 212 F.Caused by Operator Failure to Adequately Monitor Reactor Water Temp on Proper Instrumentation.Water Temp Reduced1987-03-24024 March 1987
- on 870227,shutdown Cooling HX Outlet Temp Found Greater than 212 F.Caused by Operator Failure to Adequately Monitor Reactor Water Temp on Proper Instrumentation.Water Temp Reduced
05000237/LER-1987-005-02, :on 870218,w/unit Shut Down for Refueling Outage,Snubbers 1301,1302 & 1303 Failed to Activate or Bleed within Allowable Ranges.Caused by Upgraded Test Acceptance Criteria.Two Mechanical Snubbers Installed1987-03-21021 March 1987
- on 870218,w/unit Shut Down for Refueling Outage,Snubbers 1301,1302 & 1303 Failed to Activate or Bleed within Allowable Ranges.Caused by Upgraded Test Acceptance Criteria.Two Mechanical Snubbers Installed
05000249/LER-1987-001-09, :on 870209,hourly Fire Insp for East End of Unit 2/3 Cable Tunnel Not Established Immediately Per Tech Spec 3.12.C.1.Caused by Operating Personnel Error.Hourly Fire Watch Immediately Established1987-03-0303 March 1987
- on 870209,hourly Fire Insp for East End of Unit 2/3 Cable Tunnel Not Established Immediately Per Tech Spec 3.12.C.1.Caused by Operating Personnel Error.Hourly Fire Watch Immediately Established
05000237/LER-1987-003, :on 870108,primary Containment Drywell Structural Steel Did Not Meet FSAR Design Requirements Due to Inadequate Connections Between Radial & Tangential Beams. Caused by Inadequate Confirmation of Const1987-02-0606 February 1987
- on 870108,primary Containment Drywell Structural Steel Did Not Meet FSAR Design Requirements Due to Inadequate Connections Between Radial & Tangential Beams. Caused by Inadequate Confirmation of Const
05000237/LER-1987-003-01, :on 870108,found Primary Containment Drywell Structural Steel Not Meeting FSAR Design Requirements.Caused by Inadequate Confirmation of Original Const & as-built Configuration.Repair Program Developed1987-02-0606 February 1987
- on 870108,found Primary Containment Drywell Structural Steel Not Meeting FSAR Design Requirements.Caused by Inadequate Confirmation of Original Const & as-built Configuration.Repair Program Developed
05000249/LER-1986-026-03, :on 860911,drywell Oxygen Content Exceeded 4% for Greater than 24 H W/Mode Switch in Run.Caused by Personnel Error.Operating Procedures DOP-1600-5 & DOP 1600-6 Revised & Guidance Under Development1987-02-0505 February 1987
- on 860911,drywell Oxygen Content Exceeded 4% for Greater than 24 H W/Mode Switch in Run.Caused by Personnel Error.Operating Procedures DOP-1600-5 & DOP 1600-6 Revised & Guidance Under Development
05000237/LER-1987-004-01, :on 870108,primary Containment Type B & C Local Leak Rate Test Limit Exceeded Due to Excessive Leakage Through Isolation Valve.Valve Will Be Repaired & Retested Per Dts 1600-1 When Cause Determined1987-01-29029 January 1987
- on 870108,primary Containment Type B & C Local Leak Rate Test Limit Exceeded Due to Excessive Leakage Through Isolation Valve.Valve Will Be Repaired & Retested Per Dts 1600-1 When Cause Determined
1997-03-12
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17191B4751999-10-0101 October 1999 Safety Evaluation Supporting Amends 175 & 171 to Licenses DPR-19 & DPR-25,respectively ML17191B4721999-10-0101 October 1999 Safety Evaluation Supporting Amends 174 & 170 to Licenses DPR-19 & DPR-25,respectively ML20249C8491999-09-30030 September 1999 1999 Third Quarter Rept of Completed Changes,Tests & Experiments Evaluated,Per 10CFR50.59(b)(2), for Dresden Nuclear Power Station. with ML17191B4591999-09-23023 September 1999 Safety Evaluation Supporting Amends 173 & 169 to Licenses DPR-19 & DPR-25,respectively 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 ML20210R6081999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Dresden Nuclear Power,Units 1,2 & 3.With ML17191B4101999-07-27027 July 1999 Safety Evaluation Accepting Alternative to Licenses NPF-72, NPF-77,NPF-37,NPF-66,NPF-19,NPF-25,NPF-11,NPF-18,DPR-29 & DPR-30,respectively ML20210D3071999-06-30030 June 1999 Corrected Page 8 to MOR for June 1999 for DNPS Unit 3 ML20209E1291999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML20209J3481999-06-30030 June 1999 1999 Second Quarter Rept of Completed Changes,Tests & Experiments, Per 10CFR50.59.With ML20207G3871999-06-0404 June 1999 Safety Evaluation Supporting Amend 168 to License DPR-25 ML20195G6381999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.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 ML17191B3541999-05-18018 May 1999 Safety Evaluation Concluding That Licensee Proposal to Reexamine Flaw According to Guidelines & Schedule of BWRVIP-41 Acceptable.Concludes That Licensee Leakage Evaluation Adequate ML20206N2821999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML17191B3081999-04-0101 April 1999 Safety Evaluation of Topical Rept NPSR-0111,rev 1, BWR Transient Analysis Methods. Rept Acceptable for Inclusion in Licensing Submittals ML20206B1901999-03-31031 March 1999 First Quarter Rept of Completed Changes,Tests & Experiments Per 10CFR50.59 for Dresden Nuclear Power Station. with ML20205N7491999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML17191B2811999-03-16016 March 1999 Safety Evaluation Supporting Amends 172,167,184,181,132 & 117 to Licenses DPR-19,DPR-25,DPR-29,DPR-30,NPF-11 & NPF-18, Respectively ML20207M6921999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced ML17191B2391999-02-16016 February 1999 Safety Evaluation Supporting Amends 171 & 166 to Licenses DPR-19 & DPR-25,respectively ML17191B2371999-02-0808 February 1999 Safety Evaluation Supporting Amends 170,165,183 & 180 to Licenses DPR-19,DPR-25,DPR-29 & DPR-30,respectively ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc ML20199D3261998-12-31031 December 1998 10CFR50.59 SER for 1998-04 Quarter, of Changes,Tests & Experiments.With ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with ML20199C8951998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Dnps,Units 1,2 & 3 ML20197G8591998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Dresden Nuclear Power Station.With ML20196J0061998-11-19019 November 1998 Rev 66 to Topical Rept CE-1-A, QA Program 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 ML20195D2861998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Dresden Nuclear Power Station.With ML17191A9011998-10-0101 October 1998 Safety Evaluation Concluding That Ceco Adequately Addressed Actions Requested in GL 95-07 Issued on 950817 ML20154N4131998-09-30030 September 1998 1998 Third Quarter 10CFR50.59 Rept, for Dresden Nuclear Power Station of Completed Changes,Tests & Experiments ML20154L3681998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML17191A8891998-09-23023 September 1998 Safety Evaluation Accepting Continued Operation for Plant, Unit 3 Without Repair,For Four Fuel Cycles ML20153C5061998-09-21021 September 1998 SER Accepting Qualified Unit 1 Supervisor Initial & Continuing Training Program for Dresden Nuclear Power Station,Unit 1 ML20151Y2711998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Dresden Nuclear Power Station,Units 1,2 & 3.With ML17191A8831998-08-27027 August 1998 Safety Evaluation Accepting Purge & Vent Strategy for Primary Containment Hydrogen Control at Plant ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20237A1341998-08-0707 August 1998 Safety Evaluation Supporting Amend 163 to License DPR-25 ML17191A8231998-08-0505 August 1998 SER Related to Reactor Pressure Vessel Head Stud Flaw Tolerance Evaluation for Dresden Nuclear Power Station,Unit 2 ML20237A7161998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Dresden Nuclear Power Station,Units 1,2 & 3 ML20236M6041998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Dresden Nuclear Power Station,Units 1,2 & 3 ML20236Q5851998-06-30030 June 1998 1998 Second Quarter 10CFR50.59 Rept, for Dresden Nuclear Power Station of Completed Changes,Tests & Experiments ML20236T8331998-06-30030 June 1998 COLR for Dresden Station Unit 3,Cycle 15 ML20236F8131998-06-30030 June 1998 Rev 0 to Defueled SAR Dresden Nuclear Power Station Unit 1 Commonwealth Edison Co ML20236T8391998-06-30030 June 1998 Rev 1 to EMF-96-141, Dresden Unit 3 Cycle 15 Reload Analysis Rept ML20248M3021998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Dresden Nuclear Power Station,Units 1,2 & 3 ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted ML20247F3391998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Dresden Nuclear Power Station,Units 1,2 & 3 1999-09-30
[Table view] |
text
-
- - Y*
LICENSEE EVENT REPORT (LER)~
Fac11 Hy Nam-.: ( 1) ocket Number (2)
~I Page (3) 0 I SI 0 I 0 I 0 I 21 41 9 1 I of I 1 I Dresden Nuclear Power Stat;on. un;t 3 6
Tltle (4)
LLRT Intervals D;scovered ;n Excess of 24 Months on an lnd;v;dua) Test Basis Due to Procedural Deficienc~
Event Date (5)
LER Number ( 6)
Reoort Date l7l Other Facilities Involved C8l Month Day Year Year I.II. Sequential ~~~ Revision Month Day Year Facility Names I Docket Number(s)
- Dresden Unit 2 ol SI OI OI ol 21 31 7 o I 1 o Is 8 I 8 8 I 8 --- o I o I 1 -- olo-o I 1 21 9 81 8 N/A OI SI ol ol ol I I OPERATING
.THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR MODE (9)
(Check one or more of the following) (11}
N 20.402(b) 20.405(c)
SO. 73(a)(2)( iv)
- 73. 71(b).
POWER
- 20. 405(a)( 1 )( i)
S0.36(c)( 1)
SO. 73(a)(2)(v)
- 73. 7 l(c)
LEVEL o I 9 I 20.405(a)(l)(ii) -
S0.36(c)(2)
S0.73(a)(2)(vi;)
_Other (Specify
( 10) 6 20.40S(a)(l)(iii)..L S0.73(a)(2)(i)
S0.73(a)(2)(vi;;)(A)
- n Abstract
~~~~~~~~~~~*~~~~~~~~~~~~*~~~ _. 20. 40S (a)( 1 )( iv)
S0.73(a)(2)(;i) -
S0.73(a)(2)(vii;)(B) below and in
~~~~~~~~~~~~~~~~~~~~~~~~~~ _
20.405(a)(l)(v) 50.73(a)(2)(i;i) -
S0.73(a)(2)(x)
Text)
LICENSEE CONTACT FOR THIS LER ( 12}
Name TELEPHONE NUMBER A~EA CODE Jerrv Lizalek. Technical Staff Ena;neer. X-421 8 I 1 I S 91 41 21 -I 21 91 21 o COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANUFAC-REPORT AB.LE ~~~~~~~ CAUSE SYSTEM COMPONENT MANUFAC-REPORTABLE ~~~~~~
TUR ER TO NPRDS
~~~~~~~
TURER TO NPRDS
~~~~~~
I I I I I I
~~~~~~~
I I
I I
I
~~~~~
I I I I I I 1111111 I*
I I
1 I
"/ "/"/"/"/
SUPPLEMENTAL REPORT EXPECTED (14)
Expected Month I Da~ I Year Subm;ss;on
- - IYes (If ves ccrnolete EXPECTED SUBMISSION DATE)
I I I I I ABSTRACT (Limit to 1400 spaces, ;.e, approximately fifteen s;ngle-space typewritten l;nes) (16)
On January S, 1988, w;th Unit 3 at 9S.8% rated thermal power, wh;le reviewing past Local Leak Rate.Testing (LLRT) records, 1t was discovered that 41 individual type "B" and "C" LLRTs performed during the previous
- refueling outage had individual testing intervals in excess of 24 months.
As 10 CFR SO Append;x*J requires that this testing be perfonned,on intervals no greater than 24 months, concern was raised as to whether compliance with 10 CFR SO Appendh J had been demonstrated.
The. root cause was attributed to a procedural deficiency within the scheduJ;ng of the station survei1lance track;ng program; ;t has been the station's practice to enter the test date of the last ;ndiv;dual LLRT as the entire surveillance completion date.
Corrective actions ;ncluded performing as many of the overdue LLRTs as poss;ble during reactor operation
- and upgrading the track;ng program to verify the LLRT intervals on an ind;v;dual test basis in the future.
Management changes were also implemented regard;ng admin;stration of the surveillance program. A schedular exempt;on was requested fran the Office of Nuclear Reactor *Regulat;on (NRR) for the remci;ning untested valves. Further review revealed that 28 individual Type "B" and "C" LLRTs perfonned during a past Unit 2 refuenng outage had ;ndrvidual testing ;ntervals in excess of 24 months, although current compliance was verified. Safety s;gnHkance is min;ma1 s;nce a primary contairvnent Integrated Leak Rate Test (type "A" test) was performed on Unit 2 and un;t 3 pr;or to startup w;th sat;sfactory results.* A prev;ous event involving a surve;11ance program procedure def;c;ency ;s reported by LER 87-18 on Docket #OS0249.
8802080228 880129 PDR ADOCK 05000249 S
PDR
"'--------------=L=I=CE EVENT REPORT LER FACIL.ITY NAME ( l)
DOCKET NUMBER (2)
TEXT CONTINUATIO LER NUMBER 6 Page (3)
Dresden Nuclear Power Station Unit 3 Year
~~~ Sequential ~~~
Ill Number Ill 0
5 0
0 0
2 4 9 8 8
0 0
Revision Number 0
0 0 2 OF l 6 TEXT PLANT ANO SYSTEM IDENTIFICATION:
General Electric Boiling Water Reactor - 2527 foU.Jt rated core thermal power.
Energy Industry Identificaton System (EIIS) codes are identified in the text as [XX].
Nuclear Conmitment Tracking System* (NTS) tracking code numbers are identified in the text as (XX).
EVENT IDENTIFICATION:
While reviewing past Local Leak Rate Testing (LLRT) records it was discovered that 41 individual Type 118 11 and "C" LLRTs perfornied during the previous Unit 3 refueling outage had individual testing intervals in exces~* of 24 months due to a surveillance program procedure deficiency. Further review revealed that 28.
individual Type 118 11 and "C" LLRTs perfomied during a previous Unit 2 refueling outage had individual testing intervals in excess of 24 months, although current compliance was verified.
A.
CONDITIONS PRIOR TO EVENT
Unit: 3 Event Date:. January 5, 1988 Reactor Mode:
N Mode Name: Run Reactor Coolant System (RCS) Pressure: 997 psig B.
DESCRIPTION OF EVENT
Event Time:
1000.hours Power Level: 95.8%
On January 5, 1988 with Unit 3 at 95.8% rated thennal power, a review of LLRT.records for Type 118 11 and "C" testing performed during the previous Unit 3 refueling outage (End of Cycle 9) was conducted. This.review revealed that.the LLRT program had begun on September 27, 1985 and was completed on August 13, 1986. Therefore, 41 individual LLRTs were discovered to have individual testing intervals in excess of 2 years~ As 10 CFR 50 Appendix J requires testing intervals of no greater than 24 manths, concern was raised as to whether compliance with 10 CFR 50 Appendix J had been demonstrated. These individual LLRTs which were currently in excess of the 2 year interval are listed in Table *1.
This review of LLRT testing history was a result of investigation into a previous related surveillance interval problem.
On September 30, 1987, it was discovered that a fire barrier penetration surveillance had exceeded its 18 month Technical Specification surveillance interval in 1985, although current compliance was verified. The root cause of the fire barrier penetration surveillance problem was found to be a procedural deficiency within the surveillance tracking program.
The problem identified with the tracking program resulted from certain surveillances being misclassified as due on a refueling outage basis rather than on a fixed-interval basis.
As part of the corrective actions for this event, which was reported to the Nuclear Regulatory Conmission (NRC) on Licensee Event Report #87-018 on Docket #050249, a review was performed in order.to properly classify these types of surveillances *.
LIC EVENT REPORT LER TEXT CONTINUATIO FAi:l~ITY NAME ( 1)
Do* ET NUMBER (2)
LER NUMBER Page Pl Year
~~~ Sequential
~~~ Revision Dresden Nuclear Power Station Unit 3 0
5 0
0 0
2 4 9 8 8
Ill Number Ill NUIOOer 0
0 0
0 0 3 OF TEXT Further review revealed that a schedular exemption had been previously granted by the NRC concerning a past Unit 3 refueling outage LLRT program.
Upon perfonning a detailed review of this previous exemption and 10 CFR 50 Appendix J, the current LLRT concerns were raised..
Upon discovering that the 41 individual Unit 3 LLRT dates had exceeded a 24 month interval, the Corporate Office was contacted for assistance in detennining if a violation had occurred.
The Dresden site Senior Resident Inspec~or (USNRC, Region III) was also contacted and notified of this infonnation.
Further review revealed that 28 individual Type 118 11 and "C" LLRTs performed during a past Unit 2 refueling outage had individual testing intervals in excess of 24 months, although current compliance was verified. These individual LLRTs are listed in Table 2.
C.
APPARENT CAUSE OF EVENT:
Local Leak Rate Tests at Dresden Station are perfonned under the direction of five separate_
Technical Staff surveillances 1 ist_ed in Table 3.
Leak Rate Tests of similar type are grouped together and are controlled under one of these five surveillances.
As many as 85 Local Leak Rate Tests are listed under one survei_llance*and scheduling of these surveillances is tracked by the stat ion computer survei 11 ance program.
The completion of any of these Local leak Rate surveillances can take the duration of the entire refue.ling outage due to availability of equipment for test1rig and conflicts with inaintenance. It has_been the past practice of Dresden Station to enter the date of the last test perfonned as the completion date of the affected surveillance. This single completion date rather than individual test completion dates would then be used by the computer program to schedule the next surveillance interval for all of the penetrations affected. Since all of the individual tests listed under the surveillance were completed before the end of the refueling outage, and primary containment isolation volumes were not required to be operable until the end of the refueling outage, it has been past station practice to use the last individual test date as the representative completion date to be used for surveillance tracking.
Although the 69 primary containment isolation pathways (41 currently for Unit 3 and 28 previously for Unit 2) listed in Table 1 and_ Table 2 have individual test dates that have exceeded the 2 year 10 CFR 50 Appendix J test interval, the completion dates listed for each affected surveillance in the computer tracking system do not exceed the 2 year limit. Therefore, the cause has been attributed to a surveillance program procedure deficiency.
The Dresden Unit 2 and Unit 3 Technical Specifications Section 4.7.A.2. require perfonnance of Leak Rate Testing in accordance with 10 CFR 50 Appendix J. Therefore, this report is submitted in accordance with 10 CFR 50 (a)(2)(i)(B), which requires the reporting of any operation or condition prohibited.by the Technical Specifications.*
D.
SAFETY ANALYSIS OF EVENT:
The safety significance of the Unit 3 discrepancy is mitigated by the following factors:
- 1.
On July 26, 1986, a successful primary containment Integrated Leak Rate Test (ILRT) was performed on Unit 3.
The total "as left" leakage was.5874 weight %/day and included Local Leak Rate Test results for all systems not vented. This was well within the Technical Specification maximum allowable limit of 1.6 weight %/day. Twenty-nine of the 41 penetrations affected were vented during the ILRT.
6
LIC~ EVENT REPORT (LER) TEXT CONTINUATIO~
FACH,ITY NAME ( 1)
DOCKET NUMBER (2)
LER NUMBER (6)
Page Pl Year
~~~ Sequential ~~~ Revision Dresden Nuclear Power Ill Nurrber Ill Number Station. Unit 3 o I 5 I o I o I o I 21 41 9 8-1 8 o I o I 1 -
o I 0 0 14 OF TEXT
- 2.
A review was perfonned of the "as found" and "as left" leakage rates for the 41 _affected penetr.ations during the last three refueling outages. The results are listed in Table 4.
In general, almost all "as found" leakages for the 41 penetrations were small, with the worst "as found" leakage of 87.9 SCFH found on valves 220-57A and 62A during the 1984 refueling outage.
This is well below the Technical Specification limit of 493. 116 SCFH.
The total "as found" leakage for all 41 penetrations is listed at the bottom of Table 3 for each outage.
The worst total "as found" leakage was 214.25 SCFH which was also below the Technical Specification 1 imit.
The safety significance of the Unit 2 discrepancy is mitigated by the following factors:
The total "as left" leakage was.4732 weight %/day and included Local Leak Rate Test results for all non-vented systems. This was well within the Technical Specification.maximum allowable.limit of 1.6 weight %/day.
Fifteen of the 28 penetrations affected were vented during the ILRT.
- 2.
A review was perfonned of the "as found" leakage rates for the 28 penetrations affected during the end of Cycle 10 refueling outage. It was determined that the "as found" maximum pathway leakage was 185.008 SCFH; this is well below the Technical* Specification limit of 493. 116 SCFH.
Additionally, the "as found" minimum pathway leakage for all Type "8" and "C" LLRTs was determined to be 236.789 SCFH.
This is also below the Technical Specification limit of 493. ll6 SCFH.
- 3.
Currently, Dresden Station is in compliance with the required Appendix J 24 month test interval for all Unit 2 Type "8" and "C" primary containment penetrations on an individual basis.
E.
CORRECTIVE ACTIONS
The inmediate corrective actions included tabulating the already past due 41 Unit 3 LLRTs (Table 1) and determining which of these could have LLRTs performed during reactor power operation.
Subsequent to a review, it was determined that 22 of the 41 past due surveillances could be tested.
Table 5 identifies which primary containment/valve volumes could be tested. These volumes. were inmediately Local Leak Rate Tested and the identified leakage can be seen on Table 5.
- The 19 remaining volumes could not be tested at reactor power for the reasons stated in Table 6.
Secondly, a review was performed to determine which individual LLRTs would becane overdue if not performed prior to the scheduled.March 26, 1988 outage start date. It was determined that 43
- - additional volumes (Table 7) would becane overdue prior to the March 26, 1988 outage start date. A review of these volumes revealed that 37 could be Local Leak Rate Tested during power operation.
These were inmediately tested and their _leakage can be seen on Table 8. The 6 remaining volumes could not be tested at reactor power for the reasons stated in Table 9.
A total of 25 penetrations/volumes that were overdue or would be overdue prior to the March 26, 1988 outage start date could not be tested. A request for schedular exemption from the 10 CFR 50, Appendix J, Type "B" and "C" test interval for these 25 tests was submitted on January 10, 1988 in a letter from the Dresden Nuclear Licensing Administrator to the Director of Nuclear Reactor Regulation.
In addition to these short term corrective actions, the following long term corrective actions have been initiated to restore full confidence in the Technical Specification Surveillance Program.
They are as follows:
11 6
EVENT REPORT LER FACl~ITY NAME (1)
ET NUMBER (2)
TEXT CONTINUATI0~1 LER NUMBER (
Page (3)
Year
~~~ Sequential ~~~ Revision Dresden Nuclear Power
5 0
0 0
2 4 9 8 8
"//'/
Number
"//'/
Number 0
0 0
0 0 5 OF TEXT
- 1.
A Conrnonwealth Edison Task Force has been assigned to perfonn an independent review of the surveillance program's adequacy. Appropriate improvements wi 11 then be implemented (249-200-88-00101).
- 2.
An INPO assistance visit has been scheduled for February 8, 1988.
The purpose of this visit is to provide the station and Task Force with INPO industry input on effective surveillance programs (249-200-88-00102).
- 3.
In order to improve the effectiveness and control of the surveillance program, management changes concerning its administrati.on were initiated. Overall control of surveillance tracking will be assigned to the Work Planning Department, which is headed by an Assistant Superintendent. Scheduling of surveillances will be integrated into daily work. scheduling meetings in order to ensure they are perfonned in a timely manner.
Each working group will be held responsible to comply with appropriate scheduling requirements (249-200-88-00103).
F.
PREVIOUS OCCURRENCE:
Licensee Event Report #87-018 on Docket #050249, Fire Stop 18 Month Surveillance Interval Exceeded Due to Procedural Deficiency.
As stated in Section B., this event also involved a surveillance program deficiency. Corrective actions involving reclassification of the surveillances as due on a fixed-interval basis as opposed to an operating cycle basis did not correct the underlying problem with Local Leak Rate Testing procedures in t.hat they were logged as completed upon performance of the entire group of tests.
G.
COMPONENT FAILURE DATA
Since no component failures were associated with this event, this section is not applicable.
l 6
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1-7-88 Iii Penetration Bellows X-138 12-8-85 x
1-7-88 1-i PJ Penetration Bellows x..:141
.X 1-7-88 Cll Penetration Bellows 12-8-85.
1-i X-149A x
1-7-88 Penetration Bellows X-149B 12-8-85 x
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. CRD Hatch 11/26/84 1/16/87 X-313A last Torus Train 9/2184 11/24/86
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Dresden Nuclear Power Station Unit 3 TEXT DTS-250-1 DTS-1600-1 DTS-1600-2 DTS-1600-4*
DTS-1600-15 TEXT CONTINUATIO LER NUMBER b Year
~~~ Sequential
"/"/'/
Number
~~~ Revision
"/"/"/
Number Page (3) 0 5
0 0
0 2 4 9 8 8
0 0
0 0
0 9
OF 1 6 TABLE 3 LOCAL LEAK RATE TEST SURVEILLANCE PROCEDURES Main Steam Isolation Valve Local Leak Rate (Dry) Test
. Local Leak Rate Testing of Primary Containment Isolation Valves Local Leak Rate Testing Procedure Bellows Seal Penetrations Local Leak Rate Testing For Electrical Penetrations Local Leak Rate Testing of Double Casketed Seals
FACILITY NAME ( 1)
Dresden Nuclear Power Station Unit 3 TEXT VALVES East Torus Drain West Torus Drain 4720 & 4721 4722 & Check 220-57A & 62A 1101-1 & 15 1101-1 & 16 1501-18A & 19A 1501-20A & 38A 1501-18B & 19B 1501-20B & 38B 1402-4A,8A,25A,36A 1402-4B,8B,25B,36B 1301-1 & 2 220-44'& 45 Penetration Bellows X-105A X-105B X-105C.
X-1050 X-106 X-107A X-108A X-lllA X-lllB X-128 X-123 X-124 X-125 X-126 X-138 X-147 X-149A X-149B X-107B X-109A 3702 & 3799-126 1301-17 & 20 220-1 & 2 2301-4 & 5 2301-35 & 36 Drywell Head Hanway TOTAL AS FOUND 0
AS EVENT REPORT LER TEXT CONTINUATIO ET NUMBER (2)
LER NUMBER Year
~~~ Sequential Iii' Number 5
0 0
- 0. 2 4 9 8 8
0 0
TABLE 4 UNIT 3 HISTORICAL LLRT DATA 1982 1984 FOUND AS LEFT AS FOUND AS LEFT SCFH SCFH SCFH SOFH 0
0 0
0 0
0 0
0 0
0 4.14 4.14
.24
.24 3.92 3.92 25.9 25.9 87.9 9.33
.96
.96 li.115 11.115 23.14 23.14 0
0 0
0 1.185 1.36 7.83 7.83 4.05 4.05
.38
.38 0
0 2.48 2.48 0
0
.56
.56 0
0 0
0
.918
.918 0
0 2.15 2.15 0
0 0
0 ALL BELLOW SEALS ALL BELLOW SEALS INITIAL FINAL INITIAL FINAL 4.8 4.8 8.33.
8.33
.289
.289 1.07 1.07
.649
.649 6.49 2.84 7.67 7.67 1.021 1.021 2.12 2.12 0
0 79.82 19.73 0
0 0
0 74.87 SCFH 214.25 SCFH Page Pl
~~~ Revision i'i'i' Number 0
0 1 0 OF 1 6 1985 AS FOUND AS LEFT SCFH SCFH
.1991
.1991
.2109
.2109 2.62 2.62
. 1. 949
- 1. 949 10.85 10.85 6.01 6.01 11.81
. 11. 81 1\\.534 4.534 1.94 1.94
.5917
.5917 1.294 1.294 3.44 3.44 1.821 1.821
.9365
.9365
.033
.033 3.974 3.974 0
0 0
0 0
0 0
0
- 864
.864 0
0 3.024 3.024 0
0 0
0 0
0 0
0 0
0 0
0
.432
.'132 0
0
.216
.216 1.08 1.08 0.0 0.0
.o 0
0 0
- 1. 74
- 1. 74 54.328
- 1. 736 3.44 3.44 1.437 1.437 0
0 118. 77 SCFH
FACIL.ITY NAME ( l)
Dresden Nuclear Power Stat ion Unit 3 TEXT LER NUMBER LICE'I EVENT REPORT LER TEXT CONTINUATIO DO ET NUMBER (2)
Year
~~~
Ill Sequential ~~~
Number IT/
Revision Number 0
5 0
0 0
2 4 9 8 8
0 0
Penetration Number X-105A.
X-105B X-105C X-105D X-106 X-107A X-107B X-108A X-109A X-lllA X-lllB X-128 X-123 X-124 X-125 X-126 X-138 X-147 X-149A X-149B X-313A X-313B TABLE 5 UNIT 3 OVERDUE LLRTs THAT HAVE BEEN PERFORMED AS OF 1/5/88 Voluine Being Tested Main Stearn Main Steam Main Stearn Main Steam Main Steam Drain Feedwater Feedwater Iso. Cond. Steam Iso. Cond. Condensate Shutdown Cooling Shutdown Cooling HPCI Steam RBCCW Inlet RBCCW Outlet Vent From DW Vent To OW SBLC Rx head Spray*
Core Spray Core Spray E. Torus Drain Vlvs w. Torus Drain Vlvs 0
0 1
Recent Test Leakage SCFH 3.412 0
0 0
0 0.604 0
0 0
1.9 0
0 0
0 0
0 0.562 0
0.432 0.734
OF l
6
VALVE OR PENETRATION TES'I: PROBLEMS A0-4720 & 4721 A0-4722 & Check Valve 220-57A & 62A 1101-1 & 15 1101-1 & 16 1501-lSA & 19A 1501-20A & 38A 1501-188 & 198 1501-208 & 388 H0-1402-4A, SA, 25A, 36A H0-1402-48, SB, 258, 368 H0-1301-1 & 2 A0-220-44*& 45 Drywall Pneumatic Supply Isolation Drywall Pneumatic Supply Isolaition Feedwater Check Isolation standby Liquid Control Injection Standby Liquid Control Injection LPCI Loop A LPCI Loop A
.LPCI Loop B LPCI Loop B Core Spray Injection Core Spray Injection Isolation Condenser
- steam Supply Recirc Loop Sample Isolation REASON FOR NOT TESTING AT POWER T~sting pathway could cause main steam isolation valves to close.
Testing pathway will cause main steam isolation* valves lo close.
Testing would require isolation of the reaclor feedwater system and a unit shutdown.
Test isolation valves are located in the inerted drywall and are inaccessible during operation.
Test isolation valves are located in the inerted drywall and are inaccessible during operation.
Testing pathway would require entry into a limiting condition of operation.
Testing pathway would require entry into a limiting condition of operation.
Testing pathway would require entry into a limiting concHtlon of operation.
Testing pathway would require entry into*a limiting condition of operation.
Testing pathway would require entry irito a limiting condition of operation. One.test boundary would be pressurized at 1000 psig preventing valid testing.
Testing pathway would require entry into a limiting condition of operation.
One test boundary would be pressurized at 1000 psig preventing valid testing.
Testing pathway would require entry into a limiling condition of operation.
One test boundary would be pressurized at 1000 psig preventing valid testing.
- one test boundary would be pressurized at 1000 psig preventing valid testing.
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VALVE OR PENETRATION TEST PROBLF.HS H0-3702 &
3799-126 RBCCW to Drywell Coolers Drywell Head Hanway Double Gasket Seal A0-1301-17 & 20 Isolation Condenser Vents H0-220-1 & 2 Hain steam Line Drain
. H0-2301..,.4 & 5 HPCI Steam Supply H0-2301-35 & 36 HPCI Pump Suction REASON FOR NOT TESTING AT POWER Testing of pathway results in loss of drywell cooling and could cause a reactor scram.
Test boundary inaccessible during operation.
One test boundary will be pressurized at 1000 psig preventing valid testing.
One test boundary*will.be pressurized at 1000 psig preventing valid testing.
Testing pathway would require entry into a limltlng condition of operation.
Testing pathway would require entry into a limiting condition of operation.
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A0-3-301-156A,157 Scram Discharge Volume Ven ls and Drains 1-11:..86 N
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1-9-88 L'
A0-1599-61,62 Torus to Condenser Drain 1-21-86 y
1-8-88 0
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1-9-88 L'
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y 1-':l z Ill X-202S Electrical Penetrations 3-2-86 y
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1-8-88
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1-8-88 X-204M Electrical Penetrations 3-2~86 y
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'° X-204N Electrical Penetrations 2:
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~
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j 3-2-86.
y 1-9-88 M0-1601-57,58,59 H2 Makeup To Drywall 3..,.16-86 y
1-9-88 X-313A E. Torus Drain 3-18-86 y
1-9-88 A
X3138 w. Torus Drain 3-18-86 y
1-9-88 Ill 0
Ill w
°'
LICE~ EVENT REPORT (LER) TEXT CONTINUATION~
FACH.ITY NAME ( 1)
DOCKET NUMBER (2)
LER NUMBER (6)
Page {3)
Year
~~~ Sequential ~~~ Revision Dresden Nuclear Power
"//"/
Number
"//"/
Number Station. Unit 3 o I 5 I o I o I o I 21 41 9 a I a -
o I o I 1 -
o I 0 1 15 OF 11 6 TEXT TABLE 8
. UNIT 3 PENETRATIONS TESTED SO AS NOT TO EXCEED 24 MONTH INTERVAL Penetration Leakage Number Volume Being Tested SCFH X-200C
.LV Power & Control 8.130 X-201B HV Power 1.016 X-202B CRD Indication 0.103 X-202BB CRD Indicator
- o.o X-202D HV Power o.o X-202F Thermocouples 8.965 X-202J Neutron Monitor o.o X-202N Neutron Monitor o.o X-202Q Instrumentation o.o X-202S CRD Indicators 2.571 X-202W CRD Indicators 3.074 X-203B HV Power o.o X-204A HV Power 2.376 X-204E Neutron Monitor o.o X-204H Neutron Monitor 0.103 X-204M LV Power 6.125 X-204N CRD Indicator o.oo X-204Q..
CRD Indicator 0.561 X-204S LV Power & Control 14.062 X-20SB CRD Indicator o.o X-136A Tip Flange o.o x..:.136B Tip Flange
.082 X-136C Tip Flange 0
X'."1360 Tip Flange 0
X-136E Tip Flange
.661 X-136F Tip Flange 0.0 X-313A E. Torus Drain o.o X-313B w. Torus Drain o.o NA 1S99-61 & 62 3.77 X-126,304 1601-S7, 58, 59 0.421 X-118 2001-5 & 6
.854 X-117 2001-105 & 106
.281 X-202V 2499-lA & 2A o.o X-204B 2499-lB & 2B o.o X-202V 2599-2A & 23A 3.285 X-125,318 2S99-4A & SA 9.235 X-125,318 2599-4B & SB 1.558
LICE
'-r"---------'==
EVENT REPORT LER TEXT CONTINUATION FAC~LITY NAME (1)
Dresden Nuclear Power Station Unit 3 TEXT.
DOCKET NUMBER (2) 0 5
0 0
0 2
LER NUMBER Year
~~~
'Ill 4 9 8 8
TABLE 9 6
Page Sequential
~~~ Revision Number Ill Number 0
0 0
0 1 6 OF LIST OF PATHWAYS NOT YET OVERDUE THAT COULD NOT BE TESTED (UNIT 3)
Valve or Penetration 205-2-4 & Flange 205-2-7 & Flange 3-301-156A, 157 3-301-160A, 161A.
3-301-1568, 1578 3-301-1608, 161B Reactor Head Cooling Reactor Head Cooling Scram Discharge Volume Scram Discharge Volume Scram Discharge Volume Scram Discharge Volume Test boundary inaccessible during operation Test boundary inaccessible during operation Testing of pathway could cause a reactor scram Testing of pathway could cause a reactor scram Testing of pathway could cause a reactor scram Te$ting of pathway could cause a reactor scram p) 1 6
Common~h Edison Dresden Nucl~ower Station R.R. #1 Morris, Illinois 60450 Telephone 815/942-2920 EDE LTR 1188-063 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 January 29, 1988 Licensee Event Report /188-001-0, Docket /1050249 is being submitted as required by Technical Specification 6.6, NUREG 1022 and 10 CFR 50.73(a)(2)(i)(B).
Een' en Station ger Dresden Nuclear Power Station EDE/kj 1 Enclosure cc:
A. Bert Davis, Regional Administrator, Region III File/NRC File/Numerical 0192k
|
---|
|
|
|
05000249/LER-1988-001, :on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted |
- on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) | 05000237/LER-1988-001-12, :on 880329,diesel Generator Air Start Piping Found Outside FSAR Design Criteria.Caused by Original Design Deficiency.Mods Initiated to Modify Piping Sys & Procedure DAP 5-1, Plant Mod Program, Revised |
- on 880329,diesel Generator Air Start Piping Found Outside FSAR Design Criteria.Caused by Original Design Deficiency.Mods Initiated to Modify Piping Sys & Procedure DAP 5-1, Plant Mod Program, Revised
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000237/LER-1988-001, :on 880329,diesel Generator Air Start Piping (Dgasp) Exceeded Code Stress Analysis Allowables Specified in Fsar.Caused by Original Design Deficiency.Mod of Dgasp on Unit 2 & Unit 3 Planned |
- on 880329,diesel Generator Air Start Piping (Dgasp) Exceeded Code Stress Analysis Allowables Specified in Fsar.Caused by Original Design Deficiency.Mod of Dgasp on Unit 2 & Unit 3 Planned
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000237/LER-1988-002-04, :on 880201,identified Deficiency in Weekly APRM Surveillance Procedure Operational Surveillance,Aprm Rod Block & Scram Function Test.Caused by Procedural Deficiency. Procedure to Be Revised to Ensure Weekly Tests |
- on 880201,identified Deficiency in Weekly APRM Surveillance Procedure Operational Surveillance,Aprm Rod Block & Scram Function Test.Caused by Procedural Deficiency. Procedure to Be Revised to Ensure Weekly Tests
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(A) | 05000249/LER-1988-002-01, :on 880113,emergency Diesel Generator Automatically Started W/No Initiation Conditions Indicated in Control Room.Caused by Mgt Deficiency Re Work Activity on Inservice Equipment.Station Policy Issued |
- on 880113,emergency Diesel Generator Automatically Started W/No Initiation Conditions Indicated in Control Room.Caused by Mgt Deficiency Re Work Activity on Inservice Equipment.Station Policy Issued
| 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) | 05000249/LER-1988-003-06, :on 880323,flued Head Anchor (FHA) Supports Found in Excess of FSAR Design Criteria.Caused by Design & Const Deficiencies.Repairs to FHA to Be Performed Under Mod M12-3-88-20 |
- on 880323,flued Head Anchor (FHA) Supports Found in Excess of FSAR Design Criteria.Caused by Design & Const Deficiencies.Repairs to FHA to Be Performed Under Mod M12-3-88-20
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000249/LER-1988-004-07, :on 880404,atmosphere Containment & Dilution Purge Check Valve Leaked Exceeding Tech Spec Limit.Cause of Leakage Undetermined.Cause Will Be Determined & Valve Will Be Repaired & Retested |
- on 880404,atmosphere Containment & Dilution Purge Check Valve Leaked Exceeding Tech Spec Limit.Cause of Leakage Undetermined.Cause Will Be Determined & Valve Will Be Repaired & Retested
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-1988-004-02, :on 880221,inoperable CRD E-9 Found Electrically Armed Due to Personnel Error.Crd Disarmed & out-of-svc Card Hung on Hydraulic Control Unit.Event Discussed W/Personnel at Weekly Meeting |
- on 880221,inoperable CRD E-9 Found Electrically Armed Due to Personnel Error.Crd Disarmed & out-of-svc Card Hung on Hydraulic Control Unit.Event Discussed W/Personnel at Weekly Meeting
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(i)(8) | 05000249/LER-1988-005-03, :on 880326,HPCI Sys Made Inoperable to Facilitate Preplanned Preventive Maint Testing.Caused by Mgt Decision to Perform Preventive Maint Testing of HPCI Turbine Overspeed Trip Sys.Hpci Tripped on 5,015 Rpm |
- on 880326,HPCI Sys Made Inoperable to Facilitate Preplanned Preventive Maint Testing.Caused by Mgt Decision to Perform Preventive Maint Testing of HPCI Turbine Overspeed Trip Sys.Hpci Tripped on 5,015 Rpm
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iii) | 05000237/LER-1988-005-06, :on 880426,reactor Bldg Ventilation Sys (RBVS) Tripped,Initiating Standby Gas Treatment a Sys Train.Caused by Spurious Instantaneous Spike on Unit 2 Rbvs.Calibr Check of Unit 2 RBVS Performed Satisfactorily |
- on 880426,reactor Bldg Ventilation Sys (RBVS) Tripped,Initiating Standby Gas Treatment a Sys Train.Caused by Spurious Instantaneous Spike on Unit 2 Rbvs.Calibr Check of Unit 2 RBVS Performed Satisfactorily
| 10 CFR 50.73(a)(2)(iv), System Actuation | 05000237/LER-1988-006-07, :on 880429,primary Nitrogen Inerting Makeup Sys Line Kinked & Cracked.Caused by Personnel Error & Rigging Procedure Deficiency.Rigging Procedure Improvement Initiated & Addl Rigging Operations Training Planned |
- on 880429,primary Nitrogen Inerting Makeup Sys Line Kinked & Cracked.Caused by Personnel Error & Rigging Procedure Deficiency.Rigging Procedure Improvement Initiated & Addl Rigging Operations Training Planned
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000249/LER-1988-007-04, :on 880413,standby Liquid Control Relief Valves Failed to Open During Performance.Caused by Blockage as Result of Sodium Pentaborate Solution.Procedure Will Be Revised to Indicate Testing of Relief Valves |
- on 880413,standby Liquid Control Relief Valves Failed to Open During Performance.Caused by Blockage as Result of Sodium Pentaborate Solution.Procedure Will Be Revised to Indicate Testing of Relief Valves
| 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000237/LER-1988-007-06, :on 880504,discovered That CRD Scram Testing Surveillance Interval Exceeded.Caused by Scram Testing Procedure Deficiency.Strip Chart Event Recorder Wired, Utilizing Wire plug-in Jacks |
- on 880504,discovered That CRD Scram Testing Surveillance Interval Exceeded.Caused by Scram Testing Procedure Deficiency.Strip Chart Event Recorder Wired, Utilizing Wire plug-in Jacks
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-1988-008-04, :on 880506,reactor Bldg Ventilation Sys Automatically Tripped & Standby Gas Treatment Train Auto Started.Caused by Error in Chemistry Procedure 2000-5. Radiation Source Removed from Monitor |
- on 880506,reactor Bldg Ventilation Sys Automatically Tripped & Standby Gas Treatment Train Auto Started.Caused by Error in Chemistry Procedure 2000-5. Radiation Source Removed from Monitor
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000249/LER-1988-009-03, :on 880505,Groups II & III Primary Containment Isolation Occurred During Shutdown for Scheduled Refueling Outage.Caused by Mgt Deficiency for Failure to Implement Corrective Actions from Previous Event |
- on 880505,Groups II & III Primary Containment Isolation Occurred During Shutdown for Scheduled Refueling Outage.Caused by Mgt Deficiency for Failure to Implement Corrective Actions from Previous Event
| 10 CFR 50.73(a)(2)(iv), System Actuation | 05000249/LER-1988-010-01, :on 880424,main Steam Safety Valve 3-203-4D Opened at Pressure in Excess of Tech Spec 4.6.E Required Setpoint.Caused by Setpoint Drift.Valve Overhauling,Setting & Retesting Prior to Installation Planned |
- on 880424,main Steam Safety Valve 3-203-4D Opened at Pressure in Excess of Tech Spec 4.6.E Required Setpoint.Caused by Setpoint Drift.Valve Overhauling,Setting & Retesting Prior to Installation Planned
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000249/LER-1988-011, :on 880412,Group II Primary Containment Isolation Occurred.Caused by Procedural Inadequacy.Power Restored to Monitors & Isolation Reset.Procedural Change Being Investigated |
- on 880412,Group II Primary Containment Isolation Occurred.Caused by Procedural Inadequacy.Power Restored to Monitors & Isolation Reset.Procedural Change Being Investigated
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(viii)(B) |
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