LER 95-004-00:on 950607,ESFA Rt Occurred During Unit 2 Controlled Shutdown Per TS 3.0.3.Caused by RHR Sys Inoperability.Replaced All SBF-1 Failed Protection Relays on 500 Kv BreakersML18101A809 |
Person / Time |
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Site: |
Salem ![PSEG icon.png](/w/images/c/cc/PSEG_icon.png) |
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Issue date: |
07/07/1995 |
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From: |
Hall W PENNSYLVANIA POWER & LIGHT CO. |
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To: |
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Shared Package |
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ML18101A808 |
List: |
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References |
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LER-95-004-02, LER-95-4-2, NUDOCS 9507110212 |
Download: ML18101A809 (11) |
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Similar Documents at Salem |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:RO)
MONTHYEARML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML18107A5581999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 2.With 991014 Ltr ML18107A5571999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 1.With 991014 Ltr ML18107A5301999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 2.With 990913 Ltr ML18107A5311999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 1.With 990913 ML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A5201999-08-12012 August 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#9) Second Interval,Second Period, First Outage (96RF). ML18107A4811999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 1.With 990813 Ltr ML18107A4821999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 2.With 990813 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A5211999-07-0101 July 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#10) Second Interval,Second Period,Second Outage (99RF). ML18107A4351999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 1.With 990713 Ltr ML18107A4341999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 2.With 990713 Ltr ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3441999-06-0101 June 1999 Interim Part 21 Rept Re Premature Over Voltage Protection Actuation in Circuit Specific Application in Dc Power Supply.Testing & Evaluation Activities Will Be Completed on 990716 ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A3681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 1.With 990611 Ltr ML18107A3721999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 2.With 990611 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A3711999-04-30030 April 1999 Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1 ML18107A3151999-04-30030 April 1999 Submittal-Only Screening Review of Salem Generating Station Individual Plant Exam for External Events (Seismic Portion), Rev 1 ML18107A2991999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 1.With 990514 Ltr ML18107A2971999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 2.With 990514 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18107A2881999-04-0707 April 1999 Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. ML18107A1821999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 1.With 990414 Ltr ML18107A1831999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 2.With 990414 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B1021999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 2.With 990315 Ltr ML18106B1011999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 1.With 990315 Ltr ML18106B0931999-02-25025 February 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0551999-02-0101 February 1999 Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl ML18106B0561999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 2.With 990212 Ltr ML18106B0571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 1.With 990212 Ltr ML20205P1671999-01-31031 January 1999 a POST-PLUME Phase, Federal Participation Exercise ML18106B0441999-01-29029 January 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0251998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Salem Unit 2.With 990115 Ltr 1999-09-30
[Table view] |
Text
I NRG FORM 366
LICENSEE EVENT REPORT (LER)
. U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION FORWARD AND RECORDS MANAGEMENT BRANCH (MNBB 7714). U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
Salem Generating Station DOCKET NUMBER (2)
II PAGE (3) 05000 311 1 OFlO TITLE (4) Engineered Safety Features Actuation (Reactor Trip) During Unit 2 Controlled Shutdown Per Technical Soecification 3.0.3.
EVENT DATE (5) LEA NUMBER (6 REPORT NUMBER (7) OTHER FACILITIES INVOLVED (8)
FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000 FACILITY NAME 6 07 95 95 -- 004 -- 00 07 07 95 DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more (11)
MODE (9) 1 20.402(b) 20.405(c) x 50. 73 (a) (2) (iv) 73.7.1 (b)
POWER 20.405 (a) (1) (i) 50.36(c) (1) 50. 73 (a) (2) (v) 73.71(c)
LEVEL (10) 100 20.405(a) (1 )(ii) 50.36(c)(2) 50.73(a) (2) (vii) OTHER
- 20.405(a) (1) (iii) x 50. 73 (a) (2) (i) 50.73(a) (2) (viii) (A) (Specify in Abstract below and in Text, NRC 20.405 (a)(1 )(iv) 50.73(a) (2) (ii) 50.73(a) (2) (viii) (B) i
- ., !!llil'll!;il i\'11! Iii, 20.405(a) (1) (v) 50.73(a) (2) (iii)
LICENSEE CONTACT FOR THIS LEA 12) 50.73(a) (2)(x)
Form 366A)
NAME Warren Hall, LER Coordinator T6LEPHONE NUMBER (Include Area Code) 09 339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS x BP FCV V085 y SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH I DAY~
xj YES (If yes, complete EXPECTED SUBMISSION DATE)
NO SUBMISSION DATE (15) 09 \ rs 195 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On 6/7/95, at 1828 hours0.0212 days <br />0.508 hours <br />0.00302 weeks <br />6.95554e-4 months <br />, a Unit 2 controlled shutdown from Mode 1 was initiated in accordance with the requirements of Technical Specification (TS) 3.0.3 due to the inoperability of both Residual Heat Removal (RHR) trains. The inoperability was caused by the failure of the 21RHR pump recirculation valve ( 21RH2 9) to automatically open when the 21RHR pump was started and the inability to ensure operability of the 22RHR pump recirculation valve (22RH29). With both RHR trains inoperable the Unit was operating in a condition not covered by the TSs and the requirements of TS 3.0.3 were initiated. The Unit was reducing load with the group buses already transferred to the 21 and 22.Station Power Transformers. When the Unit load was below 50 MW (14% reactor power), the Nuclear Control Operator manually tripped the turbine. 500KV Bus Section breakers, BS 1-9 and BS 9-10, both opened; however, BS 1-9 breaker failure protection circuitry actuated unexpectedly causing loss of Bus Section 1 which represents the loss of one source of off-site power to both Salem Units. This caused the loss of 4KV group buses 2F and 2G which caused the trip of 23 and 24 reactor coolant pumps, followed by an automatic reactor trip on low flow in two of four reactor coolant loops (> 10 % reactor power) at 2301 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.755305e-4 months <br /> on 6/7/95.
The investigation of the RHR System inoperability event is continuing and final results, including apparent cause and corrective actions, will be provided in a supplement to this LER, anticipated by 9/15/95.
NRC FORM 366 (5*92) 9507110212 950707 PDR ADOCK 05000311 - ...... r.
REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL 2 DOCKET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6 TOTAL 5 EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL - DOCKET NUMBER 3 IN ADDITION TO 05000 9 1 OPERATING MODE 10 3 POWER LEVEL 1
11 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1
14 SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK
Unit # 2
- Docket Number 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Page 2 of 10 Plant and System Identification:
Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes appear in the text as {xx}
Identification of Occurrence:
Engineered Safety Features (ESF) Actuation (Reactor Trip)
During Unit 2 Controlled Shutdown Per Technical Specification 3.0.3, Due To Inoperability Of Both RHR Trains.
Event Date: June 7, 1995 Report Date: July 7, 1995 Conditions Prior to Occurrence:
Mode: 1 Reactor Power: 100% Unit Load: 1120 MWe Description/Analysis of Occurrence:
On June 7, 1995, Operations requested that both RHR pump recirculation valves (21RH29 & 22RH29) be tested to assure that they would open when called upon when the pump started.
At 0150 the 22RHR pump was started and 22RH29 opened automatically as required. The pump was then stopped and 22RH29 closed automatically as required. AT 0155 the 21RHR pump was started but 21RH29 did not open automatically as required. The operator took manual control of 21RH29 and opened the valve. The 21RHR pump was then stopped and 21RH29 was placed in automatic and the valve closed. The operator then entered Technical Specification 3.5.2c, Action Statement a, which requires that the inoperable ECCS subsystem be returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the unit must be shutdown. Continuing evaluation of both RHR pump recirculation valves confirmed that 21RH29 was inoperable and that 22RH29 was operable but may be in a degraded condition. A follow up operability assessment of 22RH29 was requested to be completed in a timely manner.
Unit# 2
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number Page 3 of 10 Description/Analysis of Occurrence (cont'd):
Trouble shooting of 21RH29 was initiated and as a result of failure analysis, three relays in the control circuitry were replaced. Because the follow up assessment for operability of the 22RH29 valve could not be completed in a timely manner commensurate with the safety significance of the RHR System {BP}, 22RH29 was declared inoperable. At 1827 hours0.0211 days <br />0.508 hours <br />0.00302 weeks <br />6.951735e-4 months <br /> on June 7, 1995, Unit operators entered Technical Specification 3.0.3 and commenced a controlled shutdown of the Unit from Mode 1 due to both trains of the RHR System being inoperable.
At 1907 on June 7, 1995, Public Service Electric & Gas made a one-hour notification to the NRC in accordance with 10CFR50.72(b) (1) (i) (A).
An orderly shutdown was in progress and the Unit was reducing load at a rate of 30% per hour. As the load reached 20% power the group buses were transferred from the Auxiliary Power Transformer to the 21 and 22 Station Power Transformers (SPTs) . As the load was reduced below 50MW (14% reactor power) the operator manually tripped the turbine in accordance with normal operating procedures.
This action in conjunction with the actuation of the back-up reverse power relay caused the Generator Overload and Out of Step HEA Multi-Trip (MT) to operate. The Generator Overload MT subsequently initiated several signals to circuitry which included BS 1-9 (32X) breaker trip coil, BS 9-10 (30X) breaker trip coil, BS 1-9 breaker failure relay, BS 9-10 breaker failure relay, Exciter Field Breaker, and Overhead Alarms. As expected, this action resulted in the opening of both BS 1-9 and BS 1~10 breakers. However, BS 1-9 breaker failure protection circuitry actuated unexpectedly causing loss of Bus Section 1 which represents the loss of one source of off-site power to both Salem units. The operation of BS 1-9 breaker failure protection circuitry resulted in tripping 500KV BS 1-5, BS 1-8, 13KV BS 3-4, BS 4-5, BS C-D, BS D-E, and Gas Turbine breakers.
This caused a loss of power to 2SPT, 4SPT, 12SPT, 14SPT, 22SPT, 23SPT, and 4KV group buses lF, lG, 2F, and 2G. The loss of 4KV group buses 2F and 2G caused 23 and 24 reactor coolant pumps to trip and consequently resulted in an automatic reactor trip on low flow conditions in two of four reactor coolant loops.
The BS 1-9 breaker failure relay actuated after the breaker trip signal was received. The Generator Overload MT
Unit# 2 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number 50-311 LER Number 95-004-00
- Page 4 of 10 Description/Analysis of Occurrence (cont'd):
initiated the breaker trip signals and breaker failure timing to both BS 1-9 and BS 9-10 simultaneously, but the breaker failure time delay relay unexpectedly actuated for BS 1-9. There was no problem in opening the BS 1-9 breaker within the required time, however, the BS 1-9 breaker failure relay time delay requirement did not reset during the event thereby causing the breaker failure circuitry to actuate without the proper time delay. The purpose of the breaker failure relay time delay is to allow adequate time for the breaker to open prior to actuating the breaker failure relay. The BS 1-9 breaker and the breaker failure relay actuation occurred at approximately the same time.
This points to a deficiency or malfunction in the breaker failure relay (SBF-1) circuitry.
Tests performed under the troubleshooting plan extensively checked the breaker main contacts, auxiliary contacts, associated control circuitry, SBF-1 circuitry, and the time delay circuitry. Results of the.preliminary on-site testing found no anomaly or problems with any of the circuitry.
However, laboratory analysis of the breaker failure circuitry is being conducted and the results, if available, will be provided in the Supplement to this LER.
At 0152 on June 8, 1995, Public Service Electric & Gas made a four hour notification to the NRC in accordance with 10 CFR5 0 . 7 2 (b) ( 2 ) ( ii ) .
Apparent Cause of Occurrence:
The apparent cause of the occurrence related to the RHR system inoperability is still under investigation and will be provided in a supplement to this LER, anticipated by 9/15/95.
The cause of this occurrence related to the BS 1-9 breaker failure and the subsequent reactor trip is attributed to the malfunction of the SBF-1 breaker failure relay.
The manufacturer of the SBF-1 relay (Westinghouse/ABB) had notified users of SBF-1 relays manufactured prior to 1992 that they are susceptible to misoperation under transient voltage conditions. The malfunction has been attributed to surge voltage conditions on the 125V DC system. The energization and de-energization of the relays in the
. Unit# 2
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number Page 5 of 10 Apparent Cause of Occurrence (cont'd) control circuitry creates a very rapid switching transient of greater than 2.5KV on the 125V DC system. The opening of the main breaker contacts also impresses additional noise signal transients on the 125V DC system. These transients have the potential to momentarily short circuit the time delay device in the SBF-1 relay circuitry resulting in the premature operation of the relay. This misoperation of the relay was preliminarily determined to be the reason for the breaker failure and the subsequent reactor trip.
In August 1993, a notice from the manufacturer (Westinghouse/ABB) of the possible SBF-1 relay malfunction for relays manufactured prior to 1992 was received by the Nuclear Business Unit. This notification provided information on the use of certain types of cables with the relay and also stated that a surge suppresser kit had been developed to upgrade the surge withstand capability of the relay to prevent future relay malfunction. Relays manufactured after 1992 have the surge protection feature upgrade built in.
The corrective action recommended in the vendor notice was not acted upon until May 16,1995, when a Design Change Request (DCR) was issued by Nuclear Engineering to replace the current relays with the upgraded model. This modification had not taken place at the time of the breaker failure.
Prior Similar Occurrence:
A review of the Salem records determined that there has been no similar occurrences at the Salem Station related to either the RHR inoperability or the breaker failure.
Safety Significance:
The RHR System inoperability is reportable pursuant to 10CFR50. 73 (a) (2) (i) (A) and the reactor trip is reportable pursuant to 10CFR50. 73 (a) (2) (iv).
RHR Pump Recirculation Valves (21RH29 & 22RH29)
Testing of the 21RH29 recirculation valve determined that the valve would fail to open after placing the 21RHR pump in service. Testing of the 22RH29 recirculation valve determined that the valve would open upon receiving a start signal thereby assuring that the 22RHR pump would not be
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number Page 6 of I 0 Safety Significance (cont'd) placed in a shut off head condition. Testing did not confirm that the 22RH29 valve would close when RHR flow to the core was required for various sized LOCAs. If this valve were to remain open, flow delivered to the core would be reduced under some conditions.
No safety injection (SI) signal occurred during this event therefore no automatic signal to start the RHR pumps was generated. Consequently, no operation of the RHR pumps in either normal or shut off head conditions was required. The fact that one recirculation valve was inoperable and the operability of the other was of concern did not contribute in any way to the Unit trip.
If a SI signal had been generated and the valves responded consistently with the test results, the 22RHR pump would have had recirculation. Assuming no other single failure, 22RHR pump would have been available to perform its intended safety function, while the 21RHR pump would have been in a shut off head condition. After operating for an extended period of time in this configuration, the 21RHR pump could eventually fail sometime after 45 minutes, while the 22RHR pump would be available to deliver flow to the core. In the Salem safety analysis, the single failure assumed for the minimum safeguards analysis is the loss of one RHR pump.
This is consistent with the conditions that existed at the time the testing was conducted with the exception that it could not be assured that the 22RH29 valve would close when flow to the core was required. If the 22RH29 valve did not close, the flow to the core would be reduced by the amount of the recirculation flow. However, ample margin is provided between various conservatisms and in newer, generically approved analyses. Credit is available for Large Break LOCAs since they assume the broken line spills to 10 psig rather than 0 psig (this artificially increases the peak cladding temperature (PCT)). Additional margin would also exist by crediting the second IHSI pump which would not be compromised by failure of the RH29 valve to close. Further margin is also available in the Large Break LOCA PCT. The PCT difference between maximum and minimum safeguards flow is 102°F. The change in flow caused by an open RH29 valve would be substantially less than the difference between maximum and minimum safeguards. Hence, the increase to PCT would be expected to be less than 102°F.
This would be acceptable since there is 186°F of margin. On
Unit# 2
- Docket Number 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Page 7 of 10 Safety Significance (cont'd) the basis of these considerations, it is judged that the penalty on an RHR pump associated with an open RH29 valve would not have caused the Unit to exceed 10CFR50.46 criteria for Large Break LOCA. Small Break LOCA and LOCA hydraulic forces are not affected by RHR flows because they are not credited in the analyses. Furthermore, long term core cooling and hot leg switchover are not impacted by this issue because the recirculation valve is manually closed by the operator when recirculation begins.
Short term LOCA subcompartment analysis and pump runout concerns have also been considered. It was concluded that there would be no adverse effect on these issues due to an open RH29 valve. Consequently, if the recirculation valves behave consistently with the test results and no other failures occurred to compromise the 22RHR pump, it can be concluded that the inoperable status of the recirculation valves did not pose any undue risk to public health and safety.
With the existing condition on 21RH29, a single failure could cause 22RH29 to fail closed and the both RHR pumps would be in a shut off head condition. The current Large Break LOCA analysis uses RHR pump flows that assume the recirculation valves are closed. The reactor rapidly depressurizes to approximately 30 psia in about 30 seconds.
With a loss of off site power and maximum safety injection delay time, the RHR pumps will still deliver flow following a large break LOCA since the valves would already be closed.
No operator action would be required to open the recirculation valves following a large break LOCA since system pressure would be low and pump flow would be adequate to keep the valves closed. Therefore, the failure of both recirculation valves to open would have no impact on the safety analysis or the ability of the RHR system to perform its safety function during a large break LOCA.
For the limiting Small Break LOCA PCT analysis, operation of the RHR pumps is not credited since system pressure does not fall below the cut in pressure during the transient.
Therefore, failure of the recirculation valves to open does not impact the analysis. If the operator were to secure the pumps or open the recirculation valves there would also be no impact on the analysis. All other Chapter 15 analyses do not recognize start up or operation of the RHR pumps at RCS pressures above the shut off head of the pumps. Therefore,
Unit # 2
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number Page 8of10 Safety Significance *(cont'd) plant safety would not have been compromised and the results and conclusions of the safety analysis remain valid for the period in which the recirculation valves would not have opened.
Operation of the pumps in a shut off head condition for an extended period of time has also been reviewed. The Salem Emergency Operating Procedures (EOPs) direct the operators to secure the RHR pumps. This activity is normally performed in approximately 30 minutes. Westinghouse predicts that the Salem RHR pumps could operate at shut off head conditions for 45 minutes and still perform their required safety function after restoration of normal flows.
PSE&G has reviewed the EOPs and has confirmed that 45 minutes provides the operators with adequate time to perform these actions. Therefore, the RHR pumps would have been secured before damage could occur and would remain capable of performing their intended safety function when needed.
Effects of the Transient on the Primary Plant The transient that resulted following the reactor trip is bounded by the loss of offsite power and loss of flow analyses in Chapter 15 of the UFSAR that assumes the coast down of all four loops. Clearly the loss of flow in four loops is more severe and therefore bounds the loss of flow in two loops. The analysis is unaffected by the inoperability of 21RH29 and 22RH29. With the Unit at 14%
power at the time of the reactor trip, additional margin to Departure From Nucleate Boiling (DNB) and reactor vessel pressure limits is provided. The loss of 23 and 24 reactor coolant pumps coupled with the isolation of letdown, reduced normal pressurizer spray flow. This has no impact on the Chapter 15 analysis since spray is only assumed in the analysis if it is detrimental to the accident. In this transient, pressurizer spray was not detrimental since it would have mitigated the pressure increase. Further margin for the analysis was provided when auxiliary pressurizer spray was initiated to stop increasing pressurizer pressure.
Reactor coolant system pressure was maintained below 2350 psia throughout the transient. Based on the above discussion, this transient is bounded by the Chapter 15 accident analysis and had no impact on the reactor core or the reactor coolant system.
Unit# 2
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINO ATION Salem Generating Station Docket Number Page 9 of 10 Safety Significance (cont'd) 500KV Bus Section 1-9 Breaker Failure/Reactor Trip As a result of the failure of the breaker failure protection relay of the BS 1-9 breaker Bus Section 1 of the 500KV ring bus was de-energized. This resulted in the loss of power to the Number 2 and Number 4 Station Power Transformers (SPT) .
The Number 2 SPT normally feeds the 4KV 2F and 2G group buses through the Number 22 SPT and the Number 4 SPT normally feeds the 4KV 2B and 2C vital buses and one half of the circulating water bus through the Number 23 SPT.
When power was lost to the Number 2 SPT, all equipment fed from the 2F and 2G group buses was de-energized. All of this equipment is classified as non-safety related and safe shut down of the Unit without this equipment is within the design basis analysis of the plant.
When power was lost to the Number 4 SPT, the vital buses transferred to their alternate power source, as designed.
This transfer took place as required and all effected equipment functioned as designed. If the transfer to the alternate source failed, the diesel generators were available and would have started and sequenced on the appropriate shutdown loads. This would have allowed the operators to continue the safe and orderly shut down of the Unit. Therefore, the loss of the Number 4 SPT is within the design basis analysis of the plant.
The partial loss of off site power/reactor trip is within the design basis analysis of the plant and therefore did not pose any additional risk to public health and safety.
Corrective Action:
Corrective actions related to the RHR system inoperability will be established as a result of completing the cause investigation and will be provided in a Supplement to this LER, anticipated by 9/15/95.
All SBF-1 breaker failure protection relays on the 500KV breakers will be replaced with upgraded relays that have higher surge withstand capabilities. The SBF-1 relays on the generator output breakers (BS 1-9, BS 9-10, BS 1-5, and BS 5-6) will be replaced first followed by other 500KV breakers.
Unit# 2
- 50-311 LER Number 95-004-00 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number Page 10 of 1O Corrective Action (cont'd):
All SBF-1 breaker failure relays on the new 13KV BS A-B, BS B-C, BS C-D, and BS D-E will be replaced with up~raded relays.
Similar circuitry for other breaker failure relays at both Salem and Hope Creek will be reviewed to determine transient susceptibility and replaced with upgraded relays as necessary.
The process used to evaluate and implement vendor recommendations in a prompt and effective manner will be reviewed and changes made where necessary.
The status of these corrective actions will be provided in the supplement to this LER.
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J. C. Summers General Manager -
Salem Operations MJPJ:vs REF: SORC Mtg.95-072