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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:RO)
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17284A9001999-10-31031 October 1999 Rev 0 to COLR 99-15, WNP-2 Cycle 15,COLR GO2-99-177, LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With1999-10-0101 October 1999 LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With ML17284A8941999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for WNP-2.With 991012 Ltr ML17284A8801999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for WNP-2.With 990910 Ltr ML17284A8691999-07-31031 July 1999 Monthly Operating Rept for July 1999 for WNP-2.With 990813 Ltr ML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B7271999-06-30030 June 1999 Monthly Operating Rept for June 1999 for WNP-2.With 990707 Ltr ML17292B6961999-05-31031 May 1999 Monthly Operating Repts for May 1999 for WNP-2.With 990608 Ltr ML17292B6641999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for WNP-2.With 990507 Ltr ML17292B6391999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for WNP-2.With 990413 Ltr ML17292B5871999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for WNP-2.With 990311 Ltr ML17292B5571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for WNP-2.With 990210 Ltr ML17292B5621999-01-31031 January 1999 Rev 1 to COLR 98-14, WNP-2 Cycle 14 Colr. ML17292B5341999-01-15015 January 1999 Part 21 Rept Re Incorrect Modeling of BWR Lower Plenum Vol in Bison.Defect Applies Only to Reload Fuel Assemblies Currently in Operation at WNP-2.BISON Code Model for WNP-2 Has Been Revised to Correct Error ML17292B5331999-01-15015 January 1999 Part 21 Rept Re XL-S96 CPR Correlation for SVEA-96 Fuel. Defect Applies Only to WNP-2,during Cycles 12,13 & 14 Operation.Evaluations of Defect Performed by ABB-CE ML17292B4791998-12-31031 December 1998 Washington Public Power Supply Sys 1998 Annual Rept. with 981215 Ltr ML17292B5351998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for WNP-2.With 990112 Ltr ML17292B5741998-12-31031 December 1998 WNP-2 1998 Annual Operating Rept. with 990225 Ltr ML17284A8231998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for WNP-2.With 981207 Ltr ML17284A8081998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for WNP-2.With 981110 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7831998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for WNP-2.With 981007 Ltr ML17284A7491998-09-10010 September 1998 WNP-2 Inservice Insp Summary Rept for Refueling Outage RF13 Spring,1998. ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7681998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for WNP-2.With 980915 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7261998-07-31031 July 1998 Monthly Operating Rept for July 1998 for WNP-2.W/980810 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6751998-06-30030 June 1998 Ro:On 980617,flooding of RB Northeast Stairwell with Consequential Flooding of Two ECCS Pump Rooms.Caused by Inadequate Fire Protection Sys Design.Pumped Out Water from Affected Areas to Point Below Berm Areas of Pump Rooms ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17284A6491998-05-31031 May 1998 Rev 0 to COLR 98-14, WNP-2,Cycle 14 Colr. ML17292B4031998-05-31031 May 1998 Monthly Operating Rept for May 1998 for WNP-2.W/980608 Ltr ML17292B3921998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for WNP-2.W/980513 Ltr ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3371998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for WNP-2.W/980409 Ltr ML17292B2641998-03-0404 March 1998 Performance Self Assessment,WNP-2. ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B2911998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for WNP-2.W/980313 Ltr ML17284A7971998-02-17017 February 1998 Rev 28 to Operational QA Program Description, WPPSS-QA-004.With Proposed Rev 29 ML17292B3591998-02-12012 February 1998 WNP-2 Cycle 14 Reload Design Rept. 1999-09-30
[Table view] |
Text
AC CEMRATED t
DISIKBUTION DEMONSTRATION
! SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR-8907240014 DOC.DATE: 89/07/17 NOTARIZED: NO DOCKET N FACIL:50-397 'WPPSS Nuclear Project, Unit 2,'ashington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION ARBUCKLE,J.D. Washington Public Power Supply Sy'tem POWERS,C.M. Niagara Mohawk Power Corp.
RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 89-025-00:on 890617,ESF actuations during excess flow check valve testing due to procedural inadequacies.
W/8 ltr.
DISTRIBUTION CODE: IE22T COPIES, RECEIVED: LTR I,EN CL,-~. SIZE-, -...
TIT'LE:;..50,.73/50.9 Licensee'.Event;'Report'(LER)', Incident:Rpt, etc...,','OTES:
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KKP US 'IO EKDUCH HASTE.'URAT 'IHE DOCUMENI'ONI.'FOL DESK, IZSTB XQR DOCUMEHXS KU DGNPT NEZDt FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 43 ENCL 42
WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352 Docket No. 50-397 July 17, 1989 Document Control Desk U.S. Nuclear Regulatory Comnission Washington, D.C. 20555
Subject:
NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.89-025
Dear Sir:
Transmitted herewith is Licensee Event Report No.89-025 for the WNP-2 Plant.
This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportabi.lity, corrective action taken, and action taken to preclude recurrence.
Very truly yours, C.M. Powers (M/D 927M)
WNP -2 Plant Manager CMP:lg
Enclosure:
Licensee Event Report No.89-025 cc: Mr. John B. Martin, NRC - Region V Mr. C.J. Bosted, NRC Site (M/D 901A)
INPO Records Center - Atlanta, GA Ms. Dottie Sherman, ANI Mr. D.L. Williams, BPA (M/D 399)
~pe ~gtl1 is~
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'='9072400i4 8907i7 F'DR ADOCK 0=000 97 PDC
NRC Form 366 V.S. NUCLEAR REOULATORY COMMISSION (0.83)
APPROVED 0M B NO. 3(604)104 EXPIRES; 8/31/86 LICENSEE EVENT REPORT (LER)
FACILITY NAME 0) DOCKET NUMBER (2) PACE 3 0 5 0 0 0 1 OF
"""" Engineered Safety Feature (ESF) Actuations During Excess Flow Check Valve Testing Due to Procedural Inade uacies EVENT DATE (SI LE R NUMBER (6) REPORT DATE It) OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQVENTIAL REVISION MONTH DAY YEAR FACILITYNAMES DOCKET NUMBERISI NUMBER NUMBER 0 5 0 0 0 06 17 8 98 0 2 5 0 0 71 7 8 9 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T0 THE REOUIREMENTS OF 10 CFR (): (Check one or more ol the Iolrowinpl ill)
OPERATINO MODE ISI 60.73(e) (2) (ivl 73.71(5) 4 20.402(b) 20,405(cl POWER 20.406( ~ l(1) (i) 50.36(e)(ll 50.7 3(e) (2) (vl 73.71(c)
LEVEL 20.406( ~ l(lI (ii) 50.36(cl(2) 50,73(el(2)(viil OTHER (Specify in Ahrtrect below end In Text, IIRC Form 20.406(el(l l(iii1 50.73( ~ l(2) (I) 50.73(e l(2 l(viii)(Al 366AI O 'rirrrP 3 20A05( ~ l(1) (iv) 50,73(el(2)(ill 60.73 (e) (2 I (viii)(8)
- y. e "vM'cl<<yyrriyol oo. c6e.<< 20.40S ( ~ I (I )(v) 50,73(el(21(iiil 60,73 (~ ) (2)(x I LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER AREA CODE COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
MANUFAC EPORTABLE MANUFAC EPORTABLE CAUSE SYSTEM COMPONENT TURER TO NPRDS
~%@5~rVPP44 CAUSE SYSTEM COMPONENT TURER TO Ni'RDS v<<TPj~~(i.:
<<rogn4g4~a; SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SUB M I SS IO N DATE (151 YES Ilf yeL comPlete EXPECTED SUBhllSSIOIY DATEI NO ABSTRACT I(.imlr to l400 rpecer, l.e., epproximetNy I<<'lteen rlnole.rpece typewr<<tten linNI (16)
During the performance of the Plant procedure for Excess Flow Check Valve Testing, three separate but related events occurred which caused Engineered Safety Feature (ESF) isolations and actuations. These events are combined into one LER in accordance with the guidance procided in NUREG 1022 (Supplement No. 1) as "similar events that are part of the same activity or test program." At the time of the events the Plant was in a shutdown condition for the annual maintenance and refueling outage. The events are described as follows:
On June 17, 1989 at 1222 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.64971e-4 months <br />, the inboard Residual Heat Removal (RHR) Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated during excess flow check valve testing. The root cause of this event is procedural inadequacy in that the procedure did not caution against increasing pressure to a value that would cause the RPV high pressure isolation actuation. Plant Operators were in the process of increasing reactor pressure by raising vessel level with water from the Control Rod Drive (CRD) System. When RPV pressure reach 122 psig, a Reactor Recirculation (RRC)
System pressure switch actuated and, by design,'HR-V-9 closed which isolated RHR Shutdown Cooling.
- 2) On June 18, '1989 at 1045 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.976225e-4 months <br />, during the performance of excess flow check valve testing, a Reactor Protection System ( RPS) UAR half-scram and a High Pressure Core Spray (HPCS) Diesel Generator (DG-3) start occurred due to a pressure transient which affected several reactor vessel level instruments. The pressure transient was causea.when a Plant Instrument and Control ( ISC) Technician inadvertently opened the wrong valve during performance of the procedure. The root cause of this event is procedural inadequacy in that the procedure did not list the corresponding drain
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NRC Form 365 (8 83I
NRC Form 366A U.S. NUCLEAR REOULATORY COMMISSION (9%31 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO 3150M(04 EXPIRES: 8/31/88 FACILITY NAME (II DOCKET NUMBER (2) LER NUMBER (6) PACE (3) 5,II EEOVENTIAL NVMEER
.c~.(F REVISION NVMEER Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 8 9 0 2 5 0 0 0 2 OF TEXT illmore FPeoe /8 ver/IvtvevL rrre edd/0'ovM///RC Form J6/tv(3l I IT(
Abstract (cont'd) valves required to test specific excess flow check valves, and instead relied upon the ability of the field technicians to trace the instrument tubing from the instrumentation to the drain valves. As a result, the I&C Technicians opened the drain valve associated with the opposite side (high) of instrument MS-LT-27 instead of the low side. This caused a lowering in the pressure to instruments connected to two separate containment penetrations and resulted in the isolations and actuations.
- 3) On June 18, 1989 at 1814 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90227e-4 months <br />, the inboard Residual Heat Removal (RHR)
Shutdown Cooling Supply Valve ( RHR-V-9) again isolated on a Reactor Pressure Vessel (RPV) high pressure signal during excess flow check valve testing. The root cause of this event is procedural inadequacy in that the procedure did not explicitly state the impact of isolating instrument MS-PT-51A, nor did provide a clear caution statement directing that the Control Room Operators be it alerted that MS-PT-51A was isolated and pressure indication impaired. In accordance with the procedure, Plant I&C Technicians had valved out MS-PT-51A in order to test Excess Flow Check Valve PI-EFC-X114. When the instrument was valved out of service, the pressure at the transmitter was trapped and the indication in the Control Room stayed at that pressure.
Because the indicated-pressure remained constant, Plant Operators had no reason to vent the reactor vessel pressure during the test. As a result, eventually reactor pressure reached a point high enough to trip a Reactor Recirculation (RRC) pressure switch which, by design, closed RHR-V-9 and isolated RHR Shutdown Cooling.
Immediate corrective action consisted of resetting the isolations and actuations and restoring systems to service, and (after the third event) suspending excess flow check valve testing until a management team could investigate and develop a plan to prevent further problems. Further corrective actions consist of revising the proceaure to identify the consequences of each instrument removed from service, and performing an evaluation of RHR Shutdown Cooling isolations which have occurred.
These events posed no threat to the health and safety of either the public or Plant personnel.
Plant Conditions
'a) Power Level - OX b) Plant Mode - 4 (Cold Shutdown)
Event Descri tion During the performance of Plant Procedure (PPM) 7.4.6.3.4.1, "Excess Flow Check Valve .Testing," three separate but related events occurred which caused Engineered Safety Feature (ESF) isolations and actuations. These events are combined into one LER in accordance with the guidance provided in NUREG 1022 (Supplement No. 1) as "similar events that are part of the same activity or test program." At the time of the events the Plant was in a shutdown condition for the annual maintenance and refueling outage. The events are described as follows:
NRC FORM 386A ~ 0 ~ 8 ~ CPor 1988-820 889 00020 (9 831
NRC Form 388A U.S. NUCLEAR REGULATORY COMMISSION (943)
LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVED OMS NO. 3)80-0)04 EXPIRES: 8/31/(8 FACILITY NAME (1) OOCKET NUMBER (3) LER NUMBER (8) PAGE (3)
- 4 p4 SEOUENTIAL NUMBER IALS REVISION NUMBER Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 8 9 0 2 5 0 0 0 3 TEXT /I/more specs Is reqrreed, Iree eddrdooelHRC %%drrrr 388r('s) 07)
On June 17, 1989 at 1222 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.64971e-4 months <br />, the inboard Residual Heat Removal (RHR)
Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated on a Reactor Pressure Vessel (RPV) high pressure signal during excess flow check valve testing. The valve is a Nuclear Steam 'upply Shutoff System (Containment Isolation) valve. Closure of RHR-V-9 isolated RHR Shutdown Cooling Loop A which was in service at the time of the event.
As part of the testing, the procedure required that reactor vessel pressure be maintained at 100 psig minimum pressure. To meet this requirement, were in the process of increasing reactor pressure by raising vessel Plant'perators level with water from the Control Rod Drive (CRD) System.
When reactor pressure reached 122 psig as indicated on Main Steam Pressure Inaicator MS-PI-9 in the Control Room, a Reactor Recirculation (RRC ) System pressure switch (RRC-PS-18B) actuated and, by design, RHR-V-9 closed which isolated RHR Shutdown Cooling.
- 2) On June 18, 1989 at 1045 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.976225e-4 months <br />, during the performance .of excess flow check valve testing, a Reactor Protection System (RPS) NAN half-scram and a High Pressure Core 'pray (HPCS) Diesel Generator (DG-3) start occurred due to a
'pressure transient which affected several reactor vessel level instruments.
The pressure transient was caused when a Plant Instrument and Control (I8C)
Technician inadvertently opened the wrong valve during performance of the procedure.
Excess flow check valve testing is accomplished by isolating all instruments utilizing the containment penetration associated with the excess flow check valve being tested. When the instrumentation is isolated, the drain valve for the penetration is opened and both drain flow and Control Room excess check flow valve position indications are observed. The Test Director in the Control Room verifies indication of valve closure and the I&C technician verifies that the flow of water to the drain has stopped. A closed indication and significant flow decrease indicate a successful test.
At the time of the event, Excess Flow Check Valve PI-EFC-X72A was being tested. The I&C Technicians had valved out the instruments associated with containment penetration X72A. These instruments were MS-LT-27 and RFW-DPT-17.
Instrument MS-LT-27 is a differential pressure instrument having two sensing lines (high and low), and two drain lines associated with two different
,containment penetrations. The procedure required that the I&C technician open the drain valve associated with the side of MS-LT-27 connected to penetration X72A. However, the procedure did not list the valves necessary to test specific excess flow check valves, nor did low side.
it differentiate between'igh and NRC FORM SSBA AVAST CPOr 1988 530-589r000)0 (9431
NRC Form 36BA U.S. NUCLEAR REOULATORY COMMISSION (943)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. 3150&)04 EXPIRES: 8/31/88 FACILITYNAME () I COCKET NUMBER (3) LER NUMBER (6) PACE (3)
"SEOUENTIAL REVISION NUMBER NUMBER Washin ton Nuclear Plant - Unit 2 o 3 9 7 8 9 0 2 5 000 4o'EXT
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As a result, the I8C Technicians opened the drain valve associated with the opposite side (high) of MS-LT-27 instead of the low side. The drain valve that was opened is associated with penetration X110 and is connected to other level instrumentation. When the valve was opened, pressure to the, instruments connected to that penetration.
it caused a lowering i'n the Penetration X110 is also the variable leg of instrument MS-LIS-24C which tripped at the Reactor
,Level-3 setpoint and caused the RPS, Division 1, half-trip.
In addition, two other instruments (MS-LIS-31B and MS-LIS-31D) connected to penetration X110 share another penetration (X112: reference leg) with NS-LIS-24C and MS-LIS-100B 'hich were affected by the event. The pressure disturbance on penetration X110 was transmitted through the bellows of MS-LIS-24C and MS-LIS-100B to penetration X112. As a result, all of the instruments connected to penetration X112 were affected by this event.
Instruments MS-LIS-31B and NS-LIS-31D were affected enough to trip at their Reactor Level-2 (-50U) setpoint which caused the HPCS Diesel to start (and attain rated speed) and the HPCS injection valve (HPCS-V-4) to open. As a precautionary measure, the HPCS pump (HPCS-P-1) motor had previously been placed out of service and, as a result, no HPCS injection occurred.
- 3) On June 18, 1989 at 1814 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90227e-4 months <br />, the inboard Residual Heat Removal (RHR)
Shutdown Cooling 'Supply Valve (RHR-V-9) again isolated on a Reactor Pressure Vessel (RPV) high pressure signal during excess flow check valve testing. As during the previous event (June 17, 1989), closure of RHR-V-9 isolated RHR Shutdown Cooling. During this event, RHR Shutdown Cooling, Loop B, was in service.
In accordance with the procedure, Plant I8C Technicians valved out the instruments associated with containment penetration in order to test Excess Flow Check Valve PI-EFC-X114. One of the instruments connected to penetration X114 is MS-PT-51A which provides the reactor steam dome pressure to instruments NS-LR/PR-623A and NS-PI-9 in the main Control Room. When NS-PT-51A was valved out of service, the pressure at the transmitter was trapped and. the indication in the Control Room stayed at that pressure until the transmitter was valved back into service after the event.
Control Room Operators were controlling reactor pressure by blowing down excess water provided by the Control Rod Drive (CRD) System whenever the pressure in the vessel approached 110 psig. Because the Control Room - indicated pressure remained constant, Plant Operators had no reason to vent the reactor vessel pressure. As a result, eventual'ly the reactor pressure reached a point high enough to trip RRC-PS-18 which closed RHR-V-9 and isolated RHR Shutdown Cooling.
Immediate Corr ective Action June 17, 1989 (1222 Hours)
~ Control Room Operators 'lowered RPV pressure, reset the RHR-V-9 isolation and opened the valve. At 1301 hours, RHR Shutdown Cooling was restored.
NRC FOAM Sdda 'II~ 6 ~ CPOr )966 $ 30 $ 69 000)0 (9 83)
NRC Form 388A U.S. NUCLFAR REOULATORY COMMISSION (84)3)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3160&10&
EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER LB) PACE (3)
YEAR S&OV&NZIAL R&VISION NVM &R NVMS&R Washin ton Nuclear Plant - 0 5 0 0 0 OF Ilfmore sosse Un'EXT Is requkerL use aAIIEoosl HRC Form 3884'sl (12)
- 2) June 18, 1989 (1045 Hours)
~ Control Room Operators reset the RPS RAR half-scram and HPCS actuation signals. By 1108 hours0.0128 days <br />0.308 hours <br />0.00183 weeks <br />4.21594e-4 months <br />, the HPCS Diesel Generator was secured and restored to standby status.
- 3) June 18, 1989 (1814 Hours)
~ .
Control Room Operators lowered RPV pressure, reset the RHR-V-9 isolation and opened the valve. At 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br />, RHR Shutdown Cooling was restored.
~ As directed by the Assistant Plant Manager, excess flow check valve testing was suspended and a management team assigned to investigate and develop a plan to prevent further problems.
Further Evaluation and Corrective Action A. Further Evaluation
- l. All three events are reportable under 10CFR50.73(a)(2)(iv) as "an event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF).U
- 2. There were no structures, systems or components that were inoperable at the start of these events that contributed to the events.
- 3. The root causes of these events are as follows:
a) June 17, 1989 Event
~ Procedural Inadequacy: The procedure governing the excess flow check valve surveillance testing did not caution against increasing pressure to a value that would cause the RPV high pressure isolation actuation. The procedure specified a minimum pressure (100 psig) but did not identify a range of allowable pressures. In addition, Control Room Operators were aware of the Technical Specification Trip Setpoint of 125 psig but did not realize that, due to instrument inaccuracies and head corrections, the trip could occur at indicated pressures below 125 psig. A review of, the administratively-acceptable trip value for RRC-PS-18B indicated an acceptable trip value as low as 114 psig. Had Plant Operators been provided with either a pressure band within which to perform the test, or a caution against increasing pressure above a specified value, this event would not have occurred.
NRC FORM SSISA *0 S Ceor 1988 520 589 00010 IB 83)
NRC Foem 388A U.S. NUCLEAR REOULATORY COMMISSION (943)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3(SOM108 EXPIRES: B/31/BB FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (8) PACE (3) seaveNTIAL @j': REVISION NUMBER Washington Nuclear Plant - Unit 2 o 5 o o o 97 89 0 5 0 0 6 OF TEXT /// moFF 8/Moo /8 nqawod, vse 8~ NRC FomI 3BSAB/ (17) b) June 18, 1989 (1045 Hours) Event P roceaural Inadequacy: The procedure di d not 1 i st the corresponding drain valves required to test specific excess flow check valves, and instead relied upon the ability of the field technicians to trace the instrument tubing from the instrumentation to the drain valves. Had the drain valves been positively identified by tag number in the procedure, the need to trace tubing to the correct valve would have been eliminated and this event would not have occurred.
c) June 18, 1989 (1814 Hours) Event
~ Procedural Inadequacy: The procedure did not explicitly state the impact of isolating MS-PT-51A, nor did caution statement directing that the Control Room Operators be it provide a clear alerted that MS-PT-51A was isolated and pressure indication impaired. While the Test Director was aware that MS-PI-9 was being used by the Control Room Operators to monitor reactor pressure, it was not obvious that MS-PI-9 .input comes from MS-PT-51A. At the point where MS-PT-51A is isolated in the procedure, there is a caution to, "Inform the Control Room Operator that pressure indication for reactor vessel leakage test is affected PPM 7.4.0.5.38.U Because there were no leakage tests in progress at the time, the Test Director did not specifically inform the Control Room Supervisor of this instrument isolation. A contributing factor was that a verbal agreement had been established whereby the Test Director would inform the Control Room Supervisor whenever the technicians were moving to the next valve. In the case of PI-EFC-X114, this was not done. However, had the caution statement specifically stated that pressure indication on MS-PI-9 would not be available during testing of PI-EFC-X114 (and to alert the control room),'this event would not have occurred.
B. Further Corrective Action
- 1. June 17, 1989 Event
~ The procedure was modified to inform the Shift Manager that high reactor pressure (above 125 psig) would cause an RHR Shutdown Cooling isolation and a recommendation to control pressure at less than 125 psig was included. In addition, the limit for the minimum pressure required was reduced from 100 psig to 85 psig.
~ In addition, the procedure, was modified to provide a pressure band (85-110 psig) to avoid the possibility of a trip at an indicated valve below 125 psig.
NRC FORM 388D ~ O.S, CPOe IS88-828-SSO OOO)n (9 83)
NRC Form 38SA U.S, NUCLEAR REOULATORY COMMISSION (943)
LICENSEE EVE T REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3)50&104 EXPIRES: S/3) ISS FACILITYNAME (I) DOCKET NUMBER (2) LER NUMBER (6) PACE (3) nj SEOVENTIAL ;.>35 REVISION i??N NUMSER Washington Nuclear Plant - Unit 2 o 5 0 0 o 3 9 7 8 9 02 -5 0 0 0 7 OF 0 8 TEXT N moro g>>co Ir ro0rdrod, Irro odd)oorMl 1VRC %%drm 38543) ((7)
- 2. June 18, 1989 (1045 Hours) Event
~ The procedure was modified to require that the Test Director visually verify that the correct drain valve has been located prior to opening to test the excess flow check valve. This action was taken until such time that formal revision to the procedure could be made.
- 3. June 18, 1 989 ( 1814 Hours) Event
~ As a result of the management review previously discussed, additional changes to the procedure were made which included requiring that instruments valved out-of-service be caution tagged to indicate status.
- 4. The procedure (PPM 7.4.6.3.4.1) will be revised prior to the. next maintenance and refueling outage to identify the consequences of each instrument removed from service, and to also include specific valve, identification.
- 5. An overall evaluation of the RHR Shutdown Cooling isolations which have
- occurred is currently being performed by the Plant Technical and Nuclear Safety Assurance Groups. The generic issues associated with those RHR Shutdown Cooling isolations which occurred during the recent maintenance and refueling outage will be addressed in this evaluation. The evaluation is nearing completion and an internal formal report will be issued.
Safety Si nificance There is no safety significance associated with these events. For the loss of RHR Shutdown Cooling events, Shutdown Cooling was isolated for 39 minutes (June 17, 1989) and 31 minutes (June 18, 1989) respectively, well within the time frame allowed by the Technical Specifications.
Regarding the HPCS Diesel start, no Plant condition warranting the ESF actuation actually existed and the HPCS actuations occurred as designed.
Accordingly, these events posed no threat to the health and safety of either the public or Plant personnel-.
Similar Events There have been several events associated with the loss of RHR Shutdown Cooling; however, none with the same root cause (i.e., procedural inadequacies during excess flow check valve testing).
NRC FORM 388a 'V.S ~ CPOI )088 820-880 00010 (9 33)
NRC Form 366A U.S. NUCLEAR REOULATORY COMMISSION (84)3)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OM8 NO. 3150M(04 EXPIRES: 8/31/88 FACILITY NAME (1) OOCKET NUMBER (3)
LER NUMBER (6) PACE (3)
YEAR M4r SEOVSNTIAL +r8'EVISION NVMSSR KrrrI NVMSSA Washin ton Nuclear Plant - Unit TEXT /llmoro S>>co /S roqrr/rod, Irro odd/cior>>/HRC Form 355A'c/ (IT) 2 <<<< o 3 9 7 8 9 0 OF EIIS Information Text Reference EI IS Reference System Component Residual Heat Removal (RHR) System BO RHR-V-9 BO ISV Reactor Pressure Vessel (RPV) NH RPV Nuclear Steam Supply Shutoff System BD Control Rod Drive (CRD) System AA MS-P I -9 SB PI Reactor Recirculation (RRC) System AD RRC-PS-18B AD PS Reactor Protection System (RPS) JM High Pressure Core Spray (HPCS) BG HPCS Diesel Generator EK DG Excess Flow Check Valve PI-EFC-X72A NH V Containment Penetration X72A NH PEN MS-LT-27 SB LT RFW-DPT-17 (IB PDT Containment Penetration X100 NH PEN MS-LIS-24C SB LIS MS-L I S-31B SB LIS MS-L IS-31D SB LIS Containment Penetration X112 NH PEN MS-L I S-100B SB LIS HPCS-V-4 BG INV HPCS-P-1 BG P P I-EFC-X114 NH V Containment Penetration X114 NH PEN MS-PT-61A SB PT MS-LR/PR-623A SB L R/PR NRC FORM 388A ~ V.S. CPOr 1966 S20 SS9r00010 (8 83)