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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
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ACCEI.ERAT%0 DOCUMENT DISTR>VTION SYSTEM REGULA'l~INFORMATION DISTRIBUTIO
'YSTEM (RIDS)ACCESSION NBR:9310290064 DOC.DATE: 93/10/25 NOTARIZED:
NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFILIATION LEWIS,K.B.
Washington Public Power Supply System PARRISH,J.V.
Washington Public Power Supply System RECIP.NAME-RECIPIENT AFFILIATION
SUBJECT:
LER 93-028-00:on 930923,identified unanalyzed HELB in primary containment.
Caused by less than adequate design analysis a review of design analysis by architect engineer.Review will be performed of other piping sys.W/931025 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR L ENCL L SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES: RECIPIENT ID CODE/NAME PDV LA CLIFFORDgJ INTERNAL: ACRS AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRIL/RPEB
.NR/+SSE/~PLB RGN FILE 01 COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PDV PD AEOD/DOA AEOD/ROAB/DSP NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB, RES/DSIR/EIB COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 2 2 1 1 1 1'XTERNAL EG&G BRYCE E J~H NRC PDR NSIC POOREEW.2 2 1 1 1 1 L ST LOBBY WARD.1 1 NSIC MURPHY,G.A 1 1 NUDOCS FULL TXT'1 NOTE TO ALL"RIDS" RECIPIENTS PLEASF HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.504-2065)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDI FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 30 WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O.Box 968~3000 t eorge Washington Way~Richland, Washington 99352 October 25, 1993 G02-93-258 Docket No.50-397 Document Control Desk U.S.Nuclear Regulatory Commission Washington, D.C.20555
Subject:
NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EVENT REPORT NO.93-028-00 Transmitted herewith is Licensee Event Report No.93-028-00 for the WNP-2 Plant.This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportability, corrective action taken, and action taken to preclude recurrence.
If the corrective actions identified in the LER result in additional reportable information, the results will be documented as a supplement to this LER.Sincerely,@Crt/V.Parrish (Mail Drop 1023)Assistant Managing Director, Operations JVP/KBL/lr Enclosure cc: Mr.B.H.Faulkenberry, NRC-Region V Mr.R.Barr, NRC Resident Inspector (Mail.Drop 927N, 2 Copies)INPO Records Center-Atlanta, GA'r.D.L.Williams, BPA (Mail Drop 399)2S001S 9310Z'70064 9'310Z5 PDR ADOCK 050003'tt7 S PDR+1~28 LICENSEE EVEIOREPORT (LER)ACILITY NAME (1)Washin ton Nuclear Plant-Unit 2 DOCKET NUMB R ()PAGE (3)0 5 0 0 0 3 9 7 I OF'0 TITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT EVENT DATE (5)LER HUMBER{6)REPORT DATE (7)OTHER FACILITIES INVOLVED (8)MOHTM DAY YEAR YEAR c:?SEQUENTIAL NUMBER EV I 5 ION UMBER MONTH DAY YEAR FACILITY HAMES OCKET 50 NUMBE RS(S)0 9 2 3 PERATIHG OOE (9)0 0 1 0 2 5 9 3 9 3 9 3 0 2 8 5 0 00 HIS REPORT IS SUBMITTED PURSUAHT TO THE REQUIREMENTS OF 10 CFR E: (Check one or more of the following)
(11)I ONER LEVEL (iO)0.402(b)20.405(a)(l)(i) 20.405(a)(I)(I I)20.405(a)(1)(iii) 0.405(a)(1)(iv) 0.405(a)(1)(v) 0.405(C)50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73{a){2){iii) 0.73{a)(2){iv) 0.73(a)(2)(v) 0.73(a)(2)(vlf) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 0.73(a){2){x) 77.71{b)73.73(c)THER (Specify in Abstract elow and in Text.HRC orm 366A)LICENSEE CONTACT FOR THIS LER (12)Kurt B.Lewis, Technical Specialist REA CODE TELEPHOHE NUMBER 0 9 7 7-4 1 4 5 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IH THIS REPORT (13)CAUSE SYSTEM COMPONENT MANUFACTURER EPORTABLE 0 HPRDS CAUSE SYSTEM COMPONENT MANUFACTURER EPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14)YES (If yes, comp(ete EXPECTED SUBMISSION DATE)NO lCTIIACl OOI XPECTED SUSHI SSIOH MONTH DAY YEAR ATE (15)On September 23, 1993, with the reactor in MODE 1 at 100%power, a programmatic engineering review of high energy line break (HELB)analysis to confirm consistency between Leak Detection (LD)system capabilities and HELB analysis assumptions identified an unanalyzed HELB.The WNP-2 Final Safety Analysis Report (FSAR), Section 3.6, indicates that HELBs were analyzed for environmental effects, specifically temperature and humidity.However, the programmatic engineering review determined that these environmental effects were not analyzed for a postulated break in a four-inch Reactor Water Cleanup (RWCU)system line RWCU(5)-3.
The break is open to the Reactor Building environment.
Engineering immediately performed an operability evaluation for the unanalyzed HELB conditions and determined that the plant could continue to operate safely.Further corrective action consists of finalizing revisions to Supply System HELB analysis and requesting permanent exclusion of the postulated break.The root cause of this event was less than adequate design analysis and review of design analysis by the architect engineer.There were no contributing causes to this event, This event posed no threat to the health and safety of the public or plant personnel.
LICENSEE EVENT REPORT QR)TEXT CONTINUATION ACILITY HAHE (1).Washington Nuclear Plant-Unit 2 OOCKET HUHBER (2)0 5 0 0 0 3 9 7 ear LER HUHBER (8)umber ev.Ho.3 28 0 AGE (3)2 F 10 iTLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT Pl n ndi n Power Level-100%Plant Mode-1 (Power Operation)
Event Descri tion On September 23, 1993, with the reactor in MODE 1 at 100%power, a programmatic engineering review of high energy line break (HELB)analysis to confirm consistency between Leak Detection (LD)system capabilities and HELB analysis assumptions identified an unanalyzed HELB.The WNP-2 Final Safety Analysis Report (FSAR), Section 3.6;indicates that HELBs were analyzed for pipe whip, jet impingement, flooding, pressurization, and environmental effects, specifically temperature and humidity.However, the programmatic engineering review determined that environmental effects were not correctly analyzed for one HELB.This HELB involves a postulated break in four-inch Reactor Water Cleanup (RWCU)system line RWCU(5)-3 (Figure 1).Specifically, this break is located at RWCU system blowdown flow control valve RWCU-FCV-33 and is open to Reactor Building'loor elevation 501'..The postulated break location involves ASME Section III Class 3 piping.Postulated pipe break locations are based on guidelines provided by NRC Branch Technical Positions APCSB 3-1 and MEB 3-1 (as described in FSAR 3.6.2.1).In part, for ASME Section:III Class 3 piping, these guidelines require breaks to be postulated at terminal ends.For piping runs which are maintained pressurized for only a portion of the run, MEB 3-1 defines a terminal end as the piping connection at the first normally closed valve in the run.Accordingly, a break should be postulated at RWCU-FCV-33.
During power operation, this line operates at primary coolant pressure at an approximate temperature of 125 degrees Fahrenheit.
Immediate orrective Action Engineering performed an operability evaluation for the postulated HELB which included completing nondestructive examination testing of the postulated break location.The evaluation determined that the plant could continue to operate safely.The determination was based principally on the low stress levels at the postulated break locations, the positive results of the nondestructive examination testing, and diverse means to isolate the postulated RWCU line break automatically.
LICENSEE EVENT REPORT R)TEXT CONTINUATION AClLITY NAHE (1).Washington Nuclear Plant-Unit 2 00CKET NUHBER (2)0 5 0 0 0 3 9 7 LKR NUHBER (8)umber ev.No.3 28 0 AGE (3)3 F io ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIHARY CONTAINMENT Fu her Evalu i n R ause d orr iv Ac i n~hE 1.On September 23, 1993, at approximately 1357 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.163385e-4 months <br />, the failure to analyze two HELBs located outside Primary Containment was reported to the NRC by telephone in accordance with 10CFR50.72(b)(1)(ii)(B) which requires the Licensee, to notify the NRC within one hour of conditions outside the design bases of the plant.A further evaluation of this event concluded that more appropriate reporting criteria were 10CFR50.72(b)(1)(ii)(A) and 50.73(a)(2)(ii)(A),"...an unanalyzed condition that significantly compromises(ed) plant safety." Continued review showed only the postulated RWCU HELB met these reportability criteria.Further evaluations of the analysis associated with the second break involving the postulated Heating Steam Condensate (HCO)HELB concluded this condition was not reportable.
2.There were no structures, components, or systems inoperable prior to this event that contributed to the event.3.As stated, on September 23, 1993, a WNP-2 engineer discovered an unanalyzed HELB while performing the programmatic review as follows: a.The review was performed in response to General Electric Potentially Reportable Condition Report PRC 88-17,"Main Steam Tunnel Temperature Instrumentation and Isolation", issued in May 1989.This report consisted of three recommended actions.The first recommendation consisted of reviewing coverage of exhaust gas radiation monitoring capability.
The second recommendation consisted of reviewing reactor coolant pressure boundary leak detection/isolation initiation to confirm consistent application of the Leak Detection system design intent.These two recommendations were completed in January 1990 and March 1992respectively.
b.The third recommendation advised utilities to review"HELB analysis assumptions
...for installed LD system capability." The purpose of this recommendation was to identify potential oversights in the capability of the WNP-2 leak Detection system in light of information provided in the PRC 88-17 report.Preliminary review o'f this recommended action, as well as similar reviews performed prior to this recommendation did not discover any significant problems.Since this recommendation required extensive review for closure, the review competed with other priority tasks and resulted in an extended schedule for completion.
The review actually involved evaluation of over 160 break locations and 100 leak detection sensors.The postulated RWCU HELB was discovered near the end of the review.
LICENSEE EVENT REPORT R)TEXT CONTINUATION ACILITY HAHE (1)Mashington Nuclear Plant-Unit 2 DOCKET HUMBER (2)D 5 D 0 0 3 9 7 LER HUHBER (8)ear umber ev.Ho.3 28 0 AGE (3)4 F 10 ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED.OUTSIDE PRIMARY CONTAINMENT 4.Upon discovery of the unanalyzed HELB, engineering initiated an investigation to determine why the HELB was not evaluated.
The investigation determined that the list of bounding HELBs used to analyze corresponding environmental effects was developed from an earlier list of bounding HELBs that omitted the postulated break in line.RWCU(5)-3.
J g(~t~au e The root cause of this event was less than adequate design analysis and design analysis review by the architect engineer.The programmatic review performed by the Supply System determined that the architect engineer omitted the RWCU HELB from the list of bounding HELBs because the effects of the associated blowdown were incorrectly assessed.The architect engineer erroneously concluded that because water contained within this particular branch of piping was of moderate temperature, a postulated break at this location would have negligible environmental impact.However, the postulated RWCU HELB would cause a rapid increase in blowdown temperature until blowdown was terminated.
The rapid increase in temperature would occur because the postulated break would significantly reduce the amount of water returning to the reactor via the shell-side of the regenerative heat exchanger RWCU-HX-IA(B,C).
This would have the effect of removing the heat sink for water at reactor temperature entering the tube-side of the regenerative heat exchanger and thus causing a rapid increase in blowdown temperature.
Because this list later formed the list of bounding HELBs subsequently analyzed for environmental effects, the mistake was carried forward to that analysis.Prior to initial plant startup, the Supply System overview of the architect engineer's work was not to the level of detail where this error would be discovered.
There were no contributing causes to this event.rther orrective Acti n A review will be performed to ensure that other piping systems were not inadvertently omitted from consideration of HELB environmental effects.This review will evaluate the other piping systems eliminated from consideration by the architect engineer.This will be completed by June 1, 1994.2.A request for permanent exclusion of the postulated RWCU(5)-3 line break based on the as-built stress analysis and recent nondestructive examination results will be submitted to the NRC by December 3, 1993.
LICENSEE EVENT REPORTER)TEXT CONTINUATION ACILITT NAHE (I)Washington Nuclear Plant-Unit 2 DOCKET NUHBER (2)0 5 0 0 0 3 9 7 ear LER NUHBER (8)UIIIber ev.No.AGE (3)3 28 0 5 F 10 ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT Introdu i n The RWCU HELB is safety significant in that design basis environmental conditions in the Reactor Building would be exceeded if no automatic isolation were credited.However, the Supply System believes continued plant operation is justified based on the low stress levels associated with the postulated break, the results of recent nondestructive examination testing performed at the postulated break, and the diverse mitigation features equipment capable of isolating the postulated break if it occurred.Following is a discussion of the environmental effects, stress analysis, and mitigating features associated with the postulated RWCU HELB.Envir nmental Effect If no automatic isolation of the RWCU HELB were to occur, the environmental conditions utilized for qualifying various safety-related pieces of equipment in the Reactor Building would be exceeded.The affected area could be large.because open Reactor Building equipment access hatches could allow wide dispersement of the blowdown environment, NRC Branch Technical Positions APCSB 3-1 and MEB 3-1 require breaks to be postulated at terminal ends.For piping'runs which are maintained pressurized for only a portion of the run, MEB 3-1 defines terminal end as the piping connection at the first normally closed valve in the run.Accordingly, a break should be postulated at RWCU-FCV-33.
However, review of as-built stress analysis for the RWCU(5)-3 piping connection to RWCU-FCV-33 determined that calculated stresses due to various corresponding loading conditions are well below the ASME allowable values necessary to credibly postulate a HELB or a through-wall pipe crack:
L(CENSEE EVENT REPORT'QR)
TEXT CONTlNUATlON AClLITY NAHE (i)Washington Nuclear Plant-Unit 2 OOCKET NUMBER (2)0 5 0 0 0 3 S 7 LER NUHBER (8)ear umber ev.No.3 28 0 AGE (3)6 F 10 ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT ASME Sec.III, Class 3 Piping Stress Eq.Eq.8 Eq.9 Eq.10 Analyzed Stress Effect Drywell Tempera-ture (DWT)+Pressure DWT+Pressure+Operating Basis Earthquake (OBE)Thermal Stress Calculated Value 5,483 psi 9,838 psi 360 psi ASME Allowable Value 15,000 psi 18,000 psi 22,500 psi Stress-based pipe breaks and cracks are required to be postulated when the summation of ASME Equations 9 and 10 exceed a specified portion of the ASME Code stress allowable values (FSAR Sections 3.6.2,1.1.2 and 3.6.2.1.3).
The sum of ASME Equations 9 and 10 and the FSAR break and crack criteria are tabulated as follows: Summation of Calculated Values for ASME Eq.9 and 10 10,198 psi FSAR Stress Criteria for Full Guillotine Breaks>/=32,400 psi FSAR Stress Criteria for Through-Wall Cracks>/=16,200 psi In addition to the calculated stress values, nondestructive examination of the subject RWCU piping was performed on September 23, 1993, under Maintenance Work Request (MWR)AP5406.The results of this examination revealed no significant degradation or erosion/corrosion of the subject piping from past plant operation.
LICENSEE EVENT REPORT'R)TEXT CONTINUATION ACILITY NAHE (1)Washington Nuclear Plant-Unit 2 DOCKET NUHBER (2)0 5 0 0 0 3 9 7 LER NUHBER (8)eeI'IIIIber ev.No.3 28 0 AGE (3)7 F 10 ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT i i atin Feature There are three automatic isolation functions of the RWCU system not previously credited that will mitigate the consequences of the postulated RWCU HELB.These functions are the non-regenerative heat exchanger RWCU-HX-2A(B) high outlet temperature isolation, filter-demineralizer RWCU-DM-1A{B) high differential pressure isolation, and filter-demineralizer outlet resin trap RWCU-RST-70A(B) high differential pressure isolation.
The high temperature isolation function closes Primary Containment isolation valve RWCU-V-4 which isolates the flow of reactor coolant to the inlet of the RWCU system.The RWCU-V-4 valve is fully qualified and is designed to isolate a HELB in the RWCU system.The filter-demineralizer high differential pressure isolation and the filter-demineralizer outlet resin trap high differential pressure isolation each individually effects filter-demineralizer isolation by closure of filter-demineralizer inlet valve RWCU-V-206A{B) and filter-demineralizer outlet flow control valve RWCU-V-266A(B).
Although the three isolation functions are initiated by unqualified instruments and, with the exception of RWCU-V-4, are accomplished by unqualified valves, the Supply System believes these functions would provide reliable means of isolating the postulated RWCU HELB.None of the unqualified valves or instrumentation credited for isolating the postulated HELB would be affected by the, postulated break because this equipment is either located outside the Reactor Building or in areas unaffected by the break consequences.
Additionally, the Supply System reviewed the maintenance history of unqualified equipment including check valve RWCU-V-39 (isolates RWCU system return piping to the reactor), RWCU-V-206A(B), and RWCU-V-266A(B).
The review indicated that the valves have operated reliably.Further, the Supply System verified with the"206" and"266" valve vendor that these valves are capable of closing under the evaluated flow and differential pressure values associated with the postulated RWCU HELB.Lastly, the Supply System determined that the unqualified instrument sensors generating the referenced isolation functions have been recently calibrated.
If the postulated break were to occur, the break would cause high flow in the RWCU system, which in turn would cause the temperature of the water exiting the non-regenerative heat exchangers to increase.An associated temperature element (RWCU-TE-7) and temperature indicating switch (RWCU-TIS-8) would sense the rise in water temperature and would initiate automatic closure of RWCU-V-4 and subsequent trip of the RWCU system pumps (RWCU-P-1A(1B)).
The postulated RWCU HELB would be completely isolated, as RWCU-V-4 would isolate RWCU system inlet piping from the reactor, and check valve RWCU-V-39 would isolate RWCU system return piping to the reactor.The Supply System is confident that this particular isolation logic channel would respond as designed, as the Supply System has confirmed that the isolation function's Logic System Functional Test (surveillance procedure 7.4.3.2.2.16) was recently performed per schedule in June 1993 with successful response of the logic channel.This LSFT is of the same rigor as LSFTs associated with Class I components.
LICENSEE EVENT REPORT QR)TEXT CONTINUATION ACILITY NAME (1)Washington Nuclear Plant-Unit 2 DOCKET NUMBER (2)0 5 0 0 0 3 9 7 LER NUMBER (8)umber ev.No.3 28 0 AGE (3)8 F io ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT Our analysis indicates that RWCU system break flow would approach 2000 gpm within one second of the postulated break.This flow rate would cause the filter-demineralizer RWCU-DM-1A(B) high differential
'ressure setpoint to be exceeded which would initiate automatic closure of filter-demineralizer outlet flow control valve RWCU-FCV-266A(B) and, by interlock, closure of filter-demineralizer inlet valve RWCU-V-206A(B).
Similarly, break flow would cause the filter-demineralizer outlet resin-trap RWCU-RST-70A(B) differential pressure isolation setpoint to be exceeded which also causes automatic closure of RWCU-FCV-266A(B) and RWCU-V-206A(B).
The"206" and"266" valves are air-operated, fail-closed on loss-of-air and/or power and are spring-loaded to close.To evaluate the consequences of a postulated line break in the RWCU piping system, a conservative evaluation was performed based on the expected response of the previously referenced isolations., As stated, if the postulated break were to occur, high temperature in the RWCU system would initiate automatic closure of RWCU-V-4 and a subsequent trip of the RWCU pumps, as well as automatic closure of RWCU-V-266A(B) and RWCU-V-206A(B) due to high differential pressure across the associated filter-demineralizers and/or the filter-demineralizer resin traps, For conservatism, it was assumed that only RWCU-V-4 would close to mitigate the break.No credit for the quicker acting RWCU-V-266A(B) valves which would isolate on high differential pressure across the filter demineralizers and/or the associated'resin traps, is assumed in the analysis.Crediting RWCU-V-4 closure results in a more limiting system response time and associated environmental impact, The evaluation included the effects of high temperature water (440 degrees Fahrenheit) in the regenerative heat exchangers expanding and flashing into steam at the 501'elevation along with the pressurized lower temperature water (140 degrees Fahrenheit) on the filter-demineralizer side of the break.Results of the analysis indicated that peak temperature on the 501'levation would not exceed 160 degrees Fahrenheit and would drop below 120 degrees Fahrenheit within 600 seconds;humidity would peak at 100%.(Safety-related equipment in Reactor Building harsh environment areas has been evaluated and has been determined to be qualified for the environmental conditions of the event.The details of the above evaluation have been documented and filed with the plant problem report.imilar Events LER 85-001-01 identified nonconservative engineering assumptions associated with RWCU and Reactor Core Isolation Cooling (RCIC)system HELB calculations.
The calculations were used in developing Reactor Building environmental profiles which in turn were used in determining equipment qualification.
The Supply System made the discovery during in-house reviews of associated calculations performed by an independent contractor.
Corrective actions associated with this event were focused on modifying valve motor operators to decrease the severity of the environmental profile in the Reactor Building.
LICENSEE EVENT REPORT R)TEXT CONTINUATION ACILITY NAHE (I)Washington Nuclear Plant-.Unit 2 DOCKET NUHBER (2)0 5 0 0 0 3 9 7 ear LER NUNBER (8)unbent ev.No.3 2 8 0 AGE (3)9 F IO ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIMARY CONTAINMENT IIS Inf rma i n$g5~e g~m)~nen Leak Detection System (LD)Reactor Water Cleanup System (RWCU)Blowdown Flow Control Valve RWCU-FCV-33 Primary Containment Heating Steam Condensate (HCO)(part of Heating Steam (HS)System)Main Steam Tunnel Regenerative Heat Exchanger RWCU-HX-1A(B,C)
Drywell Non-Regenerative Heat Exchanger RWCU-HX-2A(B)
Filter-Demineralizer RWCU-DM-1A(B)
Resin Trap RWCU-RST-70A(B)
Primary Containment Isolation Valve RWCU-V-4 Check Valve RWCU-V-39 Inlet Valve RWCU-V-206A(B)
Outlet Flow Control Valve RWCU-FCV-266A(B)
Temperature Element RWCU-TE-7 Temperature Indicating Switch RWCU-TIS-8 RWCU System Pumps RWCU-P-lA(1B)
Reactor Core Isolation Cooling System (RCIC)CE CE NH SB CE NH CE CE CE CE CE CE CE CE CE CE BN FCV'HX HX FDM TRP ISV V ISV FCV TE TIS P LICENSEE EVENT REPORT Qh)TEXT CONTINUATION AGILITY NAME (I)Washington Nuclear Plant-Unit 2 OOCKET NUMBER (2)0 5 0 0 0 3 9 7 LER NUMBER (8)ear umber ev.No.3 28 0 AGE (3)10 F 10 ITLE (4)FAILURE TO ANALYZE A HIGH ENERGY LINE BREAK LOCATED OUTSIDE PRIHARY CONTAINMENT AEACTOR PRESSIIAE VESSEL (RPY)r r PRlMARV REACTOR COHTAIHMEHTr RECI ACULAll0 M Loops A" 5 8 r AEACTOR COOLANT TO AWCU SYSTEM FIGURE 1.SIMPLIFIED FLOW DIAGRAM: RWCU SYSTEM RWCU V 265A (B)RENN TRAP T OP 0P RWCV V 205A (8)TO RWCU VA ISOLATION FUNCTION TE I RWCV ALTER.DEMIMEAALQ ER (TYP 2)EACH DP SWITCH OUTPUTS TO RWCU V.265A(S)AND RWCU V 205A(B)ISOLATION FUNCTION RWCU.V4 r//r RWCU V.S TO RPV VIA PEEDWATER SYSTEM (TYP 2)RWCV PUMP RWCU V@0 POSTULATED AWCU HELD AT PIPINO CONNECTION TO VALVE RWCV RED ENERATlVE HEAT EXCHAHOER (TYP S)AWCU.V@2 RWCU NOH.REO ENEAATIVE HEAT EXCHAHO ER (TYP 2)REACTOR COOLANT RETURN TO RPV 5 RWCU(2)2 BI.OWDOWM TO MAIM CONDENSER OR RADWASTE r Rwcu(s)2 SLOWDOWN COHTROLVALVE RWCU.FOES Ir X 4 REDUCER 5 RWCU(S)4