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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17335A5641999-10-18018 October 1999 LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr ML17335A5531999-10-0707 October 1999 LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented ML17326A1291999-09-17017 September 1999 LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary ML17326A1121999-08-27027 August 1999 LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed ML17326A1011999-08-26026 August 1999 LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs ML17326A0911999-08-16016 August 1999 LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4 ML17326A0771999-08-0404 August 1999 LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830 ML17326A0741999-07-29029 July 1999 LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves ML17326A0661999-07-26026 July 1999 LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr ML17326A0651999-07-22022 July 1999 LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis ML17326A0491999-07-13013 July 1999 LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr ML17326A0331999-07-0101 July 1999 LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written ML17326A0151999-06-18018 June 1999 LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified ML17325B6311999-06-0101 June 1999 LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions ML17325B6421999-06-0101 June 1999 LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed ML17325B6351999-05-28028 May 1999 LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions ML17265A8231999-05-24024 May 1999 LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made ML17325B6001999-05-20020 May 1999 LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed ML17325B5861999-05-10010 May 1999 LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised ML17325B5811999-05-0404 May 1999 LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared ML17325B5771999-05-0303 May 1999 LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design ML17325B5581999-04-16016 April 1999 LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold ML17325B5471999-04-12012 April 1999 LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation ML17325B5321999-04-0707 April 1999 LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate ML17325B5161999-04-0101 April 1999 LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures ML17325B5221999-03-29029 March 1999 LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing ML17325B4801999-03-18018 March 1999 LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling ML17325B4741999-03-18018 March 1999 LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed ML17325B4571999-02-24024 February 1999 LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors ML17335A5171999-02-11011 February 1999 LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted ML17335A5141999-02-10010 February 1999 LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405 ML17335A5011999-02-0101 February 1999 LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr ML17335A4951999-01-29029 January 1999 LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr ML17335A4961999-01-27027 January 1999 LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr ML17335A4921999-01-19019 January 1999 LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components ML17335A4721999-01-0606 January 1999 LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing ML17335A4691999-01-0606 January 1999 LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr ML17335A4661999-01-0505 January 1999 LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing ML17335A4631999-01-0404 January 1999 LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr ML17335A4481998-12-30030 December 1998 LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr ML17335A4581998-12-28028 December 1998 LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process ML17335A4281998-12-22022 December 1998 LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event ML17335A4111998-12-17017 December 1998 LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr ML17335A4141998-12-16016 December 1998 LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122 ML17335A4181998-12-16016 December 1998 LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria ML17335A3871998-12-11011 December 1998 LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing ML17335A3821998-12-0808 December 1998 LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr ML17335A3781998-12-0707 December 1998 LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr ML17335A3771998-12-0303 December 1998 LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned ML17335A3741998-12-0202 December 1998 LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3 1999-09-17
[Table view] Category:RO)
MONTHYEARML17335A5641999-10-18018 October 1999 LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr ML17335A5531999-10-0707 October 1999 LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented ML17326A1291999-09-17017 September 1999 LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary ML17326A1121999-08-27027 August 1999 LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed ML17326A1011999-08-26026 August 1999 LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs ML17326A0911999-08-16016 August 1999 LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4 ML17326A0771999-08-0404 August 1999 LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830 ML17326A0741999-07-29029 July 1999 LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves ML17326A0661999-07-26026 July 1999 LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr ML17326A0651999-07-22022 July 1999 LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis ML17326A0491999-07-13013 July 1999 LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr ML17326A0331999-07-0101 July 1999 LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written ML17326A0151999-06-18018 June 1999 LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified ML17325B6311999-06-0101 June 1999 LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions ML17325B6421999-06-0101 June 1999 LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed ML17325B6351999-05-28028 May 1999 LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions ML17265A8231999-05-24024 May 1999 LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made ML17325B6001999-05-20020 May 1999 LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed ML17325B5861999-05-10010 May 1999 LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised ML17325B5811999-05-0404 May 1999 LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared ML17325B5771999-05-0303 May 1999 LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design ML17325B5581999-04-16016 April 1999 LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold ML17325B5471999-04-12012 April 1999 LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation ML17325B5321999-04-0707 April 1999 LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate ML17325B5161999-04-0101 April 1999 LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures ML17325B5221999-03-29029 March 1999 LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing ML17325B4801999-03-18018 March 1999 LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling ML17325B4741999-03-18018 March 1999 LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed ML17325B4571999-02-24024 February 1999 LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors ML17335A5171999-02-11011 February 1999 LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted ML17335A5141999-02-10010 February 1999 LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405 ML17335A5011999-02-0101 February 1999 LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr ML17335A4951999-01-29029 January 1999 LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr ML17335A4961999-01-27027 January 1999 LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr ML17335A4921999-01-19019 January 1999 LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components ML17335A4721999-01-0606 January 1999 LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing ML17335A4691999-01-0606 January 1999 LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr ML17335A4661999-01-0505 January 1999 LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing ML17335A4631999-01-0404 January 1999 LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr ML17335A4481998-12-30030 December 1998 LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr ML17335A4581998-12-28028 December 1998 LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process ML17335A4281998-12-22022 December 1998 LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event ML17335A4111998-12-17017 December 1998 LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr ML17335A4141998-12-16016 December 1998 LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122 ML17335A4181998-12-16016 December 1998 LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria ML17335A3871998-12-11011 December 1998 LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing ML17335A3821998-12-0808 December 1998 LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr ML17335A3781998-12-0707 December 1998 LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr ML17335A3771998-12-0303 December 1998 LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned ML17335A3741998-12-0202 December 1998 LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3 1999-09-17
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999 LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr ML17335A5531999-10-0707 October 1999 LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented ML17335A5631999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr ML17335A5621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr ML17335A5481999-09-30030 September 1999 Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3). ML17335A5451999-09-28028 September 1999 Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios. ML17326A1291999-09-17017 September 1999 LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary ML17326A1481999-09-17017 September 1999 Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2. ML17326A1211999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr ML17326A1201999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr ML17326A1121999-08-27027 August 1999 LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed ML17326A1011999-08-26026 August 1999 LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs ML17326A0911999-08-16016 August 1999 LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4 ML17326A0771999-08-0404 August 1999 LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830 ML17335A5461999-08-0202 August 1999 Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates. ML17326A0871999-07-31031 July 1999 Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr ML17326A0861999-07-31031 July 1999 Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr ML17326A0741999-07-29029 July 1999 LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves ML17326A0661999-07-26026 July 1999 LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr ML17326A0651999-07-22022 July 1999 LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis ML17326A0491999-07-13013 July 1999 LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr ML17326A0331999-07-0101 July 1999 LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written ML17326A0511999-06-30030 June 1999 Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr ML17326A0501999-06-30030 June 1999 Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr ML17326A0151999-06-18018 June 1999 LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified ML17325B6421999-06-0101 June 1999 LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed ML17325B6311999-06-0101 June 1999 LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions ML17326A0061999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr ML17326A0071999-05-31031 May 1999 Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr ML17325B6351999-05-28028 May 1999 LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions ML17265A8231999-05-24024 May 1999 LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made ML17325B6001999-05-20020 May 1999 LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed ML17325B5861999-05-10010 May 1999 LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised ML17325B5811999-05-0404 May 1999 LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared ML17325B5771999-05-0303 May 1999 LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design ML17335A5301999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr ML17335A5291999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr ML17325B5581999-04-16016 April 1999 LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold ML17325B5471999-04-12012 April 1999 LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation ML17325B5321999-04-0707 April 1999 LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate ML17325B5161999-04-0101 April 1999 LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures ML17325B5491999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr ML17325B5441999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr ML17325B5221999-03-29029 March 1999 LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing ML17325B4801999-03-18018 March 1999 LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling ML17325B4741999-03-18018 March 1999 LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed ML17325B5671999-03-0202 March 1999 Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218. ML17325B4631999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr ML17325B4621999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr ML17325B4571999-02-24024 February 1999 LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors 1999-09-30
[Table view] |
Text
AC CELE RATED D1 BUTl 0 N DE }d 04 SYSI'EM REGULATO STATION INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9002280017 'OC.DATE: 90/02/09 NOTARIZED: NO DOCKET I FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana & 05000316 AUTH. NAME 'UTHOR AFFILiATION DROSTE,J.B. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele BLIND,A.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 90-002-00:on 900110,MSIV inoperability due to condensate accumulation on vent side of operating piston.
DZSTRZBOTZON CODE: ZE22T COPZES RECEZVED:LTR j ENCL [
TITLE: 50.73/50.9 Licensee Event Report (LER), Incidenti Rpt, etc.
NOTES:
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RECIPIENT COPIES ~
RECIPIENT COPIES XD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL h PD3-1 LA 1 1 PD3-1 PD 1 1 GIITTER,J. 1 1 INTERNAL: D ACNW 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 S
DEDRO 1 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB9H3 1 1 NRR/DET/ESGB 8D 1 1 NRR/DLPQ/LHFB11 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEABll 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRRQ~R/ LB8Dl 1 1 NRR/DST/SRXB 8E 1 1 EG FXL 02 1 1 RES/DSIR/EIB 1 1 RGN3 FILE 01 1 1 EXTERNAL EG&G WILLIAMS E S 4 4 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHYgG A 1 1 R NUDOCS FULL TXT 1 1 I
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LXSTS FOR DOCUMEMIS YOU DC%IT NEED!
FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPXES REQUIRED: LTTR 35 ENCI 35
Indiana Michigan Power Company Cook Nuclear Plant P.O. Box "58
.Bndgman. Ml '9106 616 465 5901 INDIANA NICHIGAN POWER February 9, 1990 United States Nuclear Regulatory Commission
.Document Control Desk Rockville, Maryland 20852 Operating License DPR-74 Docket No. 50-316 Document Control Manager:
In accordance, with the criteria established by 10 CFR 50.73 entitled Licensee Event, Re ortin System, the following report is being submitted:
90-002-00 Sincerely, A.A. Blind Plant Manager AAB:clw Attachment cc D. H. Williams, Jr.
A.B. Davis, Region III M.P. Alexich P.A. Barrett J.E. Borggren R.F. Kroeger NRC Resident Inspector J.G. Giitter, NRC R.C. Callen G. Charnoff, Esg.
Dottie Sherman, ANI Library D. Hahn INPO S.J. Brewer/B.P. Lauzau
@00224<00/7 +00209 PDR ADOCI 0<00031b PDr:
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (689) APPROVED OMB NO. 31504104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLFCTION REQUEST: 50.0 HRS, FORWARD LICENSEE EVENT REPORT ILER) COMMENTS AEGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P630), V.S. NUCLEAR REGULATOAY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503.
FACILITY NAME (I) DOCKET NUMBER (2) PA E D. C. Cook Plant - Unit 2 0 5 0 0 0 3 1 6 1 OF "P TITLE<<I Main Steam Isolation" Valve Inopera i ity - Due to Con ensate Accumulation on Vent 1
Side of Operating Piston EVENT DATK (5) LER NUMBER (41 REPORT DATE,(7) OTHER FACILI'TIES INVOLVED (4)
MONTH DAY SEQUENTIAL IIEVISION FACILITY NAMES YEAR YEAR NVMSEtl re NVMSEII MONTH OAY YEAR DOCKET NUMBER(SI Cook Plant - Unit 1 0 5 0 0 0 3 1 5 0 1 1 0 9 0 9 0 0 0 2 0 0 0 2 0 9 9 0 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR ('): (Cntce ont or mort ot rnt lottowrnd) (11 OPE RATING MODE IS) 3 20.402(SI 20.405(c) 50.73(el(2)(lv) 73.7)(S)
POWER 20.405 1 ~ I(lI(ll 50.34(cl(l I 73.71(c)
LEVEL 50.73(el(2llv)'0.73(el(2)(vSI p p p 20 406( ~ )(1)(lll 50.34(cl (2) OTHKR (Sptcrry in Aprtrtct Strow tnd in Ter I, HRC Form 20.405 ( ~ I( I I(IIl) X 60.73(e l(2) (I) .
50.73lel(2)(vSI)(AI 3ddAI 20.406(el(II(NI 60.73(e)(2)(S) I( 5) '0.73(e)12)(el)I 20.405( ~ l(l ) (vl 50.73(e l(2) Ill)) 50,73 ( ~ I (2 I (e I LICENSEE CONTACT FOA THIS LER (12I NAME TELEPHONE NUMBER AREA CODE J. B. Droste - -Technical Engineering Superintendent 6 1 6 4 65 - 5 901 COMPLETE ONE (.INE FOA EACH COMPONENT FAILURE DKSCRIBED IN THIS REPORT (13)
\
CAUSE SYSTEM 'COMPONENT MANUFAC EPORTABLE MANUFAC TVRER TO NPRDS CAUSE SYSTEM COMPONENT TVRER B S B VT V X 9 9 9. Y
'Prr., ~ 4. N Xrr.r..
SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR FXPECTED'UbMISSION DATE (ISI YES ttt ytr, comprtrt FXPECFED SVSMISSION DAFEI X ABS'TAACT tLimrt ro tt00 Iptctt, r,t., tpprorimtttiy trttttn tin(lit toter ryptwrirttn tmttt (14)
On January 8, 1990 with Unit 2 in Mode 3- (Hot Standby) surveillance testing of the Main Steam Isolation Valves (MSIV's) was conducted. Two MSIV's exhibited a closing time in excess of the Technical Specification 3.7.1.5 limit of five seconds. This condit'ion is believed .to have, existed during power operation. Additional MSIV testing on January 9, and 10, 1990 confirmed that excessive condensation was collecting on the vent side of the MSIV operating piston. When the dump valves were opened to vent-off steam from the MSIV operating piston, the accumulated condensate would flash to steam and result in increased MSIV closure times. Following an effort to blow out the MSIV condensate drain tubes, a series of tests were performed at various time intervals. to determine if the condensate problem had been resolved and verify MSIV operability. With the failure of 2-MRV-220 on January 11, 1990, at 0916 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.48538e-4 months <br />, it was evident that the condensate p'roblem was still present. All MSIV's were declared inoperable.and an Unusual Event was declared. A cooldown to Mode 5 (Cold Shutdown) was started.
Repair activities included disassembly of the MSIV's to enlarge the condensate drain tube port and the equalizing steam nipple. Additional insulation was placed on the MSIV's, vent piping and valves.
NRC Form 366 (669)
NRC FORM 366A V.S. NUCLEAR REGULATORY COMMISSION (6J)9) APPROVED OMB NO. 31500)OS EXPIRES: S/30/92 LICENSEE EVENT REPORT (LER) ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION" AND REPORTS MANAGEMENT BRANCH (P430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555. AND TO THE PAPFRWORK REDUCTION PROJECT (3(500104). OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, OC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR SEQVENTIAL REVISION NVMEER NVMSER
-D. C. Cook Plant -
TEXT /// more spece ie reqvirN/. Irse ~ Unit 2 HRC %%drm 3(SEA's/ (12) o s o o o 3 1 6 9 0 0 0 2 0 0 0 0 CONDITIONS PRIOR TO OCCURRENCE Unit 2 in Mode 3 (Hot Standby).
DESCRIPTION OF EVENT On January 8, 1990, surveillance testing of the Main Steam Isolation Valves (MSIV'-s) (EIIS/SB-ISV) was conducted. Two of the MSIV's exhibited a closing time in excess of the Technical Specification 3.7.1.5 limit of five seconds.
Following each MSIV failure, the MSIV was declared inoperable and immediately retested. In both cases the retest demonstrated the valve closed within the Technical Specification time limit and the, MSIV was declared operable.
Additional tests were conducted on January 9, 1990. A substantial amount of water was observed to flash out of the dump valve (EIIS/SB-VTV) vent stacks during the first'est of each MSIV. Three MSIV's. exhibited a closing time in excess of the Technical Specification 3.7.1.5 limit of five seconds.
Following each MSIV failure, the MSIV was declared inoperable and retested.
In- each case the retest demonstrated the valve closed within the Technical Specification time limit and the MSIV's were declared operable. (See attached test table).
The January 9, 1990 testing verified that enough condensation could accumulate on the vent side of the piston within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to'ender the valve inoperable. Unit 2 was held in Mode 3 to further diagnose the root cause of this issue.
On January 10, '1990, a series of diagnostic tests were performed to determine the extent that condensate within the MSIV operating cylinder volume could affect MSIV stroke time. A strip chart was used to record pressure within the volume as a function of time. Closure time was measured in the control room, measuring the time from the dump valve switch actuation until the MSIV valve light indicated the valve was closed. MSIV stem travel time was measured locally. The following time. line is similar to each MSIV tested and represents typical circumstances experienced by all MSIV's.
0 1 Sec: 2 Sec. 3 Sec. 4 Sec. 5 Sec. 6 Sec.
TIME CONTROL ROOM SWITCH ACTUATION TO LIGHT MSIV STEM TRAVEL FLASHING r NRC Form 366A (669)
NRC FORM 355A U.S. NUCLKAR REGULATORY COIAMISSION APPROVED OMS NO. 31504104
~ )549)
EXPIRES. 4/30/92 ESTIMATED bURDEN PER AESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST) 500 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO THE RECORDS TEXT. CONTINUATION AND REPORTS MANAGEMENT SRANCH IP430). U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. OC 20555, AND TO 1HE PAPERWORK REDUCTION PAO/ECT 13150410i)i OFFICE OF MANAGEMENTAND SUDQET, WASHINGTON. DC 20503.
FACILITY NAME l1) DOCXET NUMSER (21 LER NUMSER IS) ~ AGE 13)
YEAR SEQUENTIAL REVISION NUM Kll @ NUM 9 II D, C. Cook Plant - Unit 2 2.' 000 OF o s o o o 3 1 6 9. 0 0 0 1 0 TEXT /// vere PPPCP )I /Pquia/, MPP //I/bbP//YRC %%dnt) 30SA'P/ )IT)
Local observation verified substantial quantities of water and steam being discharged from the vent stack from T 0 until MSIV stem travel began.
Two of the MSIV's exhibited a closing time in excess of the Technical Specification 3.7.1.5 limit of five seconds. 'Following each MSIV failure, the MSIV was declared inoperable and retested. In each case the retest was an acceptable value and the MSIV was declared operable (see attached test table).
The Action Requirements of Technical Specification 3.7.1.5 were fulfilled at all times for Modes 2 and 3. When an MSIV fai'lure occurred, an immediate retest demonstrated that the valve was operable. On January 10, 1990, it was realized that the failures which occurred on January 8, 1990, were believed to have existed during power operation. Notification to the NRC was made on January 10, 1990, reporting that we were outside the plants'esign bases on January- 8, 1990. At the time this notification was made, the MISV's were operable per Technical Specification Action Requirements for Modes 2 and 3 operation. The MSIV's were operable since the testing resulted in expelling the condensate, from the vent side of the valve's operating piston. The MSIV's would remain operable as long as the valves were cycled frequently enough td prevent excessive amounts of condensate from collecting on top of the operating piston. ~
The MSIV testing confirmed that the slower, main steam stop valve stroke times were symptoms of an unacceptab1e quantity of condensate accumulating in the upper valve volume; When a dump valve is opened the condensate flashes to steam 'this causes increased venting time due to delaying the development of sufficient Delta-P across the MSIV piston, to close the MSIV's. A"logical hypothesis was formulated that condensate was'not draining down into the lower volume through the drain tube (see attached drawings) fast enough to prevent accumulation in the upper volume. Acting under the premise that this flow path was in some way restricted, possibly due to mineral deposits plating out on the inside diameter of the drain tube, it was decided to attempt to blow-out this drain tube.
A process was devised to remove (blow-out) any possible restrictions located within each MSIV drain tubes. This was accomplished by maintaining the MSIV in its closed position and for approximately one hour with the dump valves open. In'doing so, main steam was capable of forcing its way past the parallel closure disc, through the drain tube into the upper volume, and exhausted out the vent stack., It was hoped that this exhausted steam would be of sufficient velocity to carry away any restrictive material.
NAC Form 395A 1509)
NRC FORM SSSA ISeS) t LICENSEE EVENT REPORT ILER)
TEXT CONTINUATION P
U.S. NUCLEAA REGULATORY COMMISSION APPAOVED OMS NO. S)504104 E)IPIR ES: S/20/02 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPOATS MANAGEMENT BRANCH IPe30). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PRO/ECT 121504)OS). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 2050K IrACILITYNAME )1) DOCKET NUMBER )2) LER NUMBER IS) PAGE IS)
YEAR 4g sEovsNTIAL oE IISVISION NVMSSR NVM SN D. C. Cook Plant - Unit 2 o s o o o '3 1 6 9 0 0 0 2 0 0 0 4. OF 1 0 TEXT lllmore <<eoe /e reevrw/, ow //r/orO/ //RC %%dnII SSSAB/ ) 12)
Additional testing was performed to determine if the maintenance action was successful and ensure .MSIV operability. This series of tests had staggered intervals to ensure MSIV operability w'as maintained. On January 11, 'est 1990, at 0916 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.48538e-4 months <br />, 2-MRV-220 exceeded the allowed closing time. An Unusual Event was declared as all four (4) MSIV's were declared inoperable causing a Technical Specification 3.0.3 shutdown. This was classified as a 'One-Hour
,Reportable Event and NRC notification was made at 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />. A cooldown to Mode 5 (Cold Shutdown) was commenced.
CAUSE OF EVENT Investigation revealed that the failure of'he MSIV's was caused by excessive condensate accumulation on the vent side of the MSIV 'operating piston. The excessive condensate accumulation was a result of the following contributing factors:
Insulation Less than Current Desi n-Insulation on MSIV, the vent piping and associated valves was less than current design specifications. The insufficient insulation resulted in increased condensation rates in the MSIV.
- 2. Dum Valve Seat Leaka e The quantity of condensate that can collect is related to the mass of saturated steam permitted to enter the upper equalizing steam volume.
When steam is allowed to escape this volume, it will be replaced by additional saturated steam which increases the Delta-P and prevents drainage.
- 3. MSIV Actuator Desi n The combined effects of one and two above resulted in increased condensation rates, developing greater Delta-P across the operating piston than assumed by design during normal operation (MSIV's in steady-state open position). The rate at which condensate can flow down from the upper volume through the drain tube is limited by the diameter of the drain tube port holes. Condensate will accumulate within the upper equalizing volume if it cannot be drained faster than it forms.
If the steam nipple is insufficiently sized, a Delta-P will exist across the piston restricting condensate drainage.
NRC Fomr SSSA ISeS)
NRC FORM 355A US. NUCLEAR REGULATORY COMMISSION APPROVEO OM8 NO, 3150010>>
ISJ)91 EXP IR ES: >>/30/92 ESTIMATED 8UROEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REOUESTI 50.0 HRS. FORWARD COMMENTS REGARDING 8URDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT SRANCH IP430). U.S. NUCLEAR s REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT 13150010l). OFFICE OF MANAGEMENTAND 8UDGET, WASHINGTON, DC 20503.
FACILITY NAME 111 DOCKET NUMSER 12) LER NUMSER 15) ~ AGE 13)
YEAR ,>>jj SEOVENTIAL HVM 5<<
yg: AEVOON
<~+ HVM D. C. Cook Plant - Unit 2 o s 0 o o 3 1 6 9 0 0 0 2 0 0 0 50F 1 0 TEXT ////>>P/>> <<eC>> /1 I>>qvnd, u Je PI/P5/>>/ HJIC fcnn 38559/ )12)
ANALYSES,OF EVENT In addition to Technical Specification 3.7.1.5, slow MSIV closure times affect Technical Specification items 3.3.2.1, Table 3.3-5, 5.h (High Steam Line Flow - Low Low Tavg), 6.h (Steam Line Pressure Low), and 7.c (Containment Pressure-High High). These items use the valve closure time as part of their overall time response requirement. Examination of surveillance data for these items show that item 7.c, steam line isolation on high-high containment pressure in seven seconds, would not be met. This was reported to the NRC on January 10, 1990, under the provisions of 10 CFR 50.72.b.l.ii.B as a condition that was outside of the plant's design basis.
The investigation of this problem has identified that the slower closure times were caused by water accumulation on the top of the steam piston. With water present, the piston movement is slowed as the water flashes to steam during the venting process. The water. accumulation occurs over time and is also affected by the leakage past the MSIV dump valves. Once the water is removed from the top of the piston, as would occur during a valve test,. a retest conducted immediately thereafter would indicate normal va3ve actuation times.
As water would accumulate during operation, there is evidence that the condition may have existed prior to the surveillance. Thus, it is probabl'e that Technical Specification 4.7.1.5.1 was not met.
Westinghouse has evaluated the consequences of increasing all applicable Technical Specification response times by three seconds, and they have concluded that the FSAR acceptance criteria are met using the longer response times. The mass and energy releases for an outside of containment steam line break were compared with the current FSAR values, and the'nvironmental qualification of the equipment in the affected locations has not been compromised. Thus, there has been no significant degradation of plant safety as a result of the increased MSIV closure times.
Since Unit 1 has the same type of valve, an evaluation was also performed concerning the 'impact on .Unit 1. The evaluation concluded that the MSIV's on, Unit 1 are considered to be operable and fully supportive of the Chapter 14 accident analyses. At the last surveillances, all Technical Specifications applicable to steam line isolation were met, and the historical data indicate no pattern of closure time deterioration. Further, the recent Westinghouse analysis shows that sufficient margin exists with MSIV closure time such that plant safety is not compromised if the'system response time is increased by three seconds. Thus, the recent events associated with the Unit 2 MSIV's do not cause undue concern with respect to the continued safe operation of NRC Fons 355A (5891
NRC FORM 344A U.S. NUCLEAR REGULATORY COMMISSION APPROVEO OMS NOl31500)0O
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ExpIREs: el30)92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT. CONTINUATION AND REPORTS MANAGEMENT SRANCH (P430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 131504)104), OFFICE OF MANAGEMENTAND SUOGET.WASHINGTON. DC 20503.
FACILITY NAME 11) DOCKET NUMSER 12) LER NUMSER (4) I'AGE 13)
YEAR rX<O SEQUENTIAL ?P~W IIEYISION NVMFEA i.rX NVM D. C. Cook Plant - Unit 2 0 0 o s o o o 3 1 6 9 0 0 2 0 06 oF1 0 TEXT N more a>>oo Jr reqond, Irre orroeo'oner HRC form SOD'e) ) 12)
Unit 1. This conclusion is als'o based on the belief that the present performance of the Unit 1 MSIV's do not result in the probability or consequences of an accident being increased, any new type of accident being created, and any Technical Specification margin of safety being reduced.
CORRECTIVE ACTIONS Upon reaching Mode 5, all four MSIV's were disassembled and a visual examination performed. These examinations revealed that the drain tubes had not experienced any physical degradation that could have restricted the condensate flow path. All dimensions were verified. Electric ITE de France (FDF) was contacted concerning their approach at resolving similar circumstances. EDF provided their investigation findings and valve design modifications. We implemented the EDF recommendations.
After all available sources of information were technically evaluated, it was determined that all Unit 2 MSIV's would be modified via Minor Modification 2-Mf-079. Under this modification the diameter of drain tube port holes would be increased from 1'/8" (0.125") to 5mm (0.197") and the diameter of the equalizing steam nipple would be increased from 1/8" (0.125")'o 10mm (0.394"). Two performance related functions would be improved in this manner.
- l. Increasing the size of the drain tube port holes provid'es a larger area to improve condensate flow through the actuator piston.
- 2. Increasing the size of the equalizing steam nipple will provide a less restrictive path for steam to pass up through the piston to maintain equilibrium and prevent steam from passing through the drain tube and inhibit drainage.
It was also discovered at this time that the insulation on the MSIV, vent piping and associated valves was less than the current design specification requirements.. This contributed to a larger variance in temperature (Delta-T) resulting in more heat transfer and increasing the condensation rate.
After completing all modifications and reassembling the,MSIV's, Special Procedure 2-THP.SP.MM-079 was used to verify operability. This procedure included a series of test cycles for all MSIV's at different intervals with the appropriate instrumentation required to measure and record all applicable performance variables, specifically the rate of pressure drop in the vent line.
NRC Form 344A I5$ 9)
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NRC FORM SSSA U.S. NUCLEAR REGULATORY COMMISSION (5JISI APPROVED OMS NO. S1500104 EXPIRES: A/SO/52 ESTIMATED SUROEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT SRANCH IP 5201. U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (S(5041(M). OFFICE OF MANAGEMENTAND SUDGET, WASHINGTON, OC 20502, FACILITY NAME (1) DOCKET NUMSER (21 LER NUMSER (5) ~ AGE LSI YEAR YP>', SSQVSNTIA'L II5 Y IS IO N NVM Sll NVM SA D. C. Cook Plant - Unit 2 0 0 0 I 0 70F1 0 o s o 0 0 3 1 6 9 0 2 0 TEXT IIImcm apece JI IPSw'wI, PPP aAWonV HRC RPmI ~SI (12(
The acceptance criteria used within '2-THP.SP.MM-079 was twofold. First, each full closure stroke time must be within 5 seconds in accordance with Technical Specification 3.7.1.5. Second, the rate of pressure, drop in the vent line must not change as the test interval is extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This second criteria was used to verify no additional condensate had accumulated within the upper steam volume that could lengthen stroke time and delay stem travel upon flashing to steam during the venting process.
All stroke tests performed in accordance with 2-THP.SP.MM-079, except the first test on MRV-210, provided .acceptable results and verified the generic condensate issue. had been resolved; The initi'al test performed on MRV-210 failed. The corresponding rate of pressure drop in the vent line signature verified significant quantities of condensate still present. Investigation .
revealed that Train B Dump Valve MRV-212 was not se'sting properly permitting a significant amount of- steam to be exhausted out the vent stack. Under these circumstances, additional steam was brought into the upper volume with condensate collecting at a rate faster than could be drained. MRV-212 was repaired and subsequent closure tests completed on MRV-210 verified that this
.isolated incident had been resolved. All MSIV's were returned to an operable status and Unit 2 returned to power operation.
As stated in the analysis, we do not expect to encounter similar difficulties with the Unit 1 MSIV's. However, we are currently adding insulation to the Unit 1 MSIV's, vent lines and valves to enhance the MSIV operating conditions.
I FAILED COMPONENT IDENTIFICATION Component ID: Main Steam Isolation Valves 2-MRV-210, 2-MRV-220, 2-MRV-240 Manufacturer: Hopkinsons - Ferranti Model: 2379 W PREVIOUS SIMILAR EVENTS 050-316/83-57 NRC Form SSSA (SSSI
APPROVEO NRC FORM 344A
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LICENSEE EVENT REPORT {LERI TEXT CONTINUATION t 0MB NO. 31500104 EXPIRES)4/30/93 ESTIMATED SURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION RFOUESTI 500 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO THE RF CORDS AND REPORTS MANAGEMENT BRANCH IPS30), U.S, NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 30555, AND TO THE PAPERWORK REDUCTION PROJECT 131504))041-OFFICE OF MANAGEMFNTAND SUDGET,WASHINGTON. DC 30503.
fACILITYNAME Il) DOCKET NUMSER )3) LER NUMSER )4) ~ AGE 13)
YEAR SCOVCNTIAL IICVISION NVM CR NVM CA D. C.
TEXT Cook Plant - Unit lllmorp gfdcg lg nqvssd, we ~ 2 HRC Jtgwii SRSC'4) (17) o e o 0 o 31690 0 0 2 0008 oF 1 0 MAIN STEAM ISOLATION VALVE TEST TABLE Januar 8, 1990 Testin VALVE INITIAL CLOSING TIME (SEC) RETEST (SEC) 2-MRV-210 =
5.94 3.01 2-IIRV-240 5.18 1.95
'Januar ~
9, 1990 Testin 2-MRV-210 5.07 "2. 83 2-MRV-220 5.60 2.28 2-MRV-240 5.05 2.21 Januar 10, 1990 Testin 2-MRV-210 5. 78 3.08 2-MRV-220 5. 73 2.55
'anuar 11, 1990 Testing 2-MJRV-220 5.26 No Retest Performed NOTE: During the test program 2-HRV-230 closing times were high, hut 'did not exceed the Technical Specification Limit of five Seconds.
NR C F oIIm 344A 144)9)
NRC FORM 364A US. NUCLEAR AEGULATOAYCOMMISSION APPAOVED OMS NO, 31500)04 1640)
E XP I R E 5: 4/30/02 ESTIMATED SUADEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGAROINO SUADEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT SRANCH IP430). U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. OC 20655, AND TO 1HE PAPERWORK REDUCTION PRO/ECT 13)50410E).'OFFICE
~ OF MANAGEMENTAND SU OG ET, WASHINGTON, DC 10503.
FACILITY NAME 111 Doer ET NUMSER (2) LEA NUMSER (4) PAGE IS)
YEAR SEQUENTIAL HEY>>ION NVM 4 II )rvM ER D..C. Cook Plant - Unit 2, o s o o o 3 1 6 9 0 0 0 2 0 0'0 9 QF 1 0 TEXT ///nxro s/>>ce /s r)V/rer/ 1>>e ~ PT/Or>>/ HIIC Ferrrr AXLES/ (17) t MAIN STEAM ISOLATION VALVE NOTE: VALVE SHOWN IN CLOSED POSITION OPERATING PISTON
~VENT LINE TO STEAN DUMP VALVES STEAM NIPPLE CONDENSATE DRAIN LINE CONDENSATE DRAIN PORT VALVE MODIFICATIONS CONDENSATE DRAIN PORT INITIALLY- 0 125" MODIFIED TO -.0..'197" (5MM)
STEAM NIPPLE INITIALLY" 0'125" MODIFI'ED TO 0.394" (10MM)
~ ~ 4
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NRC Form 346A (680)
NRC FORM 385A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMS NO. 315040104 (54ISI EXPIRES; 4I30JS3 ESTIMATED SURQEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLL'ECTION REQUEST: 50.0 MRS. FQRWARQ COMMENTS REOAROINO bUROEH ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAOEMEHT SRANCH If@30). U.S. NUCLEAR REOULATORY COMMISSION. WASHINOTOH, DC 30555, AHQ TQ f
1HE APFRWORK REDUCTION PROJECT 131504)104), Qf ICE f OF MANAGEMENTAHD SUDOET,WASHINGTON, DC 30503, fACILITYNAME Il) DOCKET NUMSER 13) LER NUMSER Ibl PACE ISI YEAR SEOVENTIAL II5 V IS IO N NVMSEII NVM ~ II D. C. Cook Plant - Unit 2 o s o o o 3 l 6 9 0 0 0 2 0 0 1 0 OF ] 0 TEXT IllII>>IF Z>>ce JF IP0vbe4 uw FIRRSN>>F JYRC %%dYm ~3 I l)T)
HAIN STEAM ISOLATION VALVE VENT PIPING VENTS TO ATMOSPHERE DUMP VALVES 3-WAY MOTOR OPERATED VALVE MSIV HRC F~ SeSA ISJISI