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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17326A2011999-08-23023 August 1999 LER 99-004-00:on 990724,reactor Tripped Due to Main Transformer Bushing Flashover.Plant Was Brought to SS & Components Were Tested & Performed Satisfactorily.With 990823 Ltr ML20029C7321994-04-22022 April 1994 LER 94-004-00:on 940221,discovered Corrosion of Three Nuts on One of Incor Instrumentation Reactor Vessel Head.Caused by Increase of Wet Boric Acid.Leaking Flanges Repaired.W/ 940422 Ltr ML20046B4731993-07-30030 July 1993 LER 93-005-00:on 930630,TS 3.0.3 Entered Due to Both Containment Spray Sys Inoperable.Replaced CCW Outlet Valve Actuator Connecting Link Assembly from Number 11 SDC Heat exchanger.W/930730 Ltr ML20046A4911993-07-22022 July 1993 LER 93-003-00:on 930625,SG Tripped Due to Low Water Level. Caused by Insufficient Feedwater Addition Due to Inadequate Communication.Reemphasis on Improved Communication Stressed. W/930722 Ltr ML20045G8611993-07-0909 July 1993 LER 93-003-00:on 930610,dual Unit Trip Occurred Due to Partial Loss of Offsite Power.Flashover Protection Relay for Breaker 552-61 Replaced & Training for Personnel W/ Access to Relays Will Be reinforced.W/930709 Ltr ML20045G7281993-07-0808 July 1993 LER 93-002-00:on 930608,inadvertent Arw Actuation Sys & RPS Actuations Experienced During Performance of Awf Sys Large Flow Surveillance Testing.Caused by Failure to Note Differential Pressure Condition.Valve opened.W/930708 Ltr ML20045G8661993-07-0808 July 1993 LER 93-004-00:on 930611,reactor Tripped Due to Turbine Trip Resulting from Inadequate Procedure.Procedure Changes Made to Open Appropriate FW Heater High Level Dump Valves During Plant startup.W/930708 Ltr ML20045E7361993-06-29029 June 1993 LER 93-002-01:on 930205,software Vendor Discovered Error in User Manual for Updating Basss data-input Library.Caused by Failure of QA Procedures to Require Independent Review of User Manuals.Manual Surveillances performed.W/930629 Ltr ML20029B1261991-02-28028 February 1991 LER 91-001-00:on 910129,tubing in Air Start Sys for Emergency Diesel Generator Failed During Seismic Event. Caused by Error in Design of EDG Air Start Sys.Permanent Mod to Sys installed.W/910228 Ltr ML20028H8011991-01-24024 January 1991 LER 90-002-01:on 900116,determined That 891211 Reconstitution of More than One Spent Fuel Assembly Per Time in Violation of Fuel Handling Incident Safety Analysis. Caused by Deficient procedure.W/910124 Ltr ML20044A1861990-06-20020 June 1990 LER 87-002-01:on 861203,section of Thin Wall Found on Main Steam Line W/Readings Below Allowable Min of 0.95 Inches. Caused by Grinding of Edge of Pipe to Achieve Proper fit-up for Welding.Relief from IWB-3610 granted.W/900620 Ltr ML20043G1071990-06-13013 June 1990 LER 89-019-01:on 891128,determined That for Approx 10 Yrs, from 1979-1989,requirement to Lock HPSI Discharge Header Isolation Valves Shut Not Implemented.Caused by Inadequate Mgt Attention.Test Procedures modified.W/900613 Ltr ML20043F1221990-06-0404 June 1990 LER 90-017-00:on 900505,pin Hole Leak Observed in Discharge Piping of Saltwater Pump 13.Caused by Localized Corrosion. Leaking Spool Piece Removed & Blank Flange Installed. W/900604 Ltr ML20043A7871990-05-21021 May 1990 LER 90-016-00:on 900421,determined That Waste Gas Decay Tank (Wgdt) 13 Discharged Instead of (Wgdt) 11 for Discharge Permit Issued.Caused by Inadequate Communications.Training Performed for Operators Re event.W/900521 Ltr ML20043A3441990-05-14014 May 1990 LER 90-014-00:on 900413 & 19,unit Entered Tech Spec Limiting Condition of Operation 3.0.3 Due to Potential Inoperability of Three Out of Four Reactor Protection Sys Delta T Power Channels.Caused by Lack of Procedure guidance.W/900514 Ltr ML20043A3401990-05-14014 May 1990 LER 90-013-00:on 900413,determined That Axial Shape Index Channels Out of Spec & Inoperable.Caused by Inadequate Understanding of Design Basis for Excore/Incore Comparison. Design Basis for Excore/Incore improved.W/900514 Ltr ML20042G4521990-05-0707 May 1990 LER 90-015-00:on 900407,discovered That Relay Contact Which Actuates Reactor Trip Breaker Shunt Trip Not Adequately Functionally Tested.Caused by Failure to Examine Circuit in Detail When Test developed.W/900507 Ltr ML20042F5801990-05-0404 May 1990 LER 90-012-00:on 900406,identified That Procedure for LOCA Would Not Ensure post-LOCA Core Flush Would Be Initiated in Time to Prevent Boron Precipitation.Caused by Personnel Error.Configuration Mgt Program strengthened.W/900504 Ltr ML20012E9931990-03-29029 March 1990 LER 90-008-00:on 900227,determined That Surveillance Procedure M-280-0 Did Not Include Steps to Fully Test Control Room Recorder for Hydrogen Analyzers.Caused by Personnel Error.Procedure Revised on 900308.W/900329 Ltr ML20012F0001990-03-28028 March 1990 LER 89-006-01:on 890508,containment Iodine Filters Outside Design Basis Due to Equipment Qualification.Recalculation of Total Integrated Radiation Dose to Cables for Filter Fans Demonstrated Cable qualified.W/900328 Ltr ML20012E9951990-03-28028 March 1990 LER 89-014-01:on 890723,determined That Salt Water Header Not Capable of Withstanding Seismic Event Intact.Caused by Inadequate Welding of Blind Spool Pieces in Pipe.Insp Revealed Spools Capable as installed.W/900328 Ltr ML20012E0101990-03-26026 March 1990 LER 90-009-00:on 900224,failure to Meet Action Requirement Re Tech Spec 3.7.12.Caused by Personnel Error.Cables Removed from Doorway in Charging Pump Room & Not Allowed to Be Placed in doorway.W/900326 Ltr ML20012C4971990-03-15015 March 1990 LER 90-007-00:on 900216,discovered That Supervised Circuits Associated W/Fire Detection Instruments Located in Reactor Coolant Pump Bays Not Been Included in Surveillance Test Procedure.Caused by Personnel error.W/900315 Ltr ML20012C4861990-03-12012 March 1990 LER 90-006-00:on 900209,determined That Four Fire Dampers Missing.Caused by Not Identifying Penetrations as Requiring Dampers When Fire Hazards Analysis of Plant Conducted.Hourly Fire Watch Continued.Missing Dampers installed.W/900312 Ltr ML20012B8991990-03-12012 March 1990 LER 89-023-01:on 891220,determined That Pipe Rupture in nonsafety-related Svc Water Subsystem Could Result in Rapid Draining of Subsystems That Serve Auxiliary Bldg.Task Force Formed to Determine Corrective actions.W/900312 Ltr ML20012B4221990-03-0606 March 1990 LER 89-012-01:on 891227,core Alterations Performed W/Only One of Two Containment Vent Valves Closed.Caused by Procedural Deficiency.Procedures Revised to Include Valve & Surveillance Test Program Instruction revised.W/900306 Ltr ML20011F2701990-02-27027 February 1990 LER 90-001-01:on 900109,determined That Surveillance Tests Used to Perform Channel Calibr Tests for Acoustic Flow Monitoring Devices Inadequate.Caused by Personnel Error & Inadequate Procedures.Swapped Leads restored.W/900227 Ltr ML20011F2091990-02-27027 February 1990 LER 89-026-00:on 891128,determined That Particulate Levels in Samples Taken from Lower Third of Tanks Exceeded Allowable Limits.Caused by Inadequate Sampling Technique. Tanks Cleaned & Filled W/Clean fuel.W/900227 Ltr ML20006F8601990-02-22022 February 1990 LER 90-004-00:on 900123,discovered Fire Barrier Penetration Seal Open for Indeterminate Time W/O Performing Tech Spec 3.7.12.a Required Actions.Caused by Personnel Error. Temporary Fire Seal installed.W/900222 Ltr ML20006E0521990-02-0808 February 1990 LER 90-001-00:on 891221,discovered That Acoustic Indications for One PORV & One Safety Valve Were Reversed During Surveillance Test.Caused by Personnel Error.Swapped Leads Restored to Proper configuration.W/900208 Ltr ML20006B4801990-01-26026 January 1990 LER 89-022-00:on 891227,core Alterations Performed W/Only One of Two Containment Vent Valves Closed,Violating Tech Specs.Caused by Procedural Deficiency.Surveillance Test Procedure Revised to Include Deleted valves.W/900126 Ltr ML19354D8931990-01-17017 January 1990 LER 89-024-00:on 891218,determined That Wires Which Connect Actuation Device Logic Relay Contacts to Remainder of Circuit Not Tested During Channel Calibr Test.Caused by Inadequate Test.Test Program Upgrade underway.W/900117 Ltr ML20005F1921990-01-10010 January 1990 LER 89-025-00:on 891208,Tech Spec Action Statement Entered When Ventilation Ducts Penetrating Fire Barrier Could Not Be Accessed to Determine If Fire Dampers Installed.On 891211, Fire Watch Missed.Caused by Personnel error.W/900110 Ltr ML20005E3971989-12-28028 December 1989 LER 89-019-00:on 891128,discovered That HPSI Discharge Header Isolation Valves Not Locked Shut When RCS in Water Solid Condition,Resulting in Operation Outside Design Basis. Procedure Revised to Require Valves closed.W/891228 Ltr ML19351A4551989-12-13013 December 1989 LER 89-020-00:on 891113,determined That Some Solenoid Valves & Valve Power Supplies for Saltwater Sys May Not Be Able to Perform Design Function After Design Basis Seismic Event. Cause Undetermined.Power Supplies upgraded.W/891213 Ltr ML20005D6611989-12-0606 December 1989 LER 89-018-00:on 891106,discovered That Many Air Operated Control Valves & piston-operated Dampers Which Utilize safety-related Air Accumulators Would Not Have Performed as Expected After Loss of air.W/891206 Ltr ML19325F3951989-11-10010 November 1989 LER 89-002-01:on 890228,discovered That Fire Barrier Penetration Inoperable & Action Statement Requirements Not Satisfied.Caused by Inadequate Administrative Controls. Penetration Returned to Operable status.W/891115 Ltr ML19325E8221989-11-0303 November 1989 LER 89-007-01:on 890505,evidence of Reactor Coolant Leakage from 120 Pressurizer Vessel Heater Penetrations Discovered. Caused by IGSCC of Inconel 600.All Penetrations Using J-welds & Inconel 600 Visually inspected.W/891103 Ltr ML19324B2511989-10-27027 October 1989 LER 89-012-01:on 890720,discovered That Master Solenoid to Switchgear Room Halon Sys Disconnected Since 890629.Caused by Personnel Error Resulting from Lack of Written Procedure. Procedure Revised to Apply Temporary mods.W/891027 Ltr ML19325C3281989-10-10010 October 1989 LER 89-016-00:on 890908,determined That as-found Condition of Resistance Temp Detectors Did Not Match Tested Configuration.Cause Not Stated.Subj Detectors Will Be Sealed,Per Environ Qualification requirements.W/891010 Ltr ML19325C3701989-10-0909 October 1989 LER 89-017-00:on 890907,determined That Discrepancy in Acceptance Criteria of Surveillance Test Procedure M-452-0 Resulted in Failure to Fully Comply W/Requirements of Tech Spec 3.9.12.Main Cause undetermined.W/891009 Ltr ML20024F3771983-08-25025 August 1983 LER 83-044/03L-0:on 830808,diesel Generator 12 Tripped on Low Jacket Cooling Water Pressure While Verifying Operability.Cause Not Stated.Coolant Jacket Vented & Large Amount of Air Found.No Evidence of leakage.W/830825 Ltr ML20024F5731983-08-25025 August 1983 LER 83-040/03L-0:on 830727,control Room Air Conditioner 11 Discovered W/Damaged Condenser Fan.Caused by Loose Set Screws Securing Fan in Position.Set Screws Restored. W/830825 Ltr ML20024E6761983-08-0404 August 1983 Updated LER 83-011/03X-1:on 830207,during Surveillance Testing ESFAS a Logic Sequencer Failed,Rendering Diesel Generator 12 Inoperable.Caused by Intermittent Operation of Module Test Push Button.Part replaced.W/830804 Ltr ML20024E1721983-07-14014 July 1983 Updated LER 81-015/03X-1:on 810226,sample Pump for Control Room Radiation Monitor Found Out of Svc,Rendering Automatic Recirculation of Control Room Ventilation Sys on High Radiation Inoperable.Caused by seizure.W/830714 Ltr ML20024D0071983-07-0808 July 1983 LER 83-035/03L-0:on 830610,during Normal Power Operation,Esf Actuation Sys Channel Zg Steam Generator Level Tripped. Caused by Failed Vitro Isolator Module.Module Replaced.W/ 830708 Ltr ML20024D0091983-07-0808 July 1983 LER 83-033/03L-0:on 830603,fire Detection Instrumentation in Containment Southeast Electrical Penetration Determined Inoperable.Repair Impossible Due to Inaccessability of Protecto wire.W/830708 Ltr ML20024B8231983-06-23023 June 1983 LER 83-029/03L-0:on 830524,during Normal Operation, Surveillance Testing Indicated Neither Spent Fuel Pool Exhaust Fans 11 or 12 Would Maintain Required Negative Pressure.Caused by Clogged HEPA filters.W/830623 Ltr ML20024C0141983-06-22022 June 1983 Updated LER 81-080/03X-1:on 811116,discovered Weep from Cracked Weld on Spent Fuel Cooling Pump Discharge Vent Line 12.Caused by Inadequate Support of Vent Line.Support Assembly installed.W/830622 Ltr ML20024A8881983-06-16016 June 1983 LER 83-032/03L-0:on 830523,containment Isolation Sys B Logic Module Would Not Actuate.Caused by Defective Vitro Labs Std Logic Module.Module Replaced.Failed Module Returned to Vitro Labs for Repair & testing.W/830616 Ltr 1999-08-23
[Table view] Category:RO)
MONTHYEARML17326A2011999-08-23023 August 1999 LER 99-004-00:on 990724,reactor Tripped Due to Main Transformer Bushing Flashover.Plant Was Brought to SS & Components Were Tested & Performed Satisfactorily.With 990823 Ltr ML20029C7321994-04-22022 April 1994 LER 94-004-00:on 940221,discovered Corrosion of Three Nuts on One of Incor Instrumentation Reactor Vessel Head.Caused by Increase of Wet Boric Acid.Leaking Flanges Repaired.W/ 940422 Ltr ML20046B4731993-07-30030 July 1993 LER 93-005-00:on 930630,TS 3.0.3 Entered Due to Both Containment Spray Sys Inoperable.Replaced CCW Outlet Valve Actuator Connecting Link Assembly from Number 11 SDC Heat exchanger.W/930730 Ltr ML20046A4911993-07-22022 July 1993 LER 93-003-00:on 930625,SG Tripped Due to Low Water Level. Caused by Insufficient Feedwater Addition Due to Inadequate Communication.Reemphasis on Improved Communication Stressed. W/930722 Ltr ML20045G8611993-07-0909 July 1993 LER 93-003-00:on 930610,dual Unit Trip Occurred Due to Partial Loss of Offsite Power.Flashover Protection Relay for Breaker 552-61 Replaced & Training for Personnel W/ Access to Relays Will Be reinforced.W/930709 Ltr ML20045G7281993-07-0808 July 1993 LER 93-002-00:on 930608,inadvertent Arw Actuation Sys & RPS Actuations Experienced During Performance of Awf Sys Large Flow Surveillance Testing.Caused by Failure to Note Differential Pressure Condition.Valve opened.W/930708 Ltr ML20045G8661993-07-0808 July 1993 LER 93-004-00:on 930611,reactor Tripped Due to Turbine Trip Resulting from Inadequate Procedure.Procedure Changes Made to Open Appropriate FW Heater High Level Dump Valves During Plant startup.W/930708 Ltr ML20045E7361993-06-29029 June 1993 LER 93-002-01:on 930205,software Vendor Discovered Error in User Manual for Updating Basss data-input Library.Caused by Failure of QA Procedures to Require Independent Review of User Manuals.Manual Surveillances performed.W/930629 Ltr ML20029B1261991-02-28028 February 1991 LER 91-001-00:on 910129,tubing in Air Start Sys for Emergency Diesel Generator Failed During Seismic Event. Caused by Error in Design of EDG Air Start Sys.Permanent Mod to Sys installed.W/910228 Ltr ML20028H8011991-01-24024 January 1991 LER 90-002-01:on 900116,determined That 891211 Reconstitution of More than One Spent Fuel Assembly Per Time in Violation of Fuel Handling Incident Safety Analysis. Caused by Deficient procedure.W/910124 Ltr ML20044A1861990-06-20020 June 1990 LER 87-002-01:on 861203,section of Thin Wall Found on Main Steam Line W/Readings Below Allowable Min of 0.95 Inches. Caused by Grinding of Edge of Pipe to Achieve Proper fit-up for Welding.Relief from IWB-3610 granted.W/900620 Ltr ML20043G1071990-06-13013 June 1990 LER 89-019-01:on 891128,determined That for Approx 10 Yrs, from 1979-1989,requirement to Lock HPSI Discharge Header Isolation Valves Shut Not Implemented.Caused by Inadequate Mgt Attention.Test Procedures modified.W/900613 Ltr ML20043F1221990-06-0404 June 1990 LER 90-017-00:on 900505,pin Hole Leak Observed in Discharge Piping of Saltwater Pump 13.Caused by Localized Corrosion. Leaking Spool Piece Removed & Blank Flange Installed. W/900604 Ltr ML20043A7871990-05-21021 May 1990 LER 90-016-00:on 900421,determined That Waste Gas Decay Tank (Wgdt) 13 Discharged Instead of (Wgdt) 11 for Discharge Permit Issued.Caused by Inadequate Communications.Training Performed for Operators Re event.W/900521 Ltr ML20043A3441990-05-14014 May 1990 LER 90-014-00:on 900413 & 19,unit Entered Tech Spec Limiting Condition of Operation 3.0.3 Due to Potential Inoperability of Three Out of Four Reactor Protection Sys Delta T Power Channels.Caused by Lack of Procedure guidance.W/900514 Ltr ML20043A3401990-05-14014 May 1990 LER 90-013-00:on 900413,determined That Axial Shape Index Channels Out of Spec & Inoperable.Caused by Inadequate Understanding of Design Basis for Excore/Incore Comparison. Design Basis for Excore/Incore improved.W/900514 Ltr ML20042G4521990-05-0707 May 1990 LER 90-015-00:on 900407,discovered That Relay Contact Which Actuates Reactor Trip Breaker Shunt Trip Not Adequately Functionally Tested.Caused by Failure to Examine Circuit in Detail When Test developed.W/900507 Ltr ML20042F5801990-05-0404 May 1990 LER 90-012-00:on 900406,identified That Procedure for LOCA Would Not Ensure post-LOCA Core Flush Would Be Initiated in Time to Prevent Boron Precipitation.Caused by Personnel Error.Configuration Mgt Program strengthened.W/900504 Ltr ML20012E9931990-03-29029 March 1990 LER 90-008-00:on 900227,determined That Surveillance Procedure M-280-0 Did Not Include Steps to Fully Test Control Room Recorder for Hydrogen Analyzers.Caused by Personnel Error.Procedure Revised on 900308.W/900329 Ltr ML20012F0001990-03-28028 March 1990 LER 89-006-01:on 890508,containment Iodine Filters Outside Design Basis Due to Equipment Qualification.Recalculation of Total Integrated Radiation Dose to Cables for Filter Fans Demonstrated Cable qualified.W/900328 Ltr ML20012E9951990-03-28028 March 1990 LER 89-014-01:on 890723,determined That Salt Water Header Not Capable of Withstanding Seismic Event Intact.Caused by Inadequate Welding of Blind Spool Pieces in Pipe.Insp Revealed Spools Capable as installed.W/900328 Ltr ML20012E0101990-03-26026 March 1990 LER 90-009-00:on 900224,failure to Meet Action Requirement Re Tech Spec 3.7.12.Caused by Personnel Error.Cables Removed from Doorway in Charging Pump Room & Not Allowed to Be Placed in doorway.W/900326 Ltr ML20012C4971990-03-15015 March 1990 LER 90-007-00:on 900216,discovered That Supervised Circuits Associated W/Fire Detection Instruments Located in Reactor Coolant Pump Bays Not Been Included in Surveillance Test Procedure.Caused by Personnel error.W/900315 Ltr ML20012C4861990-03-12012 March 1990 LER 90-006-00:on 900209,determined That Four Fire Dampers Missing.Caused by Not Identifying Penetrations as Requiring Dampers When Fire Hazards Analysis of Plant Conducted.Hourly Fire Watch Continued.Missing Dampers installed.W/900312 Ltr ML20012B8991990-03-12012 March 1990 LER 89-023-01:on 891220,determined That Pipe Rupture in nonsafety-related Svc Water Subsystem Could Result in Rapid Draining of Subsystems That Serve Auxiliary Bldg.Task Force Formed to Determine Corrective actions.W/900312 Ltr ML20012B4221990-03-0606 March 1990 LER 89-012-01:on 891227,core Alterations Performed W/Only One of Two Containment Vent Valves Closed.Caused by Procedural Deficiency.Procedures Revised to Include Valve & Surveillance Test Program Instruction revised.W/900306 Ltr ML20011F2701990-02-27027 February 1990 LER 90-001-01:on 900109,determined That Surveillance Tests Used to Perform Channel Calibr Tests for Acoustic Flow Monitoring Devices Inadequate.Caused by Personnel Error & Inadequate Procedures.Swapped Leads restored.W/900227 Ltr ML20011F2091990-02-27027 February 1990 LER 89-026-00:on 891128,determined That Particulate Levels in Samples Taken from Lower Third of Tanks Exceeded Allowable Limits.Caused by Inadequate Sampling Technique. Tanks Cleaned & Filled W/Clean fuel.W/900227 Ltr ML20006F8601990-02-22022 February 1990 LER 90-004-00:on 900123,discovered Fire Barrier Penetration Seal Open for Indeterminate Time W/O Performing Tech Spec 3.7.12.a Required Actions.Caused by Personnel Error. Temporary Fire Seal installed.W/900222 Ltr ML20006E0521990-02-0808 February 1990 LER 90-001-00:on 891221,discovered That Acoustic Indications for One PORV & One Safety Valve Were Reversed During Surveillance Test.Caused by Personnel Error.Swapped Leads Restored to Proper configuration.W/900208 Ltr ML20006B4801990-01-26026 January 1990 LER 89-022-00:on 891227,core Alterations Performed W/Only One of Two Containment Vent Valves Closed,Violating Tech Specs.Caused by Procedural Deficiency.Surveillance Test Procedure Revised to Include Deleted valves.W/900126 Ltr ML19354D8931990-01-17017 January 1990 LER 89-024-00:on 891218,determined That Wires Which Connect Actuation Device Logic Relay Contacts to Remainder of Circuit Not Tested During Channel Calibr Test.Caused by Inadequate Test.Test Program Upgrade underway.W/900117 Ltr ML20005F1921990-01-10010 January 1990 LER 89-025-00:on 891208,Tech Spec Action Statement Entered When Ventilation Ducts Penetrating Fire Barrier Could Not Be Accessed to Determine If Fire Dampers Installed.On 891211, Fire Watch Missed.Caused by Personnel error.W/900110 Ltr ML20005E3971989-12-28028 December 1989 LER 89-019-00:on 891128,discovered That HPSI Discharge Header Isolation Valves Not Locked Shut When RCS in Water Solid Condition,Resulting in Operation Outside Design Basis. Procedure Revised to Require Valves closed.W/891228 Ltr ML19351A4551989-12-13013 December 1989 LER 89-020-00:on 891113,determined That Some Solenoid Valves & Valve Power Supplies for Saltwater Sys May Not Be Able to Perform Design Function After Design Basis Seismic Event. Cause Undetermined.Power Supplies upgraded.W/891213 Ltr ML20005D6611989-12-0606 December 1989 LER 89-018-00:on 891106,discovered That Many Air Operated Control Valves & piston-operated Dampers Which Utilize safety-related Air Accumulators Would Not Have Performed as Expected After Loss of air.W/891206 Ltr ML19325F3951989-11-10010 November 1989 LER 89-002-01:on 890228,discovered That Fire Barrier Penetration Inoperable & Action Statement Requirements Not Satisfied.Caused by Inadequate Administrative Controls. Penetration Returned to Operable status.W/891115 Ltr ML19325E8221989-11-0303 November 1989 LER 89-007-01:on 890505,evidence of Reactor Coolant Leakage from 120 Pressurizer Vessel Heater Penetrations Discovered. Caused by IGSCC of Inconel 600.All Penetrations Using J-welds & Inconel 600 Visually inspected.W/891103 Ltr ML19324B2511989-10-27027 October 1989 LER 89-012-01:on 890720,discovered That Master Solenoid to Switchgear Room Halon Sys Disconnected Since 890629.Caused by Personnel Error Resulting from Lack of Written Procedure. Procedure Revised to Apply Temporary mods.W/891027 Ltr ML19325C3281989-10-10010 October 1989 LER 89-016-00:on 890908,determined That as-found Condition of Resistance Temp Detectors Did Not Match Tested Configuration.Cause Not Stated.Subj Detectors Will Be Sealed,Per Environ Qualification requirements.W/891010 Ltr ML19325C3701989-10-0909 October 1989 LER 89-017-00:on 890907,determined That Discrepancy in Acceptance Criteria of Surveillance Test Procedure M-452-0 Resulted in Failure to Fully Comply W/Requirements of Tech Spec 3.9.12.Main Cause undetermined.W/891009 Ltr ML20024F3771983-08-25025 August 1983 LER 83-044/03L-0:on 830808,diesel Generator 12 Tripped on Low Jacket Cooling Water Pressure While Verifying Operability.Cause Not Stated.Coolant Jacket Vented & Large Amount of Air Found.No Evidence of leakage.W/830825 Ltr ML20024F5731983-08-25025 August 1983 LER 83-040/03L-0:on 830727,control Room Air Conditioner 11 Discovered W/Damaged Condenser Fan.Caused by Loose Set Screws Securing Fan in Position.Set Screws Restored. W/830825 Ltr ML20024E6761983-08-0404 August 1983 Updated LER 83-011/03X-1:on 830207,during Surveillance Testing ESFAS a Logic Sequencer Failed,Rendering Diesel Generator 12 Inoperable.Caused by Intermittent Operation of Module Test Push Button.Part replaced.W/830804 Ltr ML20024E1721983-07-14014 July 1983 Updated LER 81-015/03X-1:on 810226,sample Pump for Control Room Radiation Monitor Found Out of Svc,Rendering Automatic Recirculation of Control Room Ventilation Sys on High Radiation Inoperable.Caused by seizure.W/830714 Ltr ML20024D0071983-07-0808 July 1983 LER 83-035/03L-0:on 830610,during Normal Power Operation,Esf Actuation Sys Channel Zg Steam Generator Level Tripped. Caused by Failed Vitro Isolator Module.Module Replaced.W/ 830708 Ltr ML20024D0091983-07-0808 July 1983 LER 83-033/03L-0:on 830603,fire Detection Instrumentation in Containment Southeast Electrical Penetration Determined Inoperable.Repair Impossible Due to Inaccessability of Protecto wire.W/830708 Ltr ML20024B8231983-06-23023 June 1983 LER 83-029/03L-0:on 830524,during Normal Operation, Surveillance Testing Indicated Neither Spent Fuel Pool Exhaust Fans 11 or 12 Would Maintain Required Negative Pressure.Caused by Clogged HEPA filters.W/830623 Ltr ML20024C0141983-06-22022 June 1983 Updated LER 81-080/03X-1:on 811116,discovered Weep from Cracked Weld on Spent Fuel Cooling Pump Discharge Vent Line 12.Caused by Inadequate Support of Vent Line.Support Assembly installed.W/830622 Ltr ML20024A8881983-06-16016 June 1983 LER 83-032/03L-0:on 830523,containment Isolation Sys B Logic Module Would Not Actuate.Caused by Defective Vitro Labs Std Logic Module.Module Replaced.Failed Module Returned to Vitro Labs for Repair & testing.W/830616 Ltr 1999-08-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G6971999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Calvert Cliffs Npp,Units 1 & 2.With ML20216J8731999-09-10010 September 1999 Rev 52 to QA Policy for Calvert Cliffs Nuclear Power Plant ML20212A4441999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ccnpp,Units 1 & 2. with ML17326A2011999-08-23023 August 1999 LER 99-004-00:on 990724,reactor Tripped Due to Main Transformer Bushing Flashover.Plant Was Brought to SS & Components Were Tested & Performed Satisfactorily.With 990823 Ltr ML20210S6091999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ccnpp,Units 1 & 2. with ML20210N6001999-07-27027 July 1999 ISI Summary Rept for Calvert Cliffs Unit 2. Page 2 of 3 in Encl 1 of Incoming Submittal Not Included ML20210B7941999-07-15015 July 1999 SER Denying Licensee Request for Changes to Current Ts,Re Deletion of Tendon Surveillance Requirements for Calvert Cliffs LD-99-039, Part 21 Rept Re Defect of Abb 1200A 4kV Vacuum Breakers. Initially Reported on 990625.Defect Results in Breaker Failing to Remain in Closed Position.Root Cause Evaluation & Corrective Action Plan Being Developed.Licensee Notified1999-06-30030 June 1999 Part 21 Rept Re Defect of Abb 1200A 4kV Vacuum Breakers. Initially Reported on 990625.Defect Results in Breaker Failing to Remain in Closed Position.Root Cause Evaluation & Corrective Action Plan Being Developed.Licensee Notified ML20209F1721999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Calvert Cliffs Npp.With LD-99-035, Part 21 Rept Re Abb 1200A 4KV Vacuum Breakers Performing Trip Free Operation When Close Signal Received by Breaker. Defect Results in Breaker Failing to Remain in Closed Position.Root Cause & CAP Being Developed1999-06-25025 June 1999 Part 21 Rept Re Abb 1200A 4KV Vacuum Breakers Performing Trip Free Operation When Close Signal Received by Breaker. Defect Results in Breaker Failing to Remain in Closed Position.Root Cause & CAP Being Developed ML20196C6981999-06-21021 June 1999 Safety Evaluation Concluding That Use of ASME Section XI Code Including Summer 1983 Addenda as Interim Code for Third 10-year Insp Interval at Calvert Cliffs Units 1 & 2 Until Review of 1998 Code Completed,Would Be Acceptable ML20195K2811999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Ccnpp,Units 1 & 2. with ML20206R5871999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ccnpp,Units 1 & 2. with ML20195B3891999-04-30030 April 1999 0 to CENPD-279, Annual Rept on Abb CE ECCS Performance Evaluation Models ML20205N2951999-04-13013 April 1999 Special Rept:On 990314,fire Detection Sys Was Removed from Svc to Support Mod to Replace SRW Heat Exchangers in Unit 2 SRW Room During Unit 2 Refueling Outage.Contingency Measure 15.3.5.A.1 Will Continue Until Fire Detection Sys Restored ML20210T5211999-04-0101 April 1999 Rev 0 to Ccnpp COLR for Unit 2,Cycle 13 ML20205P5441999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2.With ML20204H6471999-03-21021 March 1999 SER Re License Renewal of Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20207M8321999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Calvert Cliffs Nuclear Power Plant.With ML20203D4311999-02-0505 February 1999 Safety Evaluation Accepting Procedure Established for long-term Corrective Action Plan Related to Containment Vertical Tendons ML20199G4671999-01-20020 January 1999 SER Accepting USI A-46 Implementation for Plant ML20206Q3221999-01-11011 January 1999 Special Rept:On 981226,wide Range Noble Gas Effluent RM Was Removed from Operable Status.Caused by Failure of mid-range Checksource to Properly Reseat.Completed Maint & post-maint Testing & RM Was Returned to Operable Status on 990104 ML20207L0451999-01-0808 January 1999 Cost-Benefit Risk Analyses:Radwaste Sys for Light Water Reactors ML20199F4781999-01-0808 January 1999 Safety Evaluation Concluding That Bg&E Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves Susceptible to Pressure Locking.Concludes GL 95-07 Actions Were Addressed ML20198S7591999-01-0707 January 1999 SER Accepting Quality Assurance Program Description Change for Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20207M2281998-12-31031 December 1998 1998 Annual Rept for Bg&E. with ML20199E2931998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Calvert Cliffs Npp. with ML20206R9911998-12-0808 December 1998 Rept of Changes,Tests & Experiments (10CFR50.59(b)(2)). with ML20198B2631998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2.With ML20195H1001998-11-16016 November 1998 Safety Evaluation of First Containment Insp Interval Iwe/Iwl Program Alternative ML20196E2211998-10-31031 October 1998 Non-proprietary Rev 03-NP to CEN-633-NP, SG Tube Repair for Combustion Engineering Designed Plant with 3/4 - .048 Wall Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves ML20195E5281998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Calvert Cliffs Nuclear Power Station,Units 1 & 2.With ML20154Q7191998-10-21021 October 1998 Special Rept:On 980923,unit 1 Wrngm Was Removed from Operable Status.Caused by Failure of Process Flow Transducer.Completed Maint to Remove Process Flow Transducer Input to Wrngm Microprocessor & Completed Formal Evaluation ML20154G3931998-10-0505 October 1998 Safety Evaluation Concluding That Flaw Tolerance Evaluation for Assumed Flaw in Inboard Instrument Weld of Pressurizer Meets Rules of ASME Code ML20154M5841998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Calvert Cliffs Nuclear Plant,Units 1 & 2.With ML20153C2571998-09-18018 September 1998 Special Rept:On 980830,wide Range Noble Gas Monitor (Wrngm) Channel Was Removed from Operable Status.Caused by Need to Support Performance of Required 18-month Channel Calibr.Will Return Wrngm to Operable Status by 980925 ML20153C1091998-09-18018 September 1998 Part 21 Rept Re Defective Capacity Control Valves.Trentec Personnel Have Been in Contact with Bg&E Personnel Re Condition & Have Requested Potentially Defective Valves ML20151U5441998-09-0404 September 1998 Bg&E ISI Summary Rept for Calvert Cliffs ML20151T5281998-09-0101 September 1998 Special Rept:On 980819,declared Rv Water Level Monitor Channel a Inoperable.Caused by Failure of Three Heated Junction Thermocouples (Sensors) in Lower Five Sensors. Channel a & B Rv Water Level Probes Will Be Replaced ML20151Y1191998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Calvert Cliffs Nuclear Power Plant Units 1 & 2.With ML20237D4981998-08-19019 August 1998 Safety Evaluation Accepting Licensee Request for Extension of Second ten-year Inservice Insp Interval ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20237B9371998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Calvert Cliffs Nuclear Power Plant ML20237D5941998-07-22022 July 1998 Rev 2 to Ccnpp COLR for Unit 2,Cycle 12 ML20236L7521998-07-0606 July 1998 Safety Evaluation Granting Bg&E 980527 Request for Relief from Requirement of Section IWA-5250 of ASME Code for Calvert Cliffs Unit 2.Alternatives Provide Reasonable Assurance of Operational Readiness ML20236F7791998-06-30030 June 1998 Safety Evaluation Authorizing Request for Temporary Relief from Requirement of Subsection IWA-5250 of ASME Code,Section XI for Plant,Unit 1 ML20236R0881998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20236X3101998-06-19019 June 1998 Rev 1 to Calvert Cliffs Nuclear Power Plant COLR for Unit 2,Cycle 12 ML20249A9571998-06-15015 June 1998 Special Rept:On 980430,fire Detection Sys Was Removed from Svc to Support Mod to Purge Air Sys 27-foot Elevation & 5-foot Elevation East Piping Penetration Rooms.Installed Temporary Alteration & Returned Fire Detection Sys to Svc ML20249A7711998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Ccnpp,Units 1 & 2 1999-09-30
[Table view] |
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., BALTIMORE Ru' ' '
OAS AND ;
i5 4 ELECTRIC -4 b CHARLES CENTER
- BALTIMORE. MARYLAND 21203 1475' f ii J H
- CALVERT Cliff 8 NBCLEAR POWER PLANT DEPARTRIENT '- '
, CALVERT CUFF 8 NUCLEAR POWER PLANT
s . LUSOY, MARYAND 20657 .r t
4 l 1
-I March 12, 1990 ,
l p-
', i U. S Nuclear Regulatory Commission:
. Docket Nos. 50 317 & 50 318'-
License Nos. DPR 53 & 69 J Document Control De.sk
' 1 White Flint North MailiStop P1-317- .
11555 Rockville Pike. *
- Rockville, MD 20850 -
t W.
Dear' Sirs:
- The' attached supplemental LER 89-23.. Revision 1,'is being sent to you'as-
- required under 10 CFR 50.73. guidelines.
Should you have any questions regarding this report, we would be pleased to
. discuss them with'you.' 3
'Very truly yours,
, n- -
i RL.E.-Denton Manager
~i GAL:1rr-cc: . William T.. Russell .
Director, Office of Management.Information -l
.and Program Control j G. C. Creel'
. Messrs:
C. H. Cruse ex J..'R. Lemons L. B. Russell .
R. P. Heibel [y@
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" Postulated. Pipe Rupture in the Turbine Building Service Water System (SRW) Renders Both Auxiliary Building SRW Subsystems Unavailabla event aAfs e Len asunseen e aspoet oAre in l criaea e Acetaties sev0Lvoo es Mo887w DAY YeAn VeAn
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Asernact to . e.oe t.. - . an c. . an no This supplement to LER 89-023 is submitted to clearly identify our position as previously stated in the Calvert Cliffs Nuclear Power Plant's Safety Evaluation Report issued August 28, 1972, and to reclassify the reportability to "other" and the LER as 1 a voluntary report.
At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on December 20, 1989 it was postulated that a reportable condition may I have existed as a result of a plant configuration that could potentially result in <
l- unavailability of both safety-related Service Water (SRW) Subsystems. At the time of determination, Unit 1 was in cold shutdown with the Reactor Coolant System (RCS) partially filled. at atmospheric pressure, and 114 degrees F. The Unit 2 reactor was defueled, with the reactor vessel partially drained, the vessel head detensioned, and the RCS at atmospheric pressure and ambient temperature.
It was postulated that a pipe rupture in the non safety-related SRW Subsystem thatl serves the Turbine Building could result in rapid draining of both of the independent, safety-related SRW Subsystems that serve the Auxiliary Building. The loss of both Auxiliary Building SRW trains could subsequently result in unavailability of the Emergency Diesel Cenerators. The reported condition does not describe an actual event; therefore, it was not contributed to by any actual compcnent or system failures. Based on a review of the NRC's Safety Evaluation Report for Calvert Cliffs Nuclear Power Plant (CCNPP), it is clear that we were only ar.alyzed for a LOCA concurrent with a loss of off-site power. Therefore, it is clear our SRW System was not designed to cope with a seismic event and a simultaneous loss of off site power.
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rewr n m.. w , mea ,manwim I. DESCRIPTION OF EVENT -i At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on December 20, 1989 it was postulated that a reportable condition may have existed at Calvert Cliffs. The condition was the >
result of a plant configuration that could have potentially resulted in tha unavailability of both independent, safety related Service Water (SRW)-
Subsystems. This determination wrs made during a routine reportability +
review for a Non Conformance Report (NCR). . The NCR described a condition whereby a postulated pipe rupture in >the non safety-relat_ed SRW Subsystem .
that serves the Turbine Building could result in rapid draining of. both of the independent, safety related SRW Subsystems ~ that serve the Auxiliary Building. The- loss of both - Auxiliary Building . SRW Subsystems could subsequently. result in unavailability of the Emergency Diesel Generators (EDGs). :
L At the time.of determinatica, Unit I was in cold shutdown with the Reactor Coolant System (RCS) partially filled, at atmospheric pressure, and-114 degrees F. The Unit 2 resotor was defueled, with the reacter vessel partially drained,.the vessel head detensioned, and the RCS at atmospheric pressure and ambient temperature.
The- purpose .of the SRW System is to remove heat from the main
. turbine-generator plant components, containment cooling unito, spent fuel pool heat exchangers, and various EDG heat exchangers, and to transfer that heat to the Saltwater System. Although the SRW piping configuration differs slightly between Unit i and Unit 2, each unit is basically '
comprised of two independent, safety-related SRW Subsystems. in the Auxiliary . Building which operate in parallel with a single, non-safety-
, related SRW Subsystem-in the Turbine Building.
Both - Auxiliary Building - SRW Subsystems and the Turbine Building SRW Subsystem are needed during normal plant operations. For Unit 2, the two Auxiliary Building SRW Subsystems are connected to the-- Turbine Building SRW Subsystem by a common, non safety-related connection from the SRW discharge header where the SRW- System enters the Turbine Building. For ,
Unit 1, .the two Auxiliary Building SRW Subsystems are onnected to the Turbir.e Building SRW Subsystem by a common, non-safety-related pipe located where the SRW System exits the Turbine Building and connects to the SRW suction header. As a result of these common piping connections, the non-safety-related Terbine Building SRW Subsystem essentially cross-connects with the two safety-related Auxiliary Buildieg SRU Subsystems.
1 The ability to isolate the Turbine Building SRW Subsystem from the Auxiliary Building SRW Subsystems is provided by dual, air operated isolation valves on the discharge header piping of each Auxiliary Building SRW Subsystem, and by check valves in the suction header piping of each Auxiliary Building SRW Subsystem. The isolation valves are located in the l
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- safety-related Auxiliary Subsystem piping prior to connection with the Turbine Building Subsystem piping, and the check valves are located in the m
safety related portions of the Auxiliary Building SRW Subsystem suction
. header piping - The Turbine Building isolation valves can be operated from the main Control Room, and close automatically following receipt of a Safety Injection Actuation Signal (SIAS) or loss of instrument air. Other than SIAS or loss of instrument air, there are no other automatic closure l functions associar.ed with the' isolation valves.
Calculations have been - performed assuming a worst-case, double ended guillotine pipe break in 'the non safety-related Turbine Building SRW piping. It was also assumed that the break would occur'in the Unit 2 SRW piping . configuration. The piping configuration for~ Unit 2 is much less
= conservative .than Unit 1 configuration because its' cross-connection.
occurs just downstream of the non critical service water valves. The Unit 1 cross-connection is . downstream of all Turbine Building loads., The calculation results indicate that under the previously mentioned -*
conditions, breakflow could empty the SRW System in less. time than is '
required , for the isolation valves to close following receipt.of.a SIAS.
The calculations also indicate that the SRW System could be drained before an operator could act to isolate the break under non SIAS conditions. It should be noted that assumption of a double-ended guillotine type break is more conservative than is required under our licensing and design basis for a moderate energy break in a line that is designed as Seismic Category II, and was constructed to ANSI B31.1. However, informal calculations ,
indicate that even a moderately sized pipe break would result in a rapid loss of SRW' inventory.
Calvert Cliffs Nuclear Power' Plant (CCNPP) is only analyzed for a loss of. )
coolant accident (IDCA) concurrent with a loss of off-site power, and not
! for a seismic event concurrent with a loss of off-site power. Therefore, the condition described in this report is being reported under "other" as L a vcluntary LER.
, The following identifies CCNPP current design assumptions, active and L passive failures as described in our Updated Final Safety Analysis Report L (UFSAR), Section 9.5.
Current Design Assumotions I
L Single Failure Analvj h 4 Needed 4 Needed Comoonent For Normal OPS For LOCI SRW Heat Exchangers Note (a) 1 SRU Pumps 2 1 EDCs 0 1 estC pow 3eea *3 .
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s a) Two are needed. However, the two subsystems may be cross connected and one heat exchanger may be utilized to remove the full heat load.
Active Failures System Comoonent Tvoe of Failure Service Water Turbine Building. Fails to close on SIAS isolation valves. ,
Consecuences
.Valvee are actuated by a redundant channel and would shut, isolating service water as required. ,
System Component Tyne of Failure Service Water 12.EDG Supply / Fails to seek-header return CVs with pressure Consecuences Diesel generator 12 does not receive any cooling water. This could result in cross-connecting one subsystem of each unit _ and possibly draining one subsystem by over-flowing in the other unit's head tank. However, each unit would still have a subsystem in operation and this is . sufficient to remove all necessary heat. Diesel generators 11 and 21 are cooled - and provide sufficient electrical power, System- Comoonent Tvoc of Failure Service Water Turbine building Fails to close'under return check valve. reverse flow.
Consecuences Since in all cases two check valves are provided in-series,'the second valve would close_providing isolation.
Note: As shown above sufficient numbers of all other active components are supplied to provide sufficient redundancy for all modes of operation.
Passive Failure Durine Containment Sumo Recirculation System Location of Ruoture Service Water 12 EDG supply / return manual cross connect valves.
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Conseauengg.g One . subsystem from each unit ~ would be drained and rendered inoperable.' i However, one subsystem - in each unit would continue to operate. This is ,
' adequate to provide the necessary cooling'for each-unit. No single rupture in any location could cause the loss of both subsystems of a unit as two, normally closed valves are provided where two subsystems are tied together.
Flooding Due to a Passive Failure ;
Indication in ,
Structure Flooded Control Room System Ruotured i Service Water Room High level alarm Saltwater in the room with normal service water head tank level.
Consecuences Saltwater to the Service Water Room would be stopped by closing remote manual valves from the Control Room. The containment coolers would be shut down and_ heat removed fro.n the containment would be via the spray system.
Service water would continue to operate until the service water temperature reached' 120 F, which would occur approximately 21 minutes after loss of .
saltwater, based on an- initial service water. temperature of 95 F. This is !
considered. to be sufficient time,' to determine which subsystem has ruptured ;
and to re-establish saltwater flow in the other subsystem.
1 When both units are in operation, cool ' service water would be provided to diesel generator No. 12 by the other unit's Service Water System.. Saltwater for this other unit is functioning normally. This- is accomplished as follows:
- a. Diesel generator No. 12 automatically provides power to the accident unit-channel ZB for Unit 1 (ZA for Unit 2),
- b. Valves on the-discharge-of service water pump No. 13 (23) are normally open to Service Water Subsystem ZB (ZA).
- c. The circuit breaker is remotely closed to provide power to service water pump No. 13(23) from channel ZA (ZB).
- d. The service water pumps on the unit which has the ruptured Saltwater System are shutdown,
- e. The pressure seeking valves automatically supply diesel generator No.12 cooling water from the other unit.
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Structure' Flooded Control Room System Ruotured q Service Water Room- High level. alarm Saltwater i in the room with i normal service i water head tank i level, i Conseaugngga ,
.One subsystem from each un'it would be drained. However, the other subsystem would continue to operate and'is sufficient to provide all necessary service ~!
' water. The entire-contents of one Service Water-System would not-flood out ;
the service water pumps and motors.
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- Based on the above and a review of NRC's Safety Evaluation Report for CCNPP, it is clear that we were only analyzed for a 1DCA concurrent with a loss of off-site '
l' power. Therefore, it-is clear our Service Water System was not designed to cope- S
-with a seismic event with a simultaneous loss of off site power.
Report' ability as an event or condition prohibited by Technical Specifications (T.S.) is-related to T.S._3.7.4.1, which requires that "at least two independent service water loops shall be OPERABLE in modes 1, 2, 3, and 4" is.not an' issue as= ;
described in LER 89 023, Revision O. It is determined 'at - this time that the current . design and configuration- of the . SRW System meets. the - intent of this ~
Technical Specification and the original licensing' and design. bases.. The original licensing and design bases are the same as the current plant: license and d
design bases.
The condition described in this. report is'not being considered for reportability as a condition that-is outside the plant design basis for the SRW System and'the
~
EDGs. The SRW design basis is described in Section 9.5.2.2 of the. Updated Final Safety Analysis Report (UFSAR), and states that the SRW System "has been divided into two subsystems in the Auxiliary Building.to meet single failure criteria."
It also states that "during normal operation both [SRW) subsystems are . . . .;
independent to the degree necessary to assure the safe operation and shutdown of the plant assuming a single failure."
The system description for the EDGs is found in UFSAR.Section 8.4.1.2, and states that "the emergency diesel generators and their auxiliaries are designed to w ithstand Seismic Category 1 accelerations and are installed in Category 1 structures." The SRW System directly supplies cooling water to the EDGs and is considered to be auxiliary equipment to the EDCs. UFSAR Section 9.5.2.2 and 8.4.1.2 meet the intent of the original design basis, which is the same as our current design basis.
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ol0 0l 7 0F 1l0 rent u n . m .me w sca mnswim The assumptions assumed in LER 89 023, Revision 0 are more conservative than the current licensed plant design bases. Investigations perforn.ed in late 1989 previously assumed a double end guillotine pipe rupture in the non safety-related )
portion of the SRW System as the event initiator. We recognize a seismically ,
inducou pipe rupture and concurrent loss of off site power as a possible event J scenario, however, we were not required to analyze this scenario. However, an analysis of the postulated scenario and determination if a significant safety l concern exists is provided in Section III, Analysis of Event. 1 4 :
Calvert Cliffs Nuclear Power Plant Units 1 & 2 Safety Evaluation, issued August j 28, 1972', Section 3.2.5 stat'es,"in part "The Auxiliary Systems include the Chemical and Volume Control System, Shutdown Cooling System, Component Cooling Water System,, Service Water System, Saltwater System....
The; Service Water and Saltwater Systems' provide cooling required' for vital plant safety features. These systems were revised during our review to provide greater separation and redundancy so that they could sustain single failure of active or passive components without loss of the required cooling capability. l The design bases, functions, and descriptions of the Calvert Cliffs Auxiliary Systems are substantially the same as for other plants. that have been recently reviewed and approved for operating licenses. On the basis of our - comparison of these systems with those of other approved plants and our evaluation of the adequacy of each system we concluded that the Calvert Cliffs Auxiliary Systems are acceptable."~ 1 1
1 II. CAUSE OF EVENT N/A l 1
-III. ANAINSIS OF EVENT -
The postulated event was discovered during a routine reportability review for Non-Conformance Report (NCR). NCR 8391 stated a concern that "a rupture, without l-a SIAS (turbine building isolation valves do not shut), occurring in the Turbine Building will cause a loss of both subsystems. In the event of a loss of l off-site power this could render both EDGs inoperable." l An analysis of the problem and determination if a significant safety concern exists is provided below. I 1
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Information obtained from our Probabilistic Risk Assessment (PRA) Unit for East .I Coast earthquake risks put the probability for low energy earthquakes at 1.1E 7 1 per hour per event and 1E 9 per hour per event for potentially damaging earthquakes.
Even if we postulate a damaging earthquake, a catastrophic failure of the ;
non safety-related (NSR) portion of the Service Water System is unlikely because l the Turbine Building is a Seismic Class II structure. It has a working stress design for 0.08 g horizontally and 0.053 g vertically (OBE accelerations). While l a conservative analysis was not performed , the building is relatively stiff and would not collapse'under Safe Shutdown Earthquake (SSE) conditions.
Inspections of industrial mill buildings and power plants have been performad-after earthquakes much stronger than our SSE. Except for those on soft soil or. ,
4 of very unusual configurations, those buildings performed well. - In addition, there were very few instances of piping damage. These points have been nade many
' times - with' the NRC and ACRS by Seismic Qualification Utility Group '(SQUG) consultants. .
The few piping fhilures noted by SQUG consultants were caused by: ,
- Unanchored equipment
- Severe building displacement / relative motion with little piping flexibility (such as buried pipes entering buildings, closely spaced rigid supports at' expansion joints).
Steel piping is inherently rugged. '"his is borne out by the testing which lead to ASME Code Case _N-411. and by the recent reports suggesting 'far less conservative design for small bore pipe.
The NRC has placed a relatively low priority-on seismic qualification (USI A 46).
Their position is that seismic is not a major contributor to nuclear risk.
Walkdowns by expert teams on five of the oldest plants resulted in few
. corrections. Some of these plants weren't even designed for earthquakes.
l It is postulated that a seismic event could occur following a LOCA. Emergency Operating Procedure E0P-5-(Loss of Coolant) directs operators during recovery to restart Turbine Building service water and~ restart equipment such as instrument -
air compressors. A calculation was performed and showed that a seismic event following a LOCA is an extremely low probability event scenario - on the order of IE-8 events in any 30 day period.
Anothe' failure mechanism that needs to be considered is passive failure. NUREG g CR.4407 (Pipe Break Frequency Estimates for Nuclear Power Plants) puts passive
- pipe failure for balance of plant systems at 4.4E-8 failures per hour per event.
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I Our UFSAR only discusses passive failures during recirculation following a LOCA.
By comparing passive failure risk to a potentially damaging seismic risk, it is ,
obvious that the passive failure is the higher risk - even though Sth are small.
Our Abnormal Operating Procedures (AOP) recognize a pipe ruptwe as a possible )
event. AOP 78 (loss of Service Water) Sections IV and V (Rupture of a subsystem) I provides direction to operators in the event of a loss inventory in the Service ;
Water System. There is explicit direction given to isolate the Turbine Building 1 a so that a rupture in one subsystem will not drain the other subsystem".
The NRC also recognized that plants of our vintage were not designed to withstand I a loss of their safety related (SR) Service Water System during non LOCA events.
Their evaluation of our May 20, 1980 loss of service water accident concluded the loss of service water event at Calvert Cliffs did not result in damage to any j plant equipment either safety c,r non safety related, and taken by itself does nog )
represent a cause for concern. The significance of nhe event lies in the fact i that it involved two fundamental aspects consitered in the design of .
safety related systems: J
- 1. Interaction between safety and non safety related systems and components; .
- 2. Common caused failure of redundant safety systems.
The review of the event by the Office for Analysis and Evaluation of Operational Data (AOD) revealed no immediate safety concerns; however, there is a need to reevaluate the isolation provisions at the interface between the safety at d non safety related portions of the Service Water System at Calvert Cliffs as weil as generically.
The primary concern with a loss of service water during non 14CA events is loss of cooling for the EDGs. This scenario would render EDGs inoperable and may place us in a station blackout. Calvert Cliffs is currently able to maintain the plant in a safe shutdown condition for four hours with no AC power. Station Blackout Procedure (EOP-7) provides direction to operators on how to restore both i the Service Water System and AC System to operation.
Given the low probability of a damaging earthquake and the small likelihood that it will cause a catastrophic failure of the Service Water System, we can cor.clude that there is no exigent need to take immediate action to modify the system due to an earthquake risk.
There is sufficient operator guidance to cope with a postulated loss of service water with a simultaneous loss of off site power. As an additional measure, Operations has increased the frequency of leak rate monitoring of the Service Water System, ggen. m.
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olo 1l0 or 1 l0 vert u . n.w anc s mmw nn IV. 00RRECTIVR ACTIONS TAKEN A task force has been assembled to determine appropriate long term corrective actions. In the interim, prior to startup, compensatory actions will be established, although the described is not part of our current licensing and design bases. The following compensatory actions will be established prior to startup.
- Change Alarm Manual to include immediate isolation of Turbine Building header on large rupture indications.
. Inform operators of the status of this issue prior to Unit 1 startup, i
V. ADDITIONAL INfDENATION N/A l l
VI. IDENTIFICATION.0F.00KPONENTS REFERRED IV IN THE LER ,
IEEE 803 IEEE 805 ;
Componenr/Systeu Component ID Code System ID Code Auxiliary Building NF Auxiliary Feedwater System BA Containment Cholers BK Control Room NA l Emergency Diesel Generator EK p
Isolation Valve ISV l
Reactor Coolant System AB
! Reactor Vessel RCT l Safety Injection System JE/BQ/BP L Saltwater System BS l Service Water System BI Spent Fuel Pool Cooling System DA Turbine Building NM Turbine Generator T- ,
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