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Category:DEFICIENCY REPORTS (PER 10CFR50.55E & PART 21)
MONTHYEARML20140G4481997-05-0101 May 1997 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby Diesel Generator Sys.Recommends That Springs Be Inspected on Periodic Basis,Such as During Refueling Outages ML20135D8011996-11-26026 November 1996 Part 21 Rept Re Two Safety Related Valves Supplied by Velan Valve Corp Were Not in Compliance W/Originally Supplied QA Documentation.Returned Valves to Velan in May 1996 & on 961120 Velan Advised That Valves Had Been Misplaced ML20058K3841993-12-0909 December 1993 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby DG Sys,Regarding Potential Problem W/ Subcover Assembled Atop Power Head ML20127H2301993-01-15015 January 1993 Part 21 Rept Re Potential Defeat in Component of Dsrv & Dsr Enterprise Standby DG Sys.Starting Air Distributor Housing Assemblies Installed as Replacement Parts at Listed Sites ML20126B0421992-12-17017 December 1992 Final Part 21 Rept Re Potential Problem W/Steel Cylinder Heads.Initially Reported on 921125.Caused by Inadequate Cast Wall Thickness at 3/4-inch-10 Bolt Hole.Stud at Location Indicated on Encl Sketch Should Be Removed ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20118B4391992-09-11011 September 1992 Part 21 Rept Re Degradation in Abb Type 27N Undervoltage Relays Used in Electrical Switchgear.Recommends That Users Review Applications Requiring Exposures Greater than 1E03 Rads TID W/Time Delay Function Option ML20094L1271992-03-13013 March 1992 Amend 1 to Part 21 Rept 159 Re Potential Defect in Power Cylinder Liner.Initially Reported on 920115.Caused by Liner/ Block Fit & Localized Matl Microstructure.All Drawings & Specs Revised to Address Matl Design Requirements ML20005G6831990-01-0505 January 1990 Part 21 Rept Re Installation Instructions for Grommet Use Range for Patel Conduit Seal P/N 841206.Conduit Seals in Environ Qualification Applications Inspected for Proper Wire Use Range & Grommets Replaced ML20248E0121989-09-13013 September 1989 Supplemental Part 21 Rept Re Potential Problem W/Six Specific Engine Control Devices in Air Start,Lube Oil, Jacket Water & Crankcase Sys.Initially Reported on 890429. California Controls Co Will Redesign Valve Seating ML20245B6651989-06-15015 June 1989 Part 21 Rept 150 Re Potential Defect in Component of Dsr Standby Diesel Generator.Cause of Failure Determined to Be Combination of Insufficient Lubrication to Bushings.Listed Course of Action Recommended at Next Scheduled Engine Maint ML20154J4101988-05-12012 May 1988 Addendum 1 to Supplemental Part 21 Rept 145 Re Potential Problem w/F-573-156 Pressure Sensor/Lube Oil Trip.Initially Reported on 880429.Addl 1-1/8-inch Diameter W/Deep Counterbone in Pressure Head Added to Activate Valve ML20153B6191988-04-29029 April 1988 Part 21 Rept Re Potential Defect in Component of Dsr or Dsrv Standby Diesel Generators Supplied to Utils.Recommends Return of Subj Components in Onsite Warehouse Storage & Suggests Surveillance of Devices Already Installed ML20153B2811988-02-22022 February 1988 Part 21 Rept Re Defect in Operation of Gas Monitors Using Fixed Filters.Initially Reported on 880212.Caused by Error in Software for Monitors.Vendor Testing Mod to Software & in Interim Recommends Changing Filters Every 48 H SNRC-1423, Part 21 Rept Re Accdos (HP-85) & Rms Computer Codes. Initially Reported on 880126.Engineering Change Request Implemented Involving Rms Computer Code Software & Should Be Completed by Feb 19881988-01-29029 January 1988 Part 21 Rept Re Accdos (HP-85) & Rms Computer Codes. Initially Reported on 880126.Engineering Change Request Implemented Involving Rms Computer Code Software & Should Be Completed by Feb 1988 SNRC-1386, Part 21 Rept Re Forged Steel Fittings Mfg by Std Fittings Co & Inspected by X-Ray Insp,Inc.Investigation Still Going on & NRC Resident Inspector Will Be Informed of Util Findings & Remedial Action to Be Taken for Fittings Requiring Rework1987-12-21021 December 1987 Part 21 Rept Re Forged Steel Fittings Mfg by Std Fittings Co & Inspected by X-Ray Insp,Inc.Investigation Still Going on & NRC Resident Inspector Will Be Informed of Util Findings & Remedial Action to Be Taken for Fittings Requiring Rework SNRC-1388, Part 21 Rept Re Defects in GE Termination Lugs Used for Control Wiring in 4160 Volt Switchgear Equipment.Initially Reported on 871016.GE Advised That Wire Will Be Stripped W/ Stripping Tools Which Incorporate Stops to Control Length1987-10-19019 October 1987 Part 21 Rept Re Defects in GE Termination Lugs Used for Control Wiring in 4160 Volt Switchgear Equipment.Initially Reported on 871016.GE Advised That Wire Will Be Stripped W/ Stripping Tools Which Incorporate Stops to Control Length SNRC-1366, Part 21 Rept Re 2 Inch 90 Degree Elbow Forged Steel Fittings Mfg by Std Fittings Co & Inspected by X-Ray Insp Inc. Existing Plant Documents Will Be Reviewed in Attempt to Locate Listed Identified Fittings & Completed by 8709301987-08-14014 August 1987 Part 21 Rept Re 2 Inch 90 Degree Elbow Forged Steel Fittings Mfg by Std Fittings Co & Inspected by X-Ray Insp Inc. Existing Plant Documents Will Be Reviewed in Attempt to Locate Listed Identified Fittings & Completed by 870930 ML20237G5781987-08-0707 August 1987 Part 21 & Deficiency Rept 142 Re Defect in Starting Air Distributor Sys of Dsr Standby Diesel Generator.Drain Sys Will Be Permanently Installed in Start Air Distributor ML20236N7131987-08-0505 August 1987 Part 21 Rept Re Potential Problem W/Vibration in Lube Oil & Jacket Water Sys.At Rancho Seco Site,Fatigue Cracks Found in Welds of Support Members in Lubricating Oil Sys.Cracks Due to Vibration in Lube Oil Sys.Addl Preventive Action Taken ML20210S3351987-02-0404 February 1987 Part 21 Rept Re Cracking in O-ring on Latch Mechanism on Contactors in Nuclear Exciters.Initially Reported on 861216. Sprecher & Schuh Changed O-ring Matl to Viton E60C. Contactors Will Be Returned to Vendor & Rebuilt ML20206S0841986-06-30030 June 1986 Part 21 Rept Re Possible Cut Wires in Wire Harness of Bbc Brown Boveri K600/K800 Circuit Breakers.Initially Reported on 860509.Safety Implications Listed.Gear Guard Designed to Prevent Cut Wires ML20141A1081986-03-31031 March 1986 Part 21 Rept 134 Re Potential Defect in K-1 Relay in Generator Voltage Regulator.Initially Reported on 860219. Corrective Action Must Take Into Account Equipment Operating Range of 100-137.5-volt Dc ML20154K3611986-03-0505 March 1986 Part 21 Rept Re Possible Wiring Defect in Fabricated Primer Chamber Assemblies for Standby Liquid Control Valves. Initially Reported on 860214.Listed Corrective Actions Will Be Completed by 860328 ML20137X3891986-02-27027 February 1986 Part 21 Rept Re Possible Defect in Model KMG-HRH Noble Gas Radiation Monitor W/Enhanced Detector Model KDGM-HR. Initially Reported on 860222.Vendor Will Request Utils for Listed Plants to Determine If Monitors Defective ML20153G1021986-02-19019 February 1986 Part 21 Rept Re Potential Defect in K-1 Relays of Dsr or Dsrv Standby Diesel Generators Mfg by Gould/Ite.Relays Have Incorrect Voltage Range.Dc Voltage Source or Voltage Regulator Circuitry Requires Mod ML20151Y8141986-02-0404 February 1986 Part 21 Rept Re Colt-Pielstick Engine Tripping Out on High Speed When Started for Test Purposes at Seabrook.Caused by Source of Air Pressure Staying On.Engines Will Be Modified to Positively Vent Air from Rack Boost Cylinder ML20140A5281985-12-19019 December 1985 Part 21 Rept Forwarding Ltr Sent to Customers Re Check Valves Missing Lock Welds on Hinge Supports or Hinge Support Capscrews,Per 851121 Request.List of Customers Receiving Ltr Also Encl ML20198C4981985-11-0606 November 1985 Part 21 Rept Re Potential Defect in Dsr or Dsrv Standby Diesel Generator Intake & Exhaust Valve Springs Mfg by Betts Spring Co.Users Recommended to Inspect Engines for Broken Springs & Identify Original Mfg ML20137W2611985-09-30030 September 1985 Update to Part 21 Rept Re Insp for Cracked Welds Discovered in Generator Brackets.Insp for Callaway Unit 1 Postponed Until Next Refueling Outage (Probably Apr 1986).No Cracks Discovered During Insp at Wolf Creek ML20132D4141985-09-25025 September 1985 Part 21 Rept Re Defective Rochester Instrument PR-2035 Under Voltage Relays Exhibiting Tendency to Drift from Designated Setpoint &/Or Inability to Maintain Deadband Adjustment. Investigation Continuing.Final Rept by 851011 ML20132B0261985-09-0303 September 1985 Part 21 Rept Re Possibility of Engine Ingestion of Unwelded Part in American Air Filter Standby Diesel Generator Intake Silencer.Immediate Hold Should Be Placed on Diesel Engines/Intake Silencers Not Yet in Svc SNRC-1197, Interim Deficiency & Part 21 Rept Re Procurement & Installation of Coils & Starters for Square D Co 480-volt Motor Control Ctrs.Initially Reported on 850719.Matls & Stock Numbers Revised1985-07-23023 July 1985 Interim Deficiency & Part 21 Rept Re Procurement & Installation of Coils & Starters for Square D Co 480-volt Motor Control Ctrs.Initially Reported on 850719.Matls & Stock Numbers Revised ML20128N4981985-07-22022 July 1985 Part 21 Rept Re Tendency of Rochester Instrument Sys Undervoltage Relays Installed in Emergency Switchgear for Degraded Voltage Protection to Drift from Calibr Setpoints. Defective Relays Replaced SNRC-1189, Forwards Summary Description of Util Resolution Re Tdi 10CFR21 Notifications Issued After 840401.Util Will Inform NRC of Actions Re Future Tdi Part 21 Notification Until Final SER Issued in Jul 19851985-06-28028 June 1985 Forwards Summary Description of Util Resolution Re Tdi 10CFR21 Notifications Issued After 840401.Util Will Inform NRC of Actions Re Future Tdi Part 21 Notification Until Final SER Issued in Jul 1985 ML20127H5411985-05-17017 May 1985 Part 21 Rept Re Potential Defect in Engine Generator Control Panels of Dsr or DSRV-16 Standby Diesel Generator.Review of Generator Control Panel Installation by Each Facility Recommended ML20127N0621985-05-13013 May 1985 Part 21 Rept Re Incorrect Short Time Delay Band Lever Inadequately Installed in Low Voltage K-line Circuit Breaker Electromechanical Overcurrent Trip Device.Visual Insp to Determine If Correct Lever Installed Recommended ML20112J7061985-04-0202 April 1985 Part 21 Rept Re Possible Defect in mineral-insulated Cable Assemblies.Design,Mfg & Qualification Testing of high-range Containment Area Monitoring Sys Completed.Design Qualification Rept Encl ML20100E0651985-03-29029 March 1985 Part 21 Rept Re Series of Cracked Welds Discovered in Generators.Welds Occurred in Conical Baffle Section of Coil Guards.All Units Will Be Inspected.Affected Plants Listed ML20112A0701985-03-0606 March 1985 Part 21 Rept Re Cracked Weld in Coil Guards - Conical Baffle.Lilco Has Copy of Weld Procedure WPS-402-700510, Rev 0 & Will Arrange to Perform Repairs SNRC-1130, Deficiency Rept Re Failure to Install HVAC Penetration Seals to Protect Two Trains of Redundant Safe Shutdown Motor Control Ctrs.Initially Reported on 841129.Sealing of Gaps Completed on 8412081984-12-28028 December 1984 Deficiency Rept Re Failure to Install HVAC Penetration Seals to Protect Two Trains of Redundant Safe Shutdown Motor Control Ctrs.Initially Reported on 841129.Sealing of Gaps Completed on 841208 ML20101P9361984-12-20020 December 1984 Part 21 Rept Re Fatigue Cracks on Fuel Injection Pump Delivery Valve Holders of PC-2 & PC-2.3 Emergency Diesel Generators.All Valve Holders W/Improper Radius Will Be Replaced SNRC-1112, Deficiency Rept Re Apparent Accelerated Corrosion in Four Reactor Bldg Svc Water Strainers Mfg by RP Adams Co,Inc. Initially Reported on 841015.Two Strainer Welds Repaired Under ASME Section XI Article IWD-40001984-11-27027 November 1984 Deficiency Rept Re Apparent Accelerated Corrosion in Four Reactor Bldg Svc Water Strainers Mfg by RP Adams Co,Inc. Initially Reported on 841015.Two Strainer Welds Repaired Under ASME Section XI Article IWD-4000 ML20106D3831984-10-0808 October 1984 Part 21 Rept Re Adjustment of Low Voltage Shutoff & Turnon for GE Dedicated Inverters Improperly Set.Utils Notified to Check & Readjust as Necessary ML20084L1151984-05-0808 May 1984 Suppl 1 to Part 21 Rept Re United Electric Controls Fuel Oil Filter Differential Pressure Switches at Shoreham Plant. Switch Also Provided to Listed Plants.Corrective Actions Defined in 840328 Rept Will Be Applied ML20083P9801984-04-0909 April 1984 Part 21 Rept Re Broken Engine Driven Lube Oil Pump Discharge Nozzle Boss.Caused by Overtorquing of Tapered Thread Adapter.Adapter Changed to Straight Threads W/Undercut ML20087P8711984-03-28028 March 1984 Part 21 Rept Re Two Failed Fuel Oil Filter Differential Pressure Switches.Caused by Pressure Pulsation on Low Pressure Side of Switch.Switch Replaced ML20113C6001984-03-26026 March 1984 Discusses Bonney Forge Part 21 Rept.S&W Notified That Evaluation Required Re Matl Delivered to Shoreham Not Meeting Nca 3967.4(e)(2) Requirements SNRC-1018, Interim Deficiency Rept Re Failure of Turbocharger on Emergency Standby Diesel Generator.Temporary Bypass Valve Around Turbocharger Prelube Drip Line Orifice Installed. W/Certificate of Svc.Related Correspondence1984-03-0909 March 1984 Interim Deficiency Rept Re Failure of Turbocharger on Emergency Standby Diesel Generator.Temporary Bypass Valve Around Turbocharger Prelube Drip Line Orifice Installed. W/Certificate of Svc.Related Correspondence 1997-05-01
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20140G4481997-05-0101 May 1997 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby Diesel Generator Sys.Recommends That Springs Be Inspected on Periodic Basis,Such as During Refueling Outages ML20135D8011996-11-26026 November 1996 Part 21 Rept Re Two Safety Related Valves Supplied by Velan Valve Corp Were Not in Compliance W/Originally Supplied QA Documentation.Returned Valves to Velan in May 1996 & on 961120 Velan Advised That Valves Had Been Misplaced ML20080G4691995-01-26026 January 1995 Record of Telcon W/Nrc & Licensees 950126 to Clarify Position Re Dispositioning of Exempt Sources Listed in Section 6.3.3 of Shoreham Termination Survey Final Rept Dtd Oct 1994 ML20069F0001994-01-24024 January 1994 Vols 1-4 to Shoreham Decommissioning Project Termination Survey Final Rept ML20058K3841993-12-0909 December 1993 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby DG Sys,Regarding Potential Problem W/ Subcover Assembled Atop Power Head ML20057F2261993-09-30030 September 1993 Safety Evaluation Supporting Exemption Request from Requirements of 10CFR50.54(q) for License NPF-82 ML20056C7181993-07-14014 July 1993 SE Supporting Amend 10 to License NPF-82 ML20045B3551993-06-11011 June 1993 LER 93-001-00:on 930429,refueling Jib Crane Moved in Vicinity of Spent Fuel Pool Using vendor-supplied Lifting Eye in Violation of NUREG-0612.Caused by Failure to Identify Crane as Heavy Load.Meetings held.W/930611 Ltr ML20045C8881993-06-0808 June 1993 Vols 1 & 2 to Refueling Jib Crane 1T31-CRN-008A Incident Root Cause Analysis. W/One Oversize Encl ML20044C1181993-02-28028 February 1993 Shoreham Nuclear Power Station Updated Decommissioning Plan. ML20128P6451993-02-28028 February 1993 Snps Decommissioning Project Termination Survey Final Rept for Steam Turbine Sys (N31) ML20128P7431993-02-19019 February 1993 Rev 3 to 93X027, Nuclear QA Surveillance Rept ML20127H2301993-01-15015 January 1993 Part 21 Rept Re Potential Defeat in Component of Dsrv & Dsr Enterprise Standby DG Sys.Starting Air Distributor Housing Assemblies Installed as Replacement Parts at Listed Sites ML20126B0421992-12-17017 December 1992 Final Part 21 Rept Re Potential Problem W/Steel Cylinder Heads.Initially Reported on 921125.Caused by Inadequate Cast Wall Thickness at 3/4-inch-10 Bolt Hole.Stud at Location Indicated on Encl Sketch Should Be Removed ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20128B9641992-10-31031 October 1992 Rev 0 to Shoreham Decommissioning Project Termination Survey Plan ML20118B4391992-09-11011 September 1992 Part 21 Rept Re Degradation in Abb Type 27N Undervoltage Relays Used in Electrical Switchgear.Recommends That Users Review Applications Requiring Exposures Greater than 1E03 Rads TID W/Time Delay Function Option ML20099H5781992-07-31031 July 1992 Rev 4 to Shoreham Defueled Sar ML20114A6311992-07-28028 July 1992 Shoreham Decommissioning Plan ML20101K5791992-06-25025 June 1992 Long Island Power Authority Shoreham Decommissioning Project,Shoreham Nuclear Power Station,Technical Rept on Water Processing & Water Mgt Activities for Reactor Pressure Vessel & Wet Cutting Station ML20094L1271992-03-13013 March 1992 Amend 1 to Part 21 Rept 159 Re Potential Defect in Power Cylinder Liner.Initially Reported on 920115.Caused by Liner/ Block Fit & Localized Matl Microstructure.All Drawings & Specs Revised to Address Matl Design Requirements ML20082M5081991-08-26026 August 1991 Rev 3 to Shoreham Defueled Sar PM-91-125, Monthly Operating Rept for Jul 1991 for Shoreham Nuclear Power Station1991-07-31031 July 1991 Monthly Operating Rept for Jul 1991 for Shoreham Nuclear Power Station PM-91-112, Monthly Operating Rept for Jun 1991 for Shoreham Nuclear Power Station1991-06-30030 June 1991 Monthly Operating Rept for Jun 1991 for Shoreham Nuclear Power Station PM-91-075, Monthly Operating Rept for Apr 1991 for Shoreham Nuclear Power Station1991-04-30030 April 1991 Monthly Operating Rept for Apr 1991 for Shoreham Nuclear Power Station ML20024G7171991-04-22022 April 1991 LER 91-001-00:on 910324,RB Normal Ventilation Sys (Rbnvs) Outboard Exhaust Valve Closed for No Apparent Reason.Cause Inconclusive.Sys Restored to Normal Lineup & Rbnvs Outboard Valve Will Be Stroked on Routine basis.W/910422 Ltr SNRC-1806, Revised Pages 2 & 6 to Encl a of 10CFR50.59 Annual Rept for 19901991-04-15015 April 1991 Revised Pages 2 & 6 to Encl a of 10CFR50.59 Annual Rept for 1990 PM-91-058, Monthly Operating Rept for Mar 1991 for Shoreham Nuclear Power Station1991-03-31031 March 1991 Monthly Operating Rept for Mar 1991 for Shoreham Nuclear Power Station PM-91-037, Monthly Operating Rept for Feb 1991 for Shoreham Nuclear Power Station1991-02-28028 February 1991 Monthly Operating Rept for Feb 1991 for Shoreham Nuclear Power Station PM-91-016, Monthly Operating Rept for Jan 1991 for Shoreham Nuclear Power Station1991-01-31031 January 1991 Monthly Operating Rept for Jan 1991 for Shoreham Nuclear Power Station SNRC-1797, 10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 19901990-12-31031 December 1990 10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 1990 SNRC-1794, Shoreham Nuclear Power Station Annual Operating Rept,19901990-12-31031 December 1990 Shoreham Nuclear Power Station Annual Operating Rept,1990 SNRC-1799, Lilco 1990 Annual Rept1990-12-31031 December 1990 Lilco 1990 Annual Rept ML20069Q3901990-12-31031 December 1990 Shoreham Nuclear Power Station Decommissioning Plan. (Filed in Category P) ML20028H0231990-09-28028 September 1990 LER 90-007-00:on 900907,unplanned Actuation of ESF Sys Occurred During I&C Surveillance Test.Caused by Inadequate procedure.SP44.650.16 Revised to Require That Leads Lifted & Individually separated.W/900928 Ltr ML20056A2001990-07-31031 July 1990 Safety Evaluation Supporting Amend 6 to License NPF-82 PM-90-097, Monthly Operating Rept for June 1990 for Shoreham Nuclear Power Station1990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Shoreham Nuclear Power Station ML20055E3911990-06-25025 June 1990 Safety Evaluation Supporting Amend 5 to License NPF-82 PM-90-083, Monthly Operating Rept for May 1990 for Shoreham Nuclear Power Station1990-05-31031 May 1990 Monthly Operating Rept for May 1990 for Shoreham Nuclear Power Station 05000322/LER-1988-0151990-05-16016 May 1990 LER 88-015-02:on 880916,seismic Monitoring Instrumentation, Including Peak Acceleration Recorders,Removed from Svc for More than 30 Days Due to Corrosion on Scratch Plates.Cover Gasket Replaced & Thermal Barrier Mount to Be Installed 05000322/LER-1986-0391990-05-16016 May 1990 LER 86-039-01:on 861006,unplanned Initiation of Reactor Bldg Standby Ventilation Sys Occurred W/All Rods Inserted in Core.Caused by Faulty Design of Retaining Device.Warning Signs Attached to Valve Actuator & Valve Mod Initiated 05000322/LER-1986-0321990-05-16016 May 1990 LER 86-032-01:on 860728,RWCU Isolated on High Differential Flow Sensed by Steam Leak Detection Sys While Placing Filter Demineralizers in Operation.Cause Not Determined. Operating Procedures Revised to Monitor RWCU Sys 05000322/LER-1987-0091990-05-16016 May 1990 LER 87-009-01:on 870203,full Reactor Trip Occurred Due to Perturbation in Ref Leg.Caused by Spurious Low Level Reactor Pressure Vessel Water Level Signal.Existing Level & Pressure Transmitters Replaced W/Newer Models 05000322/LER-1987-0221990-05-16016 May 1990 LER 87-022-01:on 870604,HPCI Test Valve to Condensate Storage Tank,Globe Valve & Hpci/Rcic Test Valve to Condensate Storage Tank Failed to Close Against Sys Operating Pressure.Disc of motor-operated Valve 37 Modified 05000322/LER-1989-0051990-05-16016 May 1990 LER 89-005-01:on 890321,results of Local Leak Rate Test of Core Spray Suction Valve a Determined That Leakage,When Combined W/All Type B & C Penetration Leakages,Exceeded Tech Spec Limit.Caused by Normal Valve Degradation 05000322/LER-1989-0031990-05-16016 May 1990 LER 89-003-01:on 890310,emergency Diesel Generator (EDG) 102 Manually Shutdown During 18-month Surveillance Test Due to Failure of EDG Output Breaker.Cause Not Determined. Replacement Breaker Installed in Cubicle 102-8 05000322/LER-1985-0591990-05-16016 May 1990 LER 85-059-01:on 851219,half Reactor Trip,Full NSSS Shutoff Sys Isolation & Reactor Bldg Standby Ventilation Sys Initiation Occurred Due to Loss of Power to Reactor Protection Sys Bus B.Assembly Breaker Reset 05000322/LER-1987-0351990-05-16016 May 1990 LER 87-035-02:on 871221,880106 & 0330,high Energy Line Break Logic Isolations of RWCU & Main Steam Line Drain Valves Occurred.Caused by Problems W/Temp Monitoring Units. Grounding Scheme Changed & Transformers Rewired 05000322/LER-1988-0031990-05-16016 May 1990 LER 88-003-01:on 880322,unplanned Automatic Initiation of Reactor Bldg Standby Ventilation Sys Side a Occurred During Deenergization of Relay.Caused by Close Placement of Relay Terminals.Wiring Inside Electrical Panels Reworked 1997-05-01
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> 1 LONG ISLAND LIGHTING COM PANY
- SLCO!
wwww 3 SHOREHAM NUCLEAR POWER STATION ew nve r.+ m u e P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 Dkect Dial Numkr August 12, 1983 SNRC-951 Dr. Thomas E. Murley Regional Administrator Office of Inspection & Enforcement - Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Gas Analyzer Cell Catalysts Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Dr. Murley:
On July ll, 1983, in accordance with 10CFR50.55(e), we reported verbally to Region I a potential deficiency involving contain-ment gas analyzer cell catalysts. This had been previously reported under 10CFR21 by COMSIP, Inc. to Mr. R. C. DeYoung, (Director I&E). Subsequent Shoreham specific engineering review indicates that the analyzer cell deficiency applies to Shoreham and is considered a reportable deficiency. Therefore, this letter will serve as our thirty-day written report pertaining to this deficiency.
Description of Deficiency The primary containment atmosphere may contain fission fragment iodine following certain accident conditions involving core damage. Calculations performed by Stone & Webster confirm the potential for a primary containment iodine concentration of 0.016 mm Hg partial pressure. This concentration adversely affects the cell catalyst such that the usable life is decreased to a duration in the order of 8-10 days. After this period, the analyzer would provide erroneous readings for oxygen and hydrogen
, concentrations (percent by volume) in both the primary containment ,
and the recombiner discharge. Loss of indication due to analyzer failure will reduce the ability of the control room to accurately monitor hydrogen /cxygen generating events.
8308190306 830812 PDR ADOCK 05000322 PDR i S FC-8935.1 1
tb
A August 12, 1983 SNRC-951 Page 2 Corrective Action Our corrective action involves installation of upgraded analyzer cell catalysts manufactured by COMSIP, Inc. This material has been procured by our E&DCR P-4443 ~and is. currently being in-stalled by E&DCR P-4443A. The new model analyzer-cell catalysts are currently undergoing testing by COMSIP and as of July 27, 1983, have passed the 127-day mark of a scheduled 180-day con-secutive test run; The testing is bcing done with a sample atmosphere of 0.02 mm Hg partial pressure _(iodine) . The replace-ment cell catalysts will, therefore, function properly under Shoreham's post-accident monitoring environment. Upon installa-tion of the new cell catalysts, no further modifications will be required. This modification will be completed prior to exceeding a power level of five (5) percent.
Very truly yours, J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station BWC:jm ec: Mr. Richard DeYoung, Director NRC Office of Inspection and Enforcement Div. of Reactor Operations Inspection Washington, D.C. 20555 Mr. J. C. Higgins USNRC Resident Inspector NRC Site Office (TSC)
All Parties Listed in Attachment 1
l 1
. l ATTACRMENT 1 Lawrence Brenner,'Esq. . Herbert H. Brown, Esq.
Administrative Judge Lawrence Coe Lanpher, Esq.
Atomic Safety and Licensing Karla J. Letsche, Esq.
Board Panel Kirkpatrick, Lockhart, Hill U.S. Nuclear Regulatory Christopher & Phillips Commission 8th Floor
. Washington, D.C. 20555 1900 M. Street, N.W.
Washington, D.C. 20036 Dr. Peter A. Morris Administrative Judge Mr. Marc W. Goldsmith
. Atomic Safety and Licensing , Energy Research Group Board Panel 4001 Totten Pond Road U.S. Nuclear Regulatory Waltham, Massachusetts 02154 Commission e Washington, D.C. 20555 MHB Technical Associates "
, 1723 Hamilton Avenue Dr. George A. Ferguson Suite K School of Engineering San Jose, Califcrnia 95125 Howard University 2300 Fifth Street Washington, D.C. 20059 Stephen B. Latham, Esq. .
Twomey, Latham & Shea 33 West Second Street Daniel F. Brown, Esq. P.O. Box 398 Attorney Riverhead, New York 11901 Atomic Safety and Licensing Board Panel ,
U.S. Nuclear Regulatory Ralph Shapiro, Esq. C Commission Cammer and Shapiro, P.C.
Washington, D.C. 20555 9 East 40th Street New York, New York 10016 Bernard M. Bordenick, Esq.
David A. Repka, Esq. Matthew J. Kelly, Esq.
U.S. Nuclear Regulatory State of New York Commission Department of Public Service Washington, D.C. 20555 Three Empire State Plaza Albany, New York 12223 Mr. James Dougherty l
3045 Porter Street Washington, D.C. 20008 l
l le i
Attachment 2 o
4 ' SNPS-1 FSAR
,J E m 9.4.6 Drywell Air Cooling System , .
'9.4.6.1 Design Bases )
The drywell air cooling system circulates the atmosphere within the drywell to remove heat and maintain design temperatures. The drywell will be maintained at a maximum temperature of 146 F (avg. 127 F) during normal operation. The control rod drive. area design temperature is 150 F, while maximum allowed temperature in the area is 165 F. In the proximity of electrical equipment, maximum temperature is 130 F. The system is not a safety related system.
9.4.6.2 System Description 1
During normal operation air is circulated as shown in Fig. 9.4.6-1 within the drywell by two unit coolers each
, comprised of four cooling coils, four 10l000 cfm fans, intake and discharge dampers, and supply and return ductwork. Air is supplied by three of the four fans in each unit to the lower level of the drywell and returned to the unit from the top of the drywell. Water from the reactor building closed loop cooling water system is used as the cooling medium for the coils. No control of air flow or water flow is provided for this system.
However, should the requirements for cooling decrease, one or more of the operating fans can be shut down manually from the main control room. g NJ To ensure continuous operation during loss of offsite power and no accident signal present, the drywell unit cooler fans, dampers, and valves are connected to the emergency power supply.
The units are designed to operate during al'1 normal plant operations, j 9.4.6.3 Safety Evaluation The unit coolers are designed to meet the cooling requirements of the drywell with three fans in each unit running and one fan as a spafe. Upon failure of any one of the three running fans or -
associated dampers in each unit, an alarm will sound in the main control room and the spare fan will be started manually from the main control room. ,,
Upon indication of high pressure in the drywell or low reactor water level signals, the drywell unit coolers are automatically shut down, and - all primary containment isolation valves in the cooling water piping are closed automatically.
I The system is not safety related. However, all ductwork and .
! equipment are seismically supported to ensure they stay in place and do not damage safety related equipment in the area.
O 9.4-16
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. SNPS-1 FSAR A w 9.4.6.4 Tests and Inspections All components are tested and inspected as separate components and as integrated systems. After the system is completely installed, air flows are measured and adjusted to meet design flow rates. During plant normal shutdown, the system will be inspected and readjusted, if required, to meet design flow rates.
9.4.6.5 Instrumentation Application Drywell unit cooler controls including selector switches, monitors, and system alarms are located on panels in the main control room.
Alarms are provided in the main control room for the following conditions:
- 1. Control Rod Drive area high temperature
- 2. Upper drywell exhaust high temperature
- 3. Drywell head area exhaust high temperature
- 4. Reactor building closed loop cooling water return water high temperature
- 5. Drywell unit cooler high supply air temperature
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9.4-17 Revision 1 - March 1976
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- Attachment 3 PtEF & Et;_P!/ 2py
- , CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE ,
l LIMITING CONDITION FOR OPERATI'ON 3.6.1.7 Drywell average air temperature shall not exceed 145*F.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With the drywell average air temperature greater than 145*F, reduce the cverage air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volume weighted
(- average of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hoers:
Elevation Azimuth
- a. 68'-0" 13*, 320 -
- b. 80'-0" 190*, in CRD area
- c. 83'-0" 25*, 135*, 255*
- d. 110'-0" '165*, 350'
- e. 145'-0" 55*, 230*
- f. 162'-6" ,
Reactor Vessel Centerline i
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PLANT SYSTEMS g 'g g]
(, 3/4.7.9 AREA TEMPERATURE MONITORING
, LIMITING CONDITION FOR OPERATION 3.7.9 The temperatureof each area shown in Table 3.7.9-1 shall be maintained within the limits indicated.
PPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.
ACTION:
With one or more areas exceeding the temperature limit (s) shown in Table 3.7.9-1:
- a. For more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, in lieu of any report required by Specification 6.9.1, prepare and submit' a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days providing a record of the amount by which and the cumulative time the temperature in the affected area exceeded its limit and an analysis to demonstrate the continued OPERABILITY of the affected equipment.
b.
By more than 30*F, in addition to the Special Report required above,
' within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.
SURVEILLANCE REQUIREMENTS 4.7.9 The temperature in each of the areas shown in Table 3.7.9-1 shall be determined to be within its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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.. PR00F & REVH COPY
,, TABLE 3.7.9-1 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F)
- a. Control Room 90
- b. Chiller Equipmen't Room (El 63') 104
- c. Relay Room 104
- d. Emergency Switchgear Rooms 104
- e. Battery Rooms 104
- f. bieselGeneratorRooms 120
- g. Screenwell House 104
- h. Reactor Building - Secondary Containment
- 1. General Areas 104
- 2. Refueling Area .
110 ,
- i. Reactor Building - Primary Containment
- 1. General Areas 150 (f . 2.
3.
Area Beneath RPV Drywell Head Area 150*
185 -
A 165*F during Scram.
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CONTAINMENT SYSTEMS g !g { ,
i BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY .
~This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48 psig in the event of a.LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The ifmi'tations on drywell and suppression chamber internal pressure -
ensure that the containment peak prassure of 46.0 psig does not exceed the design, pressure of 48 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 5.7 psid. The upper limit of Figure 3.6.1.6-1 will limit the total pressure to 46.0 psig which is less than the design pressure and is consistent with the safety analysis. The lower limit of Figure 3.6.1.6-1 ensures that the peak LOCA temperature will not exceed the safety analysis value and the containment external pressure requirements are satisfied.
3/4.6.1.7 DRYVELL AVERAGE AIR TEMPERATURE .
The limitation on drywell average air temperature ensures that the l containment peak air temperature does not exceed the design temperature of l 340*F during LOCA conditions and is consistent with the safety analysis. .
3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The 18-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be sealed closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA.or steam line break accident. Maintaining these valves sealed closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. To provide assurance that the 18-inch valves cannot be inadvertently opened, they are sealed closed in accordance with Standard Review Plan 6.2.4, which includes me'chanical devices to seal or lock the valve closed or prevent power from being supplied to the valve operator.
The use of the drywell and suppression chamber purge lines.is restricted to the 4-inch and the 6-inch purge supply and exhaust isolation valves since, unlike the 18-inch valves, the 4-inch and the 6-inch valves will close during a LOCA or steam line' break accident and therefore the SITE BOUNDARY dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during PURGING operations. The design of the 4-inch and the 6-inch purge supply and exhaust
- isolation valves meets the requirements of Branch Technical Position CSB 6-4,
- . " Containment Purging During Normal Plant Operations."
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r PLANT SY. STEMS BASES hk h {}ljg i
3/47[7 FIRE SUPPRESSION SYSTEMS i
The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available ,to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located. The fire suppression system consists of the water system, deluge, CO2 systems, Halon systems and fire hose stations. The collective capability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the, facility fire protecticn program.
In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.
When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is all' owed to provide an alternate means of fire fighting than if the inoperatile equipment is the primary means of fire suppression.
The surveillance requirements provide assurances that the minimum
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OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient' volume of Halon in the Halon cylinders by verifying the weight and pressure of the tanks.
In the event the fire suppression water system becomes inoperable, imediate corrective measures must be taken since this system provides the '
major fire suppression capability of the plant. ~
3/4.7.8 FIRE RATED ASSEF"dLIES . .
The OPERABILITY fire damage of the fire barriers and barrier penetrations ensure that will be limited.
These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.
I 3/4.7.9 AREA TEMPERATURE MONITORING The araa temperature limitations ensure that safety related equipment will
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not be subjected to temperatures in excess of their environmental qualification temperatures.
Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.
3/4.7.10 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with
.{ the assumptions of the feedwater controller failure analysis for FSAR Chapter 15.
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