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Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO LER)
MONTHYEARML20216E5401999-09-0707 September 1999 Special Rept:On 990826,discovered That Meteorological Sys Upper Wind Speed Cup Set Broken,Causing Upper Wind Channel to Be Inoperable.Cup Set Replaced & Channel Restored to Operable Status on 990826 ML20210R1051999-08-0606 August 1999 Special Rept:On 990628,cathodic Protection Sys Was Declared Inoperable After Sys Did Not Pass Acceptance Criteria of Bimonthly Surveillance.Work Request 98085802 Was Initiated & Connections on Well Anode Were Cleaned or Replaced ML20206P2081998-12-31031 December 1998 Special Rept:On 981218,inoperability of Meteorological Monitoring Instrumentation Channels,Was Observed.Caused by Data Logger Overloading Circuit.Replaced & Repaired Temp Signal Processor ML20198B1341998-12-14014 December 1998 Revised Special Rept:On 980505,discovered That Certain Fire Barriers Appeared to Be Degraded.Caused by Removal of Firestop Damming Boards.Hourly Fire Watches Established in Affected Areas ML20196D4041998-11-19019 November 1998 Rev 1 to Special Rept:On 980618,determined That Method Used to Calibrate Wind Speed Instrumentation Loops of Meteorological Monitoring Instrumentation Sys Does Not Meet TS Definition for Channel Calibration.Procedure Revised ML20153B0531998-09-16016 September 1998 Special Rept:On 980817,errors in Implementation of Selected Licensee Commitment Testing Requirements on Fire Protection Sys Instruments,Was Discovered.Caused by Error in Interpretation of SLC Requirement.Will Revise Procedures ML20236M9151998-07-0707 July 1998 Special Rept:On 980611,determined That Required Firewatch Patrol Had Been Missed.Caused by Firewatch Being Performed on Wrong Unit Due to Human Error.Employee Was Verbally Counseled on Firewatches & Documentation Was Corrected ML20236G4451998-07-0101 July 1998 Special Rept:On 980618,declared Wind Speed Instrumentation Loops of Meteorological Monitoring Instrumentation Sys Inoperable.Caused by Failure to Meet TS Definition of Channel Calibr.Will Revise Selected Licensee Commitment ML20248K1431998-06-0202 June 1998 Special Rept:On 980505,discovered That Certain Fire Barriers Appeared to Be Degraded.Caused by Shrinkage of Foam & Improper Installation During Construction of Plant.Posted Fire Watches & Repaired Firestop F-AX-348-W-134 ML20247H5351998-04-12012 April 1998 Special Rept:On 980415,missed Insp of Fire Hose Caskets Was Discovered.Caused by Error in Transferring Info from One Procedure to Another.Planned Rev of Applicable Procedure to Include Gasket Insp at Appropriate Frequency ML20216B0211998-04-0606 April 1998 Special Rept:On 980325,determined That Loose Parts Monitoring Sys Being Inoperable for Greater than Thirty Days.Caused by Incorrect Testing.All Channels of Loose Parts Monitoring Sys Tested Utilizing Revised Test Method ML20217K9271998-03-26026 March 1998 Special Rept:On 971229,procedure Step for Closing Safety Injection Pump Cold Leg Injection Isolation Valve Was Inadvertently Skipped.Caused by Injection of Water Into RCS from Rwst.Simplified Procedures & Discussed Event ML20216D5641998-03-0505 March 1998 Special Rept:On 980204,discovered That Fire Detection Panel Was Apparently Not Communicating W/Several Local Fire Detectors.Caused by Faulty Computer Sys Cards.Replaced Four Computer Cards in Sys ML20202C4701998-02-0505 February 1998 Ro:On 971228:Unit 1 Loose Parts Monitoring Sys Channel 6 Was Declared Inoperable Due to Excessive Static on Channel. Caused by Loose Connection.Work Order Has Been Written to Pursue Repairs ML20138E6851997-04-24024 April 1997 Special Rept:On 970318,Unit 1 Loose Parts Monitoring Sys Channel 13 Was Declared Inoperable Due to Sporadic Electical Static.Channel Was Removed from Svc & Entered Into TS Action Item Logbook as Inoperable ML20149M7251997-01-20020 January 1997 Special Rept:On 961209,Unit 1 Loose Parts Monitoring (Lpm) Sys Channel 20 Declared Inoperable Due to No Signal Being Received from Field.Lpm Channel 20 & 22 Operable & Providing Monitoring Coverage for Primary Side of 1D S/G ML20134K4901996-11-0606 November 1996 Special Rept:On 961009,selective Licensee Commitment for Operability of Fire Protection Sprinkler Sys Not Maintained. Continuous Fire Watch Established within One H Following Identification of Incorrect Remedial Action ML20134H1331996-11-0404 November 1996 Special Rept:On 961004,Unit 1B DG Failed Due to Failure of Motor Operated Pot,Electronic & Mechanical Governor,Governor Droop Relay & Mechanical Binding of Fuel Rack Control Linkage.Dg Procedures Will Be Revised ML20113A1801996-06-17017 June 1996 Special Rept:On 960521,declared Detectors A01 for Zone 69 & A04 for Zone 60 Inoperable Because Detectors Effectively Isolated from Area in Intended Protection.Detectors Relocated,Tested & Declared Operable on 960524 ML20100H9801996-02-20020 February 1996 Special Rept:On 960111,Unit 1 Loose Parts Monitoring Channel 21 Declared Inoperable,Due to Spurious,Unexplainable Electronic Bursts.Work Request Initiated to Pursue Corrective Action ML20100H9751996-02-20020 February 1996 Special Rept:On 960111,Unit 2 Loose Parts Monitoring Sys Channel 7 Declared Inoperable,Due to pre-amp Bias Voltage Indicating Zero Volts Twice During Previous Seven Days.Work Request Written to Pursue Corrective Action ML20097F5011996-02-11011 February 1996 Special Rept:On 960102,Unit 2 Loose Parts Monitoring Sys Channel 17 Was Declared Inoperable.Two Other Channels Operable & Providing Coverage Against Loose Parts ML20096E7731996-01-12012 January 1996 Special Rept:On 951214,unit 2 DG Valid Failure Occurred. Caused by Fuel Line Fitting Backing Off from Cylinder Head Connection,Which Resulted in Fuel Oil Leakage.Dg Successfully Started,Run & Declared Operable on 951215 ML20096A8761995-12-18018 December 1995 Special Rept:On 951120,during Periodic Surveillance Testing, Lpms Channel 5 Declared Inoperable.Caused by Erratic Preamp Bias Voltage Indications.Work Request 95048483 Initiated to Perform Corrective Maint During Unit 1 Cycle 9 ML20094Q5811995-11-13013 November 1995 Special Rept:On 951014,auxiliary Bldg Filtered Exhaust Sys Pump Room Heater Declared Inoperable Due to Blown Fuse & Not Restored to Operable Status within 7 Days Per Ts. Technical Investigation Will Be Performed ML20094B8291995-10-25025 October 1995 Special Rept:On 950919,loose Parts Monitoring Sys Channel 1 Declared Inoperable Due to Erratic Preamp Bias Indication. Work Request Written to Investigate & Repair Channel ML20098A4641995-09-19019 September 1995 Special Rept:On 950817,Unit 2 Lpms Channel 12 Was Declared Inoperable Due to Channel Sensor Failing Acceptance Criteria During Performance of PT/O/A4600/03 ML20092G6041995-09-14014 September 1995 Special Rept:On 950815,CNS Unit 1 DG 1A Invalid Failure Occurred Due to Main Bearing High Temp Trip Signal.Caused by Failed Splice Installed in Circuit for RTD 1LDRD5630.New RTD Installed in Main Bearing 5 ML20086H1401995-07-12012 July 1995 Special Rept:On 950615,Channel 4 Was Declared Inoperable Due to Noise Uncharacteristic of Healthy Channel Detected Via Vibration & Loose Parts Monitoring Sys.Corrective Maint Will Be Performed During 1EOC9 Outage ML20086H1431995-07-11011 July 1995 Special Rept:On 950608,Channel 13 Was Declared Inoperable. Trending of Bias Voltage & Background RMS Evaluated to Conclude Channel Was Experiencing Periodic Failures. Corrective Maint Will Be Performed During 1EOC9 Outage ML20086C6441995-06-29029 June 1995 Special Rept:On 950523,Unit 1 Train a Fuel Handling Ventilation Filter Heaters Declared Inoperable.Evaluation Done to Determine Fault ML20085M4061995-06-20020 June 1995 Special Rept:On 950501,lower Rv Tube 4 Was Declared Inoperable ML20084N7271995-05-25025 May 1995 Special Rept:On 950425,valid Failure of DG 1A Occurred. Caused by Jacket Water Thermostatic Control Valve Sticking in Position Which Reduced Engine Cw Flow Through Heat Exchanger.Thermostatic Cv Internals Removed & Replaced ML20082L2711995-04-17017 April 1995 Special Rept:On 950308,Unit 2 Cathodic Protection Sys Was Declared Inoperable & Remained Inoperable Greater than 10 Days ML20081D4851995-03-13013 March 1995 Special Rept:On 950211,actuation of PORV 1NC32B Occurred. Procedure OP/1(2)/A/6100/02 Revised to Require More Emphasis on Monitoring Pressure Indication During Sensitive Evolutions ML20080Q8701995-03-0202 March 1995 Special Rept:On 950202,Unit 1 DG 1B Invalid Failure Due to Overcurrent Breaker Trip During Governor Troubleshooting ML20149H7821994-12-20020 December 1994 Special Rept:On 941129,discovered That Selective Licensee Commitment (SLC) for Visual Insp of Fire Rated Assemblies Exceeded Due to Misinterpretation of Requirements of SLC 16.9-5.Fire Barriers Visually Inspected ML20078R0021994-12-12012 December 1994 Special Rept:On 941103,Channel 3 (Upper Rv a) Declared Inoperable.Caused by Channel Sensor Failure of Acceptance Criteria During Performance of PT/0/A/4600/03.Repair Planned for End of 2EOC7 Outage Due to Containment Entry Required ML20078K7361994-11-17017 November 1994 Special Rept on 941021,DG 1A Invalid Failure Occurred Due to Main Bearing High Temp Trip.Operability Performance Test Successfully Completed & Engine Declared Operable on 941022 ML20149G8041994-11-0101 November 1994 Special Rept:On 940922,CNS,Unit 2 Cathodic Protection Sys Declared Inoperable & Remained Inoperable for Greater than 10 Days.Wo 94080948-01 Initiated to Replace Prepackaged Anode Well 1.WO Scheduled for 941114 ML20076F3191994-10-0404 October 1994 Special Rept:On 940908,valid Failure of D/G 1A Occurred Due to Air Start Valve Sticking Open.Maint Procedure MP/0/A/7650/99 Revised,New Air Roll Criteria Developed & Sixteen Starting Air Valves Replaced ML20072P4251994-08-23023 August 1994 Ro:On 940719,channel 9 (S/G a Channel 2) Declared Inoperable.Work Request Was Generated to Repair Channel During Future Outage of Sufficient Length Since Containment Entry Required for Work ML20072E5961994-08-15015 August 1994 Special Rept:On 940715,inoperability of Unit 2 Vibration & Loose Parts Monitoring System Channel 4 & 6 Occurred.Caused by Leds Not Lighting During Performance of PT/0/B/4600/03. Work Orders 94051250-01 & 94051251-01 Initiated ML20071N8511994-07-28028 July 1994 Special Rept:On 940711,main Steam Relief Valve Exhaust Monitors Declared Inoperable Due to Engineering Calculation Concerns.Engineering Calculation CNC-1229.00-00-0047 re-performed Using EPA-400 Methodology ML20071N7441994-07-28028 July 1994 Special Rept:On 940711,delta-t Channel on Chart Recorder Found to Be out-of-tolerance Due to Drifting of Analog to Digital (A/D) Converter Card.A/D Card Replaced & delta-t Channel Chart Recorder Declared Operable ML20070K0191994-07-18018 July 1994 Special Rept:On 940630,re Inoperability of Main Steam Line Radiation Monitor 2EMF12.Work Request 94026262 Generated to Reattach 2EMF12 to Main Steamline.Work Request Completed on 940701 ML20069H0861994-05-31031 May 1994 Special Rept:On 940501,Unit 2 DG 2A Invalid Failure Occurred Due to Right Bank Turbocharger Vibration Trip.Based on Cooper-Enterprise Recommendations,Procedure Changes Made to Calibr Procedures for All Four DGs ML20065K5011994-04-13013 April 1994 Special Rept:On 940314,invalid Failure of Diesel Generator 1B Occurred Due to Output Tripping During Calibration of Electronic Governor.Dg 1B Started Successfully on 940315 & Declared Operable ML20064G2911994-03-15015 March 1994 Special Rept:On 940203,SG Channels 9 & 11 Declared Inoperable.Channels Failed to Meet Band Limited RMS Acceptance Criteria During Performance of 18 Month Channel Calibr Test.Work Request Written to Repair Channels ML20064G2771994-03-0707 March 1994 Special Rept:On 940125,Channels 6,7 & 10 Were Declared Inoperable.Channels Failed to Meet Acceptance Criteria During Performance of 18 Month Channel Calibr Test.Work Request Written to Repair Channels 1999-09-07
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20212J1891999-10-0101 October 1999 Safety Evaluation Supporting Exemption from 10CFR54.17(c)re Schedule to Apply for Renewed Operating Licenses ML20212A6271999-09-30030 September 1999 Rev 0 to WCAP-15243, Anaylsis of Capsule V & Capsule Y Dosimeters from Duke Energy Catawba Unit 2 Reactor Vessel Radiation Surveillance Program ML20217H0201999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Catawba Nuclear Station,Units 1 & 2 05000414/LER-1999-004-01, :on 980906,error During Tagout Caused de-energization of Vital Bus & Actuation of Ltop.Caused by Inadequate Work Practices.Individuals Involved Were Counseled1999-09-27027 September 1999
- on 980906,error During Tagout Caused de-energization of Vital Bus & Actuation of Ltop.Caused by Inadequate Work Practices.Individuals Involved Were Counseled
05000413/LER-1999-015, :on 990616,discovered That Auxiliary Bldg Filtered Ventilation Exhaust Sys Was Inoperable.Caused by Improperly Positioned Vortex Damper.Damper Was Repositioned Correctly & Sys Was Retested Successfully1999-09-27027 September 1999
- on 990616,discovered That Auxiliary Bldg Filtered Ventilation Exhaust Sys Was Inoperable.Caused by Improperly Positioned Vortex Damper.Damper Was Repositioned Correctly & Sys Was Retested Successfully
ML20212G2511999-09-22022 September 1999 Safety Evaluation Supporting Amends 180 & 172 to Licenses NPF-35 & NPF-52,respectively 05000413/LER-1999-008, :on 990610,operations Prohibited by TS 3.5.2, Was Violated.Caused by Inoperable Centrifugal Charging Pump. Operators Swapped to CCP 1A & Sys Parameters Were Returned to Normal.With1999-09-21021 September 1999
- on 990610,operations Prohibited by TS 3.5.2, Was Violated.Caused by Inoperable Centrifugal Charging Pump. Operators Swapped to CCP 1A & Sys Parameters Were Returned to Normal.With
05000414/LER-1999-005-02, :on 990727,missed Emergency DG TS Surveillance Concerning Verification of Availability of Offsite Power Sources,Was Declared.Caused by Defective Procedure.Revised Affected Procedure1999-09-20020 September 1999
- on 990727,missed Emergency DG TS Surveillance Concerning Verification of Availability of Offsite Power Sources,Was Declared.Caused by Defective Procedure.Revised Affected Procedure
05000413/LER-1999-009, :on 990518,inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits Was Noted. Caused by Inadequate Retest Following Surveillance Test Failure.Valve Was Retested & Returned to Service1999-09-15015 September 1999
- on 990518,inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits Was Noted. Caused by Inadequate Retest Following Surveillance Test Failure.Valve Was Retested & Returned to Service
ML20216E5401999-09-0707 September 1999 Special Rept:On 990826,discovered That Meteorological Sys Upper Wind Speed Cup Set Broken,Causing Upper Wind Channel to Be Inoperable.Cup Set Replaced & Channel Restored to Operable Status on 990826 05000413/LER-1999-014, :on 990816,missed Surveillances & Operation Prohibited by TS Was Noted.Caused by Defective Procedures or Programs Inappropriate TS Requirements.Affected Procedures/ Programs Were Revised & Testing Was Performed1999-09-0101 September 1999
- on 990816,missed Surveillances & Operation Prohibited by TS Was Noted.Caused by Defective Procedures or Programs Inappropriate TS Requirements.Affected Procedures/ Programs Were Revised & Testing Was Performed
05000414/LER-1999-004, :on 990616,CIV 2NM-221A Was Returned to Svc Without Testing,As Required by TS 3.6.3.Caused by Programmatic Deficiency.Test Procedure Has Been Revised & Subject Valve Was Successfully Tested & Returned to Svc1999-09-0101 September 1999
- on 990616,CIV 2NM-221A Was Returned to Svc Without Testing,As Required by TS 3.6.3.Caused by Programmatic Deficiency.Test Procedure Has Been Revised & Subject Valve Was Successfully Tested & Returned to Svc
ML20212B4711999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 05000414/LER-1999-003, :on 990612,unplanned Actuation of ESFAS Occurred Due to a SG High Level.Caused by Inadequate Procedural Guidance.Msiv 2SM-7 Was Closed & SG 2A Level Was Returned to Normal1999-08-31031 August 1999
- on 990612,unplanned Actuation of ESFAS Occurred Due to a SG High Level.Caused by Inadequate Procedural Guidance.Msiv 2SM-7 Was Closed & SG 2A Level Was Returned to Normal
ML20211B1281999-08-31031 August 1999 Dynamic Rod Worth Measurement Using Casmo/Simulate ML20217H0321999-08-31031 August 1999 Revised Monthly Operating Rept for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 05000413/LER-1999-012, :on 990727,adverse Sys Interaction Between Annulus Ventilation Sys & Auxiliary Building Ventilation Sys Was Discovered.Caused by Inadequate Design.Compensatory Actions Developed & Implemented.With1999-08-26026 August 1999
- on 990727,adverse Sys Interaction Between Annulus Ventilation Sys & Auxiliary Building Ventilation Sys Was Discovered.Caused by Inadequate Design.Compensatory Actions Developed & Implemented.With
ML20211A9791999-08-20020 August 1999 Safety Evaluation Granting Licensee Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section Xi,For Plant,Unit 2 ML20211C1291999-08-17017 August 1999 ISI Rept Unit 1 Catawba 1999 RFO 11 ML20211F3441999-08-17017 August 1999 Updated non-proprietary Page 2-4 of TR DPC-NE-2009 ML20210U8341999-08-13013 August 1999 Safety Evaluation Supporting Amends 179 & 171 to Licenses NPF-35 & NPF-52,respectively ML20210R1051999-08-0606 August 1999 Special Rept:On 990628,cathodic Protection Sys Was Declared Inoperable After Sys Did Not Pass Acceptance Criteria of Bimonthly Surveillance.Work Request 98085802 Was Initiated & Connections on Well Anode Were Cleaned or Replaced ML20212B4871999-07-31031 July 1999 Revised Monthly Operating Rept for July 1999 for Catawba Nuclear Station,Units 1 & 2 ML20210S2891999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Catawba Nuclear Station,Units 1 & 2 05000413/LER-1999-009, :on 990518,inoperability of Containment Valve Injection Water Sys Valve Was Noted in Excess of TS Limits. Caused by Inadequate Testing Following Surveillance Test Failure.Valve Was Retested & Restored to Service1999-07-19019 July 1999
- on 990518,inoperability of Containment Valve Injection Water Sys Valve Was Noted in Excess of TS Limits. Caused by Inadequate Testing Following Surveillance Test Failure.Valve Was Retested & Restored to Service
05000414/LER-1999-004-02, :on 990610,violation of TS 3.6.3 Was Noted Due to CIV 2NM-221A Being Returned to Service Without Testing. Caused by Procedure Deficiency.Civ 2NM-221A Was Tested & Returned to Operable Status1999-07-15015 July 1999
- on 990610,violation of TS 3.6.3 Was Noted Due to CIV 2NM-221A Being Returned to Service Without Testing. Caused by Procedure Deficiency.Civ 2NM-221A Was Tested & Returned to Operable Status
ML20209E4361999-07-0909 July 1999 SER Agreeing with Licensee General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation 05000414/LER-1999-003-02, :on 990612,unplanned Actuation of Esfa Sys Due to a SG High Level Was Noted.Caused by Inadequate Procedural Guidance.Long Term Corrective Actions to Prevent Recurrence of Event Are Being Developed1999-07-0808 July 1999
- on 990612,unplanned Actuation of Esfa Sys Due to a SG High Level Was Noted.Caused by Inadequate Procedural Guidance.Long Term Corrective Actions to Prevent Recurrence of Event Are Being Developed
ML20196K6631999-07-0707 July 1999 Safety Evaluation Supporting Licensee 990520 Position Re Inoperable Snubbers ML20209H4501999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Catawba Nuclear Station,Units 1 & 2 ML20210S2951999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Catawba Nuclear Station,Units 1 & 2 05000414/LER-1999-002-03, :on 990504,plant Was Forced to Shutdown as Result of Flow Restriction Caused by Corrosion of Afs Assured Suction Source Piping Due to Inadequate Testing. Affected Piping Was Cleaned & Flow Tested1999-06-0303 June 1999
- on 990504,plant Was Forced to Shutdown as Result of Flow Restriction Caused by Corrosion of Afs Assured Suction Source Piping Due to Inadequate Testing. Affected Piping Was Cleaned & Flow Tested
ML20209H4561999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Catawba Nuclear Station,Units 1 & 2 ML20196A0001999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Catawba Nuclear Station,Units 1 & 2 ML20196L1881999-05-31031 May 1999 Non-proprietary Rev 1 to DPC-NE-3004, Mass & Energy Release & Containment Response Methodology ML20206T4771999-05-31031 May 1999 Rev 3 to UFSAR Chapter 15 Sys Transient Analysis Methodology ML20206P5201999-05-14014 May 1999 Safety Evaluation Accepting GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20206N8391999-05-0404 May 1999 Rev 16 to CNEI-0400-24, Catawba Unit 1 Cycle 12 Colr ML20196A0041999-04-30030 April 1999 Revised Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20206R1811999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20206N8261999-04-22022 April 1999 Rev 15 to CNEI-0400-24, Catawba Unit 1 Cycle 12 Colr. Page 145 of 270 of Incoming Submittal Not Included ML20205S5551999-04-21021 April 1999 Safety Evaluation Accepting Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions 05000413/LER-1999-004, :on 990310,operation Prohibited by TSs Was Noted.Caused by Incorrect TS Requirements for Cravs & Auxiliary Bldg Filtered Ventilation Exhaust Sys Actuation Instrumentation.Submitted Lar.With1999-04-12012 April 1999
- on 990310,operation Prohibited by TSs Was Noted.Caused by Incorrect TS Requirements for Cravs & Auxiliary Bldg Filtered Ventilation Exhaust Sys Actuation Instrumentation.Submitted Lar.With
ML18016A9011999-04-12012 April 1999 Part 21 Rept Re Defect in Component of DSRV-16-4,Enterprise DG Sys.Caused by Potential Problem with Connecting Rod Assemblies Built Since 1986,that Have Been Converted to Use Prestressed Fasteners.Affected Rods Should Be Inspected ML20205N3651999-04-12012 April 1999 Safety Evaluation Accepting IPE of External Events Submittal ML20205N2381999-04-0909 April 1999 Safety Evaluation Supporting Amends 178 & 170 to Licenses NPF-35 & NPF-52,respectively ML20205N2121999-04-0808 April 1999 Safety Evaluation Supporting Amends 177 & 169 to Licenses NPF-35 & NPF-52,respectively ML20206R1931999-03-31031 March 1999 Revised Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20205P9521999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Catawba Nuclear Station,Units 1 & 2 ML20205B3101999-03-26026 March 1999 Safety Evaluation Supporting Amends 176 & 168 to Licenses NPF-35 & NPF-52,respectively 1999-09-07
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Catauh KuclearStation 1:0 lha 2%
Cimer, iC29710 DUKEPOWER September 26, 1990 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Catawba Nuclear Station Docket 50-413 and 50-414 C90-021-1, Rev 1 1-C90-0075, Rev 1: 2-C90-0248 f
Gentlemen:
Attached is a report on SETPOINT DRIFT ON PRESSURIZER SAFETY VALVES DURING SUCCEESIVE SURVEILLANCE TESTS.
This report'. is being submittod as a "Special" Report with potential 10CFR Part 21 reportability concerns.
Very truly yours,
'l y
Tony
. Owen Station Manager ken: REPORT.SP xc:
Mr. S. D. Ebneter American Nuclear Insure ~s Regional Administrator, Region II c/o Dottie Sherman,^ANI Library.
U. S. Nuclear Regulator Commission The Exchange, Suite 74%
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101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT 06032 t
i M & M Nuclear Consultants Mr. K. Jabbour s
1221 Avenues of the Americas U. S. Nuclear Regulatory Commission New York, NY 10020 Office of Nuclear Reactor Regulation Washington, D. C.
20555 INPO Records Center Suite 1500 Mr. W. T. Orders 1100 Circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station
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DUKE POWER COMPANY r
CATAWBA NUCLEAR STATION j
PROBLEM INVESTIGATION REPORT NO. 1-C90-0075, Rev. 'l SETPOINT DRIFT ON PRESSURIZER SAFETY VALVES ~
DURING SUCCESSIVE SURVEILLANCE TESTS ABSTRACT On February 3-4, 1990, two of'the thhee Pressurizer Safety Valves (PSVs) were removed from the Unit 1 Pressurizer to meet the Techt leal Specification' (T/S)
-l Surveil. lance test ;guirement. Both valves were shi., ped to the test facility at Wyle Laboratories. On February 23, Maintenance Engineering Services (MES),
l personnel were contacted that one valve-failed to neet the 2485 psig +/ - 1%.
setpoint requirement'during its as-received setpcint test. Unit I was in "No.
Mode",' core defueled, at this time.. Additionally, on June 22, 1990, two of the.
3 three PSVs were. removed from the Unit 2 Pressurizer.and shipped to Wyle-Laboratories. On July 22, MES personnel were contacted that both valves failed to meet the 2485 psig +/- 1% setpoint requirement.
Unit 2 was in "No Mode" at i
this time. A review of past surveillance tests indicated that the PSVs have i
failed to meet the setpoint requirement in 60% of their as-received setpoint L
tests. Setpoint drift on the PSVs beyond the T/S allowed +/- 1% acceptable range is attributed to a (manufacturer) Functional Design' Deficiency. The cause of the setpoint drift is attributed to the affect normal plant operation and system condition can have on PSV performance.
A Safety Evaluation will be performed, and station documents will be revised, as needed, 'o evaluate increasing the acceptable range to +/- 3% on the PSV setting.. This report is being submitted as a special Report wi'h potential 10CFR Part 21 reportability concerns.
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DUKE POWER COMPANY / CATAWBA NUCLEAR' STATION t
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.'P*.R 1-C90-0075/Specicl Report,.R:v. 1 Pag) 2 -
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t BACKGROUND The Reactor Coolant [EIIS:AB) (NC) System consists of four heat transfer loops connected in pr.rallel to the Reactor Vessel [EIIS:VSL].
Each loop contains a i
Reactor Coolant Pump [EIIS:P] and a Steam Generator [EIIS:HX] (S/G). The B loop'
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alscs includes a Pressurizer, a Pressurizer Relief Tank (PRT), interconnecting piping [EIIS: PSP] and instrumentation necessary for operational control, j
NC System pressure is controlled by the use of the Pressurizer where water and a
steam are maintained in equilibrium by electric heaters [EIIS:EHTR] and water sprays. -Steam can be formed (by the heaters) or condensed (by the Pressurizer-spray) to reduce pressure variations due to contraction and expansion of-the Reactor coolant. Three spring loaded safety valves-[EIIS:V] 1(2)NC-1, 2.and 3, are connected to the Pressurizer and discharge to the Pressurizer Relief. Tank (PRT).
The three Pressurizer Safety Valves (PSVs) per Unit are of the totally enclosed l
pop-type.
The valves are manufactured by Dresser,.Model 6-31749A-2-XNC019,'and are spring-loaded, self-activated with back pressure compensation. The combined i
capacity of the valves is equal to, or greater than, the maximum surge rate resulting from complete loss of load without Reactor Trip or any other control.
'i Temperature indicators [EIIS:XI].in the safety valve discharge' manifold alert the Operator to the passage of steam due either to leakage or valves lifting.
Each Unit's Pressurizer is equipped with three Power Operated Relief Valves (PORVs) which limit system pressure for a large power mismatch and thus prevent actuation of the fixed high-pressure Reactor trip. The PORVs are operated automatically or by remote manual control.
The operation of these valves also limits the undesirable opening of the spring-loaded safety valves. Remotely I
operated valves are provided to isolate the inlet to each PORV.if-excessive l
leakage occurs.
Each Unit's PRT condenses and cools the discharge from the PSVs and PORVs.
l Steam is discharged through a sparger pipe under the water level. The PRT is equipped with an internal spray and a drain which are used to cool the tank following a discharge. 'The PRT is protected against a discharge exceeding the design value by two rupture discs which discharge into the Reactor Containment.
Technical Specification (T/S) 3.4.2.1 requires that a minimum of one PSV be operable with a lift setting of 2485 psig +/- 1% for Unit operation in Mode.4, l
Ilot Shutdown, or Mode 5, Cold Shutdown. With no PSV operable, inuediate suspension of all operations involving positive reactivity changes and placement of an operable residual heat removal loop into operation is required.
5 T/S 3.4.2.2 requires all PSVs be operable with a lift. setting.of 2485 psig +/-
1% for Unit operation in Mode 1, Power Operation, Mode'2, Startup, and Mode 3.
l Hot Standby. With one PSV inoperable, action shall be taken to restore the I
inoperable PSV to operable status within 15 minutes, or be in at least Mode 3..
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least Mode 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
'C
DUKE POWER COMPANY / CATAWBA NUCLEAR STATION
- PIR 1-C90-0075/Specicl R: port, R;v. 1 l
- P g3 3 T/S 4.0.5 identifies that inservice inspection and testing of the PSVs shall be performed in accordance with surveillance requirements identified in the ASME Boiler and Pressure Vessel Code,Section XI, 1983 Edition including Addenda through the Summer 1983 Addenda (applicable Edition and Addenda required by 10CFR part 50, Section 50.55a(g)).
Subsection IWV-3510 of the ASME Code identifies the testing schedule which requires all PSVs be tested once every-five years. Subsection IWV-3510 required that setpoint testing be in accordance with ASME PTC 25.3-1976, Safety and Relief Valves performance Test Codes.
The 1986 Edition of the ASME Section XI Code,7 Subsection INV-3510, references that setpoint testing be in accordance with ANSI /ASME OM-1-1981, Requirements For Inservice Performance Testing of Nuclear Power Plant Pressure Relief i
Devices.
This document allows a +/- 3% acceptable range for setpoint test.
Surveillance testing to subsequent edition of codes and addenda, or portions thereof, is acceptable to 10CFR Part 50, Section 50.55a(g)(3)(v).. However,.the present T/S limitation require that the +/- 1% acceptance range be maintained at Catawba.
EVENT DESCRIPTION On February 3-4, 1990, PSVs INC-001 and 1NC-002, Serial Numbers (S/Ns) BS-02867 and BS-02872, were removed from the Unit 1 Pressurizer to be tested per the T/S l
Surveillance requirements (reference Work Requests 4401. SWR and 4402 SWR).
Replacement valves S/Ns BS-02870 and BS-02871, previously tested and verified to l
be at the required setpoint, were installed. On February 10, the valves removed-j were transported to the test 'acility at Wyle Laboratories.
On February 23, l
Maintenance Engineering Servis.es (HES) personnel were contacted that valve S/N l
BS-02867 (INC-001) failed to meet the 2485 psig +/- 1% requirement during its as-received setpoint test. Unit I was operating in "No Mode", core defueled, at this time.
Station Compliance and MES personnel discussed the concerns of PSVs setpoints drifting outside of the T/S allowed tolerance. An MES review of past setpoint test results indicated that of the 13 tests performed on the PSVs, there were 7 valves that failed the Wyle Laboratories as-received setpoint test.'
MES initiated Problem Investigation Report (PIR) 1-C90-0075 identifying the recent drift in setpoint of valve S/N BS-02867 as a continuing problem.
Valve S/Ns BS-02867 and BS-02872 were refurbished, and final setpoint testing was acceptable.
These valve were returned to Catawba as replacement spares.
On June 22, 1990, PSVs 2NC-001 and 2HC-002, S/Ns BS-02868 and BS-02866, were removed from the Unit 2 Pressurizer to be tested per T/S (reference Work Requests 4472 SWR and 4499 SWR).
The valves removed were then transported to Wyle Laboratories for a setpoint verification test. On July 22, MES personnel were contacted that both valves failed to meet the 2485 psig +/- 1% requirement
.during their as-received setpoint test. Unit 2 was operscing in "No Mode" at this time. Valve S/Ns BS-02867 and BS-02872 were instal.cd as replacements.
t!
s.
. DUKE POWER COMPANY / CATAWBA NUCLEAR STATION
- PIR 1-C90-0075/Special R: port, Rev. 1 l
- p ge 4 CONCLUSION This incident is attributed to a (manufacturer) Functional Design Deficiency due to.the PSVs not maintaining the T/S setpoint of 2485 psig within the +/- 1%
acceptable range. Catawba has experienced a drift in setpoints outside of the
+/- 1% tolerance in 60% of the PSVs tested at Wyle Laboratories (9 failures of 15 tests performed).
A search of the Nuclear Plant Reliability Database System (NPRDS) identified 62 failures of PSVs to be within their setpoint requirements during surveillance testing (valves tested in response in equipment problems, t
1.e. seat leakage, were'not included). Of these 62 failures, which.does not include any unreported incidents, 48 involved valves manufactured.by Crosby valve and Gauge and 14 involved valves manufactured by Dresser Industries.
Infor:aation on 11 of 23 valves indicated that they had drifted more than +/.3%
(the remaining 39 valves did not identify the degree of setpoint drift). The cause of setpoint test failure was identified as undetermine6 or as normal setpoint drift due to aging and thermal cycling in 55 of the 62 failures.
Improper test methods and procedures resulted in 7 of the failures. The corrective action taken in 68% of the failures was to adjust the.setpoint within the acceptable range with no additional actions required. 'In 21% of the failed valves, disassembly, inspection, and cleaning did not identify a cause other than normal setpoint drift. The failure history of the PSVs tested for Catawba is consistent with that of the NPRDS data. There have been 9 tests failing to meet the +/- 1% acceptance requirement at Catawba. Of these 9, only1 was greater than +/- 3% (at 3.6% low).
The ability of safety valves to repetitively and consistently perform within the
+/- 1% Code requirement is recognized as an industry wide concern. Recent developments include the effects of loop seal on the ability of safety valve to consistently repeat lifts within the desired setpoint range. -The Westinghouse Owners Group is presently pursuing discussions to address the loop seal concerns in an effort to develop test methods and guidelines to improve setpoint repeatability. Catawba does not utilize loop seals in the psi application and therefore does not have loop seal concerns. Other developments include test I
studies of safety valves to reseat at a desired pressure below:setpoint without l
an extended blowdown period (EPRI Test Report NP-2770-LD and NP-2628-LD).
Results from this study have been used to develop adjustments and tighter l _
contro.ls on the PSV ring settings. The affects of_an extended blowdown on the PSVs are presently being evaluated for the Catawba feedwater line break accident
~
(LER 413/90-001).
The utility industry's input was solicited using the NUCLEAR NETWORK and the NPRDS system. From the response that was obtained.(10 operating plants), it can be concluded that pSVs are not meeting their as-found +/- 1% setpoint acceptance-requirement at a 42% failure rate.
The variance in the information obtained is large with some stations having little trouble _ meeting acceptance requirements.
to stations experiencing a'75% failure rate.
From this information, it is concluded that there are generic issues that need to be addressed concerning pSV setpoint drift.
i
9 UKE POWER COMPANY / CATAWBA NUCLEAR STATION
+ P D.1-C90-0075/Specicl R: port, R;v. 1 l
' Page 5
,.t Catawba, PIR 0-C90-0026 has been issued to address concerns of setpoint drif co the Main Steam Safety Valves (MSSVs). These concerns were originated when the valve manufacturer changed the equipment calibration constants that effected the tolerance by 1.0% to 1.5% of the original setting, which have a,+ 1%
acceptance tolerance of setpoint. PIR 0-C90-0026 identified that setpoint drifts greater than the +/- 1% experienced on 55% of these tests, indicating that the MSSVs and the PSVs have similer setpoint drift concerns.
The ability of the PSVs (or MSSVs) to consistently repeat setpoint lifts at any given pressure can be affected by many uncontrollable variables, including changes in temperatures, pipe loads and stresses, rate of pressure increase, condensate, and seat leakage.
These variables may affect the relief setting in that the test conditions and the actual conditions may not be the same. Also,
'i actual conditions may in themselves vary enough to causo variances in lift settings. Wyle Laboratories utilizes special tests end control procedures and methods to simulate the actual conditions that are err ~'ted to exist et Catawba (Wyle Test Procedure 1028). These efforts have contributed to maintaining relief settings consistently within +/- 3% of setpoint during successive surveillance testing. The single test that was outside of +/- 3% is considered to be an anomaly and is not expected to be recurring. Wyle Laboratories will refurbish this valve and inform MES of any observed deficiencies that may have contributed to its setpoint variance. A Safety Evaluation;will be performed to evaluate the effect of relief setting varying +/- 3% of setpoint dua to the uncontrollable variables that affect valve performance. These efforts will be I
to evaluate the acceptability of as-found (Wyle Laboratories as-received) tests varying +/- 3% of setpoint. However, it is expected that the as-left (Wyle Laboratories as-shipped) test settings will remain within +/- 1% of setpoint.
This will provide a 2% buffer, on both the high and low side, for setpoint drift without involvin'; an operability question.
Thisposition,if.deterkned acceptable, should be documented in the T/Ss, T/S Bases, or T/S Interpretations for Safety /Rollei /alves.
l The inability of safety valves to meet performance of the +/- 1%-test 4
requirement is identified as a recurring problem.
The roccmmended and in-progress evaluation and the Westinghousd Owners Group offorts is expected to address these concerns and implement changes that will improve safety valve performance. A review of the Operating Experience Program (OEP) database revealed no previous events where equipment setpoints varied greater than their i
designed tolerance. However, the two incidents identified in the report meet the Nuclear Safety Assurance detinition of a recurring problem.
This report is being submitted as a Special Report in that setpoint drift greater than the +/- 1% required setting range could occur in subsequent as-found surveillance tests.
The cause of the setpoint drift could not be attributed to a known condition that would recult in the inopelanility of the PSVs. The setpoint drift is attributed to the effect that normr'. plant operation and system conditions can have on the ability of the PSV to consistently relief at the required setting.
Therefore, no T/S violat. ion has occurred.
e
DUKE. POWER-COMPANY / CATAWBA NUCLEAR STATION
~
f PIR 1-C90-0075/Specici R: port, R;v. 1
' Pag 6 6 This report is being submitted as potentially 10CFR Part 21 reportable; the PSV is considered a " basic component"; a drift.in setpoint greater than +/- 1% is considered a " defect"; and the existence of a." substantial safety hazard" is dependent upon the affect this retpoint drift.can have on protecting the Reactor Coolant System for the individual' plant normal operation and analyzed accident conditions. The requirement that the PSVs function within +/- 1% of setpoint-was identified within manufacturer purchase specification CNS 1205.09-00-0001, Pressurizer and Main Steam Safety. Valves..'Setpoint drifts outside +/- 1% of-setpoint due to normal operating conditions is in violation of the purchase specification.
CORRECTIVE ACTION SUBSEQUENT 1)
An NPRDS search was performed to identify an industry wide concern on setpoint drifts in PSVs.
PhANNED 1)
A Safety Evaluation will be performed to address setpoints drifts of
+/- 3% as an acceptable condition on the PSVs.
2)
Station Documents (T/Ss, T/S Bases, or T/S Interpretations) will be revised / developed to incorporate the results of the Safety Evaluation.
3)
The Safety Analysis of this report will be revised to incorporate the results of the safety evaluation.
l
[
4)
For the >3% test results, this valve will be inspected by Wyle Laboratories for any abnormalitica that could have affected the.
setpoint.
SAFETY ANALYSIS l
Variances in the PSV's setting. greater than +/- 1% of setpoint have not been considered in the present safety analysis assumptions.
The PSVs are required to
' function to limit NC System pressure during incidents invol.ving a decrease in q
heat removal by the secondary system, decrease in Reactor coolant flow rate, and anomalies in reactivity and power distribution. A safety evaluation will be-performed to address variances in PSV settings to +/- 3% of setpoint. The results of the safety evaluation will be incorporated into the. Safety Analysis.
of this report.
4 r