LER 95-004-00:on 950407,SG Tube Degradation Occurred Due to Iga/Scc That Caused QA Manual Reportable Limits to Be Reached.Sleeved or Plugged Affected Tubes W/Accepted Industry Repair MethodsML17263B054 |
Person / Time |
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Site: |
Ginna ![Constellation icon.png](/w/images/b/be/Constellation_icon.png) |
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Issue date: |
05/08/1995 |
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From: |
St Martin J ROCHESTER GAS & ELECTRIC CORP. |
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To: |
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Shared Package |
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ML17263B053 |
List: |
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References |
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LER-95-004, LER-95-4, NUDOCS 9505180167 |
Download: ML17263B054 (12) |
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Similar Documents at Ginna |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:RO)
MONTHYEARML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999 Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis ML17265A7621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr ML17265A7531999-09-23023 September 1999 Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped ML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7471999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr ML17265A7291999-07-29029 July 1999 Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7131999-07-22022 July 1999 Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521 ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7211999-07-19019 July 1999 ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp. ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A7661999-06-30030 June 1999 1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr ML17265A7011999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6761999-06-16016 June 1999 Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17265A6681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr ML17265A6651999-05-27027 May 1999 Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001 ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6591999-05-17017 May 1999 Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6381999-05-0707 May 1999 Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG ML17265A6391999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr ML17265A6361999-04-23023 April 1999 Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001 ML17265A6301999-04-18018 April 1999 Rev 1 to Cycle 28 COLR for Re Ginna Npp. ML17265A6251999-04-15015 April 1999 Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment ML17265A6551999-04-0909 April 1999 Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open ML17265A6291999-03-31031 March 1999 Rev 0 to Cycle 28 COLR for Re Ginna Npp. ML17265A6241999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A5661999-03-0101 March 1999 Rev 26 to QA Program for Station Operation. ML17265A5961999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr ML17265A5371999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr ML17265A5951998-12-31031 December 1998 Rg&E 1998 Annual Rept. ML17265A5001998-12-21021 December 1998 Rev 26 to QA Program for Station Operation. ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4761998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4531998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A4291998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr 1999-09-30
[Table view] |
Text
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150 ~ 0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.
LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE IHFORHATION AND RECORDS MANAGEHENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, (See reverse for required nunber of digits/characters for each block) WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAME (1) R.E. Ginna Nuclear Power Plant DOCKET NUMBER (2) PAGE'3) 05000244 'I OF 12 TITLE (4) Steam Generator Tube Degradation Due to IGA/SCC, Causes Quality Assurance Manual Reportable Limits to be Reached EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER HONTH DAY YEAR YEAR HONTH DAY YEAR NUHBER NUMBER 04 07 95 95 --004-- 00 05 08 95 FACILITY NAME DOCKET NUHBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)
MODE (9) 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73. 7'I (b)
POWER 20.405(a )(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 000 LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) X Oj'HER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify in 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Abstract and in Text, below 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) HRC Form 366A)
LICENSEE CONTACT FOR THIS LER (12)
NAME John T~ St. Martin - Technical Assistant TELEPHONE NUMBER (Include Area Code)
(315) 524-4446 COHPLETE ONE LINE FOR EACH COHPONENT FAILURE DESCRIBED IN THIS REPORT (13)
REPORTABLE REPORTABLE CAUSE SYSTEH COHPONENT HANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS TBG H314 SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED HOHTH DAY YEAR YES SUBMISSION (If yes, complete EXPECTED SUBHISSION DATE).
X NO DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
During the 1995 Annual Refueling and Maintenance Outage, subsequent to the eddy current examination performed on both the "A" and "B" Westinghouse Series 44 steam generators, 88 tubes in the "A" steam generator and 122 tubes in the "B" steam generator required corrective action due to tube, sleeve, or plug degradation.
The immediate cause of the event was that the "A" and "B" steam generator tube degradation was in excess of the Ginna Station Quality Assurance Manual reportable limits.
The underlying cause of the tube degradation is a common steam generator problem of a partially rolled tube sheet crevice with recurring intergranular attack/stress corrosion cracking (IGA/SCC) and Primary Water Stress Corrosion Cracking (PWSCC) attack on steam generator tubing. This event is NRC Performance Indicator System Cause Code 5.8.4.3 and NUREG-1022 Cause Code (B).
Corrective action taken was to either sleeve or plug the affected tubes with accepted industry repair methods.
NRC FORM 366 (5-92) 9505180167 950508 PDR ADOCK 05000244 8 PDR
f NRC FORM 3666 U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50 ' HRS.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTONR DC 20555 0001 AND TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 N M 2 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PRE-EVENT PLANT CONDITIONS:
The plant was in the cold/refueling shutdown condition for the 1995 Annual Refueling and Maintenance Outage. The Reactor Coolant System (RCS) was depressurized and RCS temperature was approximately 100 degrees F ~ Steam Generator (S/G) eddy current examination was in progress'I.
DESCRIPTION OF EVENT:
A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
April 7, 1995, 1700 EDST: The number of degraded S/G tubes was known to exceed Ginna Station Quality Assurance (QA)
Manual reportable limits. Event date and time.
April 7, 1995, 1700 EDST: Discovery date and time.
April 10, 1995, 1430 EDST: Oral notification made to the NRC Office of Nuclear Reactor Regulation (NRR) that the number of degraded S/G tubes exceeded QA Manual reportable limits.
April 13, 1995, 1200 EDST: All eddy current programs completed, and the evaluation of the 1995 inservice inspection of S/G tubes completed.
April 15, 1995, 2100 EDST: S/G repairs completed.
April 24, 1995: A Special Report was sent to the USNRC, reporting the number of tubes plugged or sleeved in each S/G.
NRC FORH 366A (5-92)
NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 NRS.
FORWARD COMHENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HAHAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MAHAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR Q TIAL REVIS' R.E. Ginna Nuclear Power Plant 05000244 3 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
B. EVENT:
During the 1995 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the "A" (EMS01A) and "B" (EMS01B) Westinghouse Series 44 design recirculating steam generators.
The purpose of the eddy current examination was to assess any corrosion or mechanical damage that may have occurred during the cycle since the 1994 examination.
The examination was performed by personnel from Rochester Gas and Electric (RG&E) and Asea Brown Boveri - Combustion Engineering (ABB-CE). All personnel were trained and qualified in the eddy current examination method and have been certif ied to a minimum of Level I for data acquisition and Level II for data analysis.
The initial eddy current examinations of the "A" and "B" S/Gs were performed utilizing a standard bobbin coil technique with data acquisition being performed with the EDDYNET Acquisition System. The frequencies selected were 400, 200, 100, and 25 KHz.
Additional eddy current examinations of the "A" and "B" S/Gs were performed utilizing the Zetec 3-coil Motorized Rotating Pancake Coil (MRPC) probe to examine the roll transition region, selected crevices and support plates. The frequencies used for these examinations were 400, 300, 100, and 25 KHz.
Sleeves were inspected using the Zetec "Plus Point" probe, which allows for improved inspection capability of the parent tube behind the sleeve. Since this advanced probe is more sensitive, it also can identify volumetric indications on the sleeve inside surface.
NRC FORH 366A (5-92)
NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVEO BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COHHENTS REGARDING'URDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSIOH, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) DOCKET NUHBER (2) LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M
4 OF 95 004 00 12'EXT (If more space is required, use additional copies of NRC Form 366A) (17)
The inlet or hot leg examination program plan was generated to provide the examination of 100. of each open unsleeved S/G tube from the tube end through the first tube support plate, along with 20% of
,these tubes being selected and examined for their full length [20-:
random sample as recommended in the Electric Power Research Institute (EPRI) guidelines] with the bobbin coil. In addition, '20% of each type of sleeve was examined and the remaining tube examined full length. All Row 1 and Row 2 U-Bend regions were examined with the MRPC between the N6 tube support plate hot side and the N6 tube support plate cold side from the cold leg side.
Results of the above examinations indicated that 88 tubes in the "A" S/G required action (13 new repairs by plugging and 75 new repairs by sleeving). 122 tubes in the "B" S/G required action (31 new repairs by plugging and 91 new repairs by sleeving). Corrective actions were therefore taken for 88 tubes in the "A" S/G and for 122 tubes in the Il ll B S/G On April 7, 1995, at approximately 1700 EDST, with the RCS depressurized and temperature at approximately 100 degrees F, final review of the 1995 S/G eddy current examination results was completed. More than one percent of the total tubes inspected were degraded (imperfections greater than the repair limit). Because of the above, the results of the inspection are considered a reportable event pursuant to 10 CFR 50.73 per Appendix B of the QA Manual.
On April 10, 1995, at approximately 1430 EDST, oral notification was made to the NRC Office of Nuclear Reactor Regulation pursuant to Appendix B of the QA Manual.
On April 24, 1995, a Special Report listing the number of tubes required to be plugged or sleeved in each S/G was reported to the NRC, pursuant to Appendix B of the QA Manual.
NRC FORH 366A (5-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, WASHINGTON, OC 20555-0001, ANO TO THE PAPERWORK REDUCTIOH PROJECT (3150-0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Po~er Plant 05000244 NUM 5 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:
None E. METHOD OF DISCOVERY:
This event was apparent during the review of the "A" and "B" S/G eddy current examination results.
F. OPERATOR ACTION:
The Control Room operators were notified that the number of degraded tubes exceeded the reportable limits of the QA Manual.
The Control Room operators completed the notifications and evaluations required by the A-25.1 (Ginna Station Event Report),
submitted for the event by .the S/G examination and repair supervision.
G. SAFETY SYSTEM RESPONSES:
None III. CAUSE OF EVENT:
A. IMMEDIATE CAUSE:
The immediate cause of the event was the "A" and "B" S/G tube degradation was in excess of the QA Manual reportable limits.
NRC FORH 366A (5-92)
I
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 6 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
B. ROOT CAUSE:
- 1. TUBE DEGRADATION The results of the examination indicate that Intergranular Attack ( IGA) and Intergranular Stress Corrosion Cracking (IGSCC or SCC) continue to be active within the tubesheet crevice region on the inlet side of each S/G. As in the past, IGA/SCC is much more prevalent in the "B" S/G with 80 new crevice indications reported in 1995. In the "A" S/G, 35 new crevice indications were reported in 1995.
In 1994, 42 new crevice indications were reported in the "A" S/G, and 74 new crevice indications were reported in the "B" S/G. Comparison of 1994 and 1995 results does not suggest any significant change in the rate of tube degradation due to IGA/SCC.
The majority of the inlet tubesheet crevice corrosion indications are IGA/SCC of the Mill Annealed Inconel 600 tube material. This form of corrosion is believed to be the result of an alkaline environment forming in the tubesheet crevices. 'his environment has developed over the years as deposits and active species, such as sodium and phosphate, have reacted, changing a neutral or inhibited crevice into the aggressive environment that presently exists.
Along with IGA/SCC in the crevices, Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition continued to be active during the last operating cycle. This mechanism was first addressed in 1989 and this year there were 60 roll transition (PWSCC) indications in the "A" S/G and 32 roll transition (PWSCC) indications in the "B" S/G These numbers ~
include tubes that may have PWSCC in combination with IGA or SCC in the crevice.
This event is NRC Performance Indicator System Cause Code 5.8.4.3, "Maintenance Equipment Failure", and NUREG-1022 Cause Code (B),
"Design, Manufacturing, Construction / Installation." The tube degradation does not meet the NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", definition of a "Maintenance Preventable Functional Failure".
HRC FORM 366A (5-92)
NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.
FORWARD COMMENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSIONS WASHINGTON, DC 20555-0001,'ND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MAHAGEHENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)
YEAR L ON R.E. Ginna Nuclear Power Plant 05000244 7 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
- 2. SLEEVE INDICATIONS Eddy current examination of 515 ABB-CE welded sleeves was performed using a "Plus Point" probe. This examination identified eleven (11) sleeves with inside surface (ID) volumetric indications at the upper weld location.
Subsequent visual examination (VT) of the upper weld joint for these 11 sleeves identified five (5) sleeves with weld "tail-off" indications, three (3) sleeves with pinholes in the weld, two (2) sleeves with blowholes in the weld, and one (1) sleeve with no weld in the upper weld zone.
The eight (8) sleeves with weld tail-off and pinhole indications were left in service since these conditions did not affect either the structural integrity or the leak tightness of the upper weld. One of the sleeves with a blowhole indication was also left in service. The blowhole was located in the upper portion of the weld. Sufficient weld material existed below the blowhole location to provide both sleeve/tube structural integrity and a leak tight weld.
The second blowhole was located in the lower portion of the sleeve weld. Sufficient fusion existed for weld structural integrity. However, the possibility existed for a leak path to develop from the sleeve to the secondary side of the S/G.
Although the sleeve could have been accepted as a leak-limiting sleeve, as a precautionary measure, the sleeve was repaired by plugging.
The one sleeve with no upper weld was identified as being a curved sleeve originally installed in 1990. In discussion with ABB-CE it was concluded'hat the lack of any weld in the sleeve was a result of an equipment problem with the flexible welding tool used to weld curved sleeves. Consequently, all installed curved sleeves were examined either by a review of the Plus Point data or by performing a VT inspection to verify the presence of an upper weld on the sleeve ID. This re-examination process discovered a second curved sleeve with no upper sleeve weld.
NRC FORH 366A (5-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OHB NO. 3150.0104 (5 92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COHMENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Qinna Nuclear Power Plant 05000244 8 OF 12 95 004 00 TEXT (If more space is required, use additions( copies of NRC Form 366A) (17)
Both curved sleeves with missing upper welds were installed in 1990, and both welds were examined and accepted by UT to confirm the existence of weld fusion. The UT examinations were performed by the same UT inspector. As a result of the failure of the UT examination to discover the lack of an upper weld, all of the 113 sleeves examined by the one Level II UT inspector (in 1990) were re-examined by a different ABB-CE Level III UT inspector examination process discovered an additional six (6) sleeves in 1995. This UT re-that had inadequate weld fusion. All eight of the sleeves discovered with either missing or inadequate weld fusion during the 1995 outage were repaired by plugging.
The condition of these 8 ABB-CE sleeves was not identified during the original installation inspection because of the inexperience of the Level II UT inspector who performed the examination. The lack of fusion indications at the weld location resulted in the UT inspector mistaking the sleeve expansion region for a weld.
NRC FORM 366A (5-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS ~
FORWARD COMMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB /714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAME (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)
SEOUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M 9 OF 12 95 004 00 TDT (If more space is required, use additional copies of NRC Form 366A) (17)
IV. ANALYSIS OF EVENT:
This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (Other) and the QA Manual Appendix B which requires that, "If the number of tubes in a generator falling into categories (a) or (b) below exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73." The tube degradation in the "A" and "B" S/Gs exceeded the criterion of (b) which states, "more than 1 percent of the total tubes inspected are degraded (imperfections greater than the repair limit) ".
An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:
There were no operational or safety consequences resulting from the S/G tube degradation in excess of the QA Manual reportable limits because:
The degraded tubes were identified and repaired prior to any significant leakage or S/G tube rupture occurring.
Even assuming a complete severance of a S/G tube at full power, as stated in the RE E. Ginna Nuclear Power Plant Updated Final Safety Analysis Report (Ginna UFSAR) section 15.6.3, (Steam Generator Tube Rupture), the sequence of recovery actions ensures early termination of primary to secondary leakage with or without offsite power available thus limiting offsite radiation doses to within the guidelines of 10 CFR 100.
Sleeve indications do not present any operational safety consequences because there was no major defect in the design, construction, or installation of the ABB-CE welded sleeve which would have resulted in a structural failure of the installed sleeve. Additionally, the lack of adequate fusion during the installation process does not prevent the installed sleeve from functioning as a leak-limiting sleeve.
The parent tube would remain constrained by the tubesheet and the installed sleeve. Therefore, RCS pressure boundary integrity has been maintained.
NRC FORM 366A (5-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150 ~ 0104 (5 92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COMMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON OC 20503.
FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)
SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 UMB 10 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
~ No significant increase in radiation exposure or release to the general public would have occurred. Based on the maximum calculated leakage where a sleeve has a completely missing weld, the resulting primary to secondary leakage is well below the bounding leakage for the existing Ginna S/G tube rupture accident itallcantimes.
analysis'ased on the above, be concluded that the public's health and safety was assured at CORRECTIVE ACTION:
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
Of the 88 tubes repaired in the "A" S/G, 75 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 75 tubes will remain in service. The remaining 13 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 228 tubes in the "A" S/G are currently plugged and 885 tubes are sleeved.
Of the 122 tubes repaired in the "B" S/G, 91 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 91 tubes will remain in service. Of the remaining 31 tubes, 29 tubes were removed from service by plugging both the hot and cold leg tube ends. The other 2 tubes were previously plugged and exhibited minor leakage indications. For these 2 tubes, the tube plugs were removed and new plugs installed.
A total of 343 tubes in the "B" S/G are currently plugged and 1464 tubes are sleeved.
All sleeves that were inspected by the inexperienced Level II UT inspector in 1990 were re-examined in 1995 by a different ABB-CE Level III UT inspector, to verify acceptable weld fusion between the sleeve and the parent tube.
NRC FORH 366A (5-92)
NRC FORH 366A U.S. NUCLEAR REGULATORY COMHISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COMMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714)i U S NUCLEAR REGULATORY COMMISSIONS WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 11 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
The occurrence/presence of IGA, SCC, and PWSCC is a PWR S/G problem. Utilities with susceptible tubing and partially rolled crevices must deal with this recurring attack on S/G tubing.
R.E. Ginna Nuclear Power Plant will continue careful monitoring of both primary RCS and secondary side water chemistry parameters.
These water chemistry parameters will continue to be evaluated against accepted industry guidelines in order to minimize harmful primary and/or secondary side environments.
Degraded S/G tubes shall be sleeved or plugged in accordance with the inservice inspection program and accepted industry repair methods.
To ensure that this lack of experience was limited to the 1990,installation of ABB-CE welded sleeves, the records all of the lead UT inspectors used by ABB-CE from 1986 of to 1993 at Ginna Station were reviewed. This review determined that, excluding the one Level II UT inspector used in 1990, all of the other UT inspectors used at Ginna Station had prior experience with the performance of UT examination of these sleeves' review of ABB-CE's complete sleeve installation and inspection program since 1984 has shown that this Level II UT inspector was not employed by ABB-CE for any other sleeving program at any other utility.
NRC FORM 366A (5-92)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVEO BY OMB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS. INFORHATION COLLECTION REQUEST: 50.0 HRS.
FORWARD COHMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.
FACILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 N M 12 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
VI. ADDITIONAL INFORMATION:
A. FAILED COMPONENTS:
The degraded tubes are Inconel 600 Mill Annealed U-Bend tubes having an outside diameter of 0.875 inches and a nominal wall thickness of 0.050 inches. The tubes were manufactured by Huntington Alloy Company.
B. PREVIOUS LERs ON SIMILAR EVENTS:
A similar LER event historical search was conducted with the following results: The crevice indications are similar to those reported in AO-74-02, A0-75-07, RO-75-013, and LERs76-008, 77-008,78-003, 79-006,79-022, 80-003,81-009, 82-003,82-022, 83-013, 89-001,90-004, 91-005,92-005, 93-002, and 94-006.
C. SPECIAL COMMENTS:
A more detailed final report will be submitted to the NRC, as required by the Ginna QA Manual.
As a note of interest, RG&E has ordered new steam generators for R.E. Ginna Nuclear Power Plant to be installed in 1996.
NRC FORH 366A (5-92)