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Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO LER)
MONTHYEAR3F0399-04, Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease1999-03-10010 March 1999 Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease 3F0498-28, Ro:On 980421,unplanned Release of Effluent Water from Crystal River Nuclear Plant Regeneration Waste Neutralization Tank.Root Cause Analysis Currently Being Conducted.Pipe Isolated for Repairs1998-04-23023 April 1998 Ro:On 980421,unplanned Release of Effluent Water from Crystal River Nuclear Plant Regeneration Waste Neutralization Tank.Root Cause Analysis Currently Being Conducted.Pipe Isolated for Repairs 3F0398-21, Special Rept:On 980302,insp Vendor Notified FPC That Grease Sample for Tendon 51H25 Had Water Content of 14.9% Volume by Weight.Caused Undeterminate.Subject Tendon Was Partially re-greased During Performance of Surveillance1998-03-10010 March 1998 Special Rept:On 980302,insp Vendor Notified FPC That Grease Sample for Tendon 51H25 Had Water Content of 14.9% Volume by Weight.Caused Undeterminate.Subject Tendon Was Partially re-greased During Performance of Surveillance 3F0198-07, Special Rept 97-09:provides Details of Conditions Found Not Meeting Acceptance Criteria During Ongoing Twentieth Year Tendon Surveillance of Containment Post Tensioning Sys. Commitments Encl1998-01-0808 January 1998 Special Rept 97-09:provides Details of Conditions Found Not Meeting Acceptance Criteria During Ongoing Twentieth Year Tendon Surveillance of Containment Post Tensioning Sys. Commitments Encl 3F0198-04, Special Rept 97-08:on 971208,mid Range & High Range Noble Gas Stack Monitors Found to Be Inoperable Greater than Seven Days.Cables to low-medium-high Valve Controllers re-connected Inside Radiation Monitoring Panel1998-01-0303 January 1998 Special Rept 97-08:on 971208,mid Range & High Range Noble Gas Stack Monitors Found to Be Inoperable Greater than Seven Days.Cables to low-medium-high Valve Controllers re-connected Inside Radiation Monitoring Panel 3F1297-06, Special Rept 97-07:re Hoop Tendon 51H26 That Was Found to Have Normalized lift-off Force of More than 10% Below Predicted Value.Out of Tolerance Tendons Were Returned to Proper pre-stress Level.Commitment Attached1997-12-0606 December 1997 Special Rept 97-07:re Hoop Tendon 51H26 That Was Found to Have Normalized lift-off Force of More than 10% Below Predicted Value.Out of Tolerance Tendons Were Returned to Proper pre-stress Level.Commitment Attached 3F1197-14, Special Rept 97-03:on 971021,mid Range & High Range Noble Gas Monitors Found Inoperable Greater than Seven Days.Filter Holder Immediately re-installed1997-11-19019 November 1997 Special Rept 97-03:on 971021,mid Range & High Range Noble Gas Monitors Found Inoperable Greater than Seven Days.Filter Holder Immediately re-installed 3F8097-33, Special Rept 97-02:on 970714,fire Suppression Sys Required by Fire Protection Plan Out of Svc for Greater than Fourteen Days.Two Continuous Fire Watches Put in Place as Compensatory Measures as Required by CR-31997-08-22022 August 1997 Special Rept 97-02:on 970714,fire Suppression Sys Required by Fire Protection Plan Out of Svc for Greater than Fourteen Days.Two Continuous Fire Watches Put in Place as Compensatory Measures as Required by CR-3 3F1096-22, Special Rept:On 960902,unit Shutdown Due to Leak in Turbine Lube Oil Sys.Issue Will Be Resolved Before Startup from Current Outage1996-10-28028 October 1996 Special Rept:On 960902,unit Shutdown Due to Leak in Turbine Lube Oil Sys.Issue Will Be Resolved Before Startup from Current Outage 3F0896-25, Special Rept:On 960712,declared Seismic Monitoring Instrumentation Inoperable for More than 30 Days.Caused by Triaxial Peak Accelographs Failures.Instruments Will Be Omitted from FSAR Section 2.5.4.41996-08-28028 August 1996 Special Rept:On 960712,declared Seismic Monitoring Instrumentation Inoperable for More than 30 Days.Caused by Triaxial Peak Accelographs Failures.Instruments Will Be Omitted from FSAR Section 2.5.4.4 3F0496-30, Special Rept 96-02:on 960313,fire Detection Zones Required by Fire Protection Plan Out of Svc for Greater than Fourteen Days.Hourly Fire Watch Established1996-04-26026 April 1996 Special Rept 96-02:on 960313,fire Detection Zones Required by Fire Protection Plan Out of Svc for Greater than Fourteen Days.Hourly Fire Watch Established 3F0396-21, Special Rept 96-01:on 960218,RM-A1 Taken Out of Svc Due to Failure of Detector to Respond to Check Source.Replacement Detector from Stores & Bench Calibrate.Expects to Return RM-A1 to Svc Prior to Restart from Refueling Outage1996-03-26026 March 1996 Special Rept 96-01:on 960218,RM-A1 Taken Out of Svc Due to Failure of Detector to Respond to Check Source.Replacement Detector from Stores & Bench Calibrate.Expects to Return RM-A1 to Svc Prior to Restart from Refueling Outage 3F0895-16, Ro:On 950713,1 H non-emergency Rept Made Re Condition Suspected to Be Outside Design Basis of Plant.Based on Conclusions of Design Basis Evaluation,Rescinds 1 H non-emergency Rept (NRC Event 29062)1995-08-11011 August 1995 Ro:On 950713,1 H non-emergency Rept Made Re Condition Suspected to Be Outside Design Basis of Plant.Based on Conclusions of Design Basis Evaluation,Rescinds 1 H non-emergency Rept (NRC Event 29062) 3F0895-15, Special Rept 95-01:on 950630,ODCM Required Waste Gas Analyzer (WGDA-1) Declared Inoperable & Unavailable for Greater than Seven Days Due to Deficiency in Microprocessor. Microprocessor Replaced & WGDA-1 Returned to Operation1995-08-11011 August 1995 Special Rept 95-01:on 950630,ODCM Required Waste Gas Analyzer (WGDA-1) Declared Inoperable & Unavailable for Greater than Seven Days Due to Deficiency in Microprocessor. Microprocessor Replaced & WGDA-1 Returned to Operation 3F1094-07, Special Rept:On 940919,determined That Two Transmitters Which Comprise Rv Level Indication Sys Portion of Rc Inventory Tracking Sys Not Functioning Normally Due to Line Blockage in Common Tubing.Sys Restoration outage-dependent1994-10-10010 October 1994 Special Rept:On 940919,determined That Two Transmitters Which Comprise Rv Level Indication Sys Portion of Rc Inventory Tracking Sys Not Functioning Normally Due to Line Blockage in Common Tubing.Sys Restoration outage-dependent 3F1293-07, Special Rept 93-03:on 931124,ODCM Required Radiation Monitor Taken Out of Svc Due to Failure to Pass Check Source Functional Test & Unavailable for Greater than 7 Days. Subj Monitor Restored to Normal Operation on 9312061993-12-23023 December 1993 Special Rept 93-03:on 931124,ODCM Required Radiation Monitor Taken Out of Svc Due to Failure to Pass Check Source Functional Test & Unavailable for Greater than 7 Days. Subj Monitor Restored to Normal Operation on 931206 3F1093-17, Special Rept 93-02:on 931011,primary Meteorlogical Sys Taken Out of Svc for Performance of TS Surveillance 4.3.3.4 & TS 3.3.3.4 Entered Since Backup Sys Out of Svc.Surveillance Performed & Sys Returned to Operation on 9310191993-10-27027 October 1993 Special Rept 93-02:on 931011,primary Meteorlogical Sys Taken Out of Svc for Performance of TS Surveillance 4.3.3.4 & TS 3.3.3.4 Entered Since Backup Sys Out of Svc.Surveillance Performed & Sys Returned to Operation on 931019 3F0992-18, Special Rept 92-003:on 920815,auxiliary Bldg high-range Noble Gas Monitor Failed Calibr & Declared Inoperable for More than 7 Days.Automatic Isotopic Monitoring Sys Available for Use.Amplifier Components & detector-sensor Replaced1992-09-24024 September 1992 Special Rept 92-003:on 920815,auxiliary Bldg high-range Noble Gas Monitor Failed Calibr & Declared Inoperable for More than 7 Days.Automatic Isotopic Monitoring Sys Available for Use.Amplifier Components & detector-sensor Replaced 3F0692-10, Special Rept 92-02:on 920514,waste Gas Decay Tank Hydrogen & Oxygen Monitoring Channels Removed from Svc & Not Returned to Operable Status within 14 Days.Caused by Need to Facilitate Maint.Ts Amend Will Be in Place by Sept 19921992-06-12012 June 1992 Special Rept 92-02:on 920514,waste Gas Decay Tank Hydrogen & Oxygen Monitoring Channels Removed from Svc & Not Returned to Operable Status within 14 Days.Caused by Need to Facilitate Maint.Ts Amend Will Be in Place by Sept 1992 3F0592-17, Special Rept 92-01,on 920506,reactor Bldg Purge Exhaust Duct Monitor Was Taken Out of Svc.Caused by Inability to Calibrate RM-A1 Noble Gas Activity.Monitor mid-range Channel Was Recalibrated1992-05-26026 May 1992 Special Rept 92-01,on 920506,reactor Bldg Purge Exhaust Duct Monitor Was Taken Out of Svc.Caused by Inability to Calibrate RM-A1 Noble Gas Activity.Monitor mid-range Channel Was Recalibrated 3F0591-02, Special Rept 91-001:on 910320,waste Gas Analyzer Removed from Svc to Allow Isolation of Waste Gas Compressor from Maint.Moisture Discovered in Sample Tubing.Work Request to Correct Moisture Intrusion in Analyzer Sample Lines Written1991-05-0101 May 1991 Special Rept 91-001:on 910320,waste Gas Analyzer Removed from Svc to Allow Isolation of Waste Gas Compressor from Maint.Moisture Discovered in Sample Tubing.Work Request to Correct Moisture Intrusion in Analyzer Sample Lines Written 3F1190-13, Special Rept 90-04:on 901015,portion of Mounting Platform for Triaxial Peak Accelograph on Piping on Top of Steam Generator Melted1990-11-21021 November 1990 Special Rept 90-04:on 901015,portion of Mounting Platform for Triaxial Peak Accelograph on Piping on Top of Steam Generator Melted 3F0989-08, Special Rept 89-001:on 890826,dose Equivalent Iodine Dei of RCS Samples Exceeded Tech Specs Limit.Sampling & Analysis of RCS Continued at 4 H Intervals Until Dei Decrease Below Tech Spec Limit.Dei Analysis Results Presented1989-09-21021 September 1989 Special Rept 89-001:on 890826,dose Equivalent Iodine Dei of RCS Samples Exceeded Tech Specs Limit.Sampling & Analysis of RCS Continued at 4 H Intervals Until Dei Decrease Below Tech Spec Limit.Dei Analysis Results Presented 3F0389-01, Ro:On 890118,reactor Coolant Pump 1A Failed.Cause Not Discussed.Sequence of Events Encl1989-03-0101 March 1989 Ro:On 890118,reactor Coolant Pump 1A Failed.Cause Not Discussed.Sequence of Events Encl 3F1188-01, Special Rept 88-001:on 881009,RCS Sample Obtained & Analyzed for Iodine Activity Indicated Dose Equivalent Iodine (Dei) Exceeding Tech Spec Limit.Chemists Continued Sampling & Analysis of RCS at Intervals Until Dei Decreased1988-11-0101 November 1988 Special Rept 88-001:on 881009,RCS Sample Obtained & Analyzed for Iodine Activity Indicated Dose Equivalent Iodine (Dei) Exceeding Tech Spec Limit.Chemists Continued Sampling & Analysis of RCS at Intervals Until Dei Decreased 3F0688-19, Ro:On 880527,emergency Feedwater Piping Near Flow Control Valve Found to Be Warmer than Normal.Cause Not Stated.Water Added to Reseat & to Cool Emergency Feedwater Line & Valve Closed to Prevent Inadvertent Manual Addition of Water1988-06-28028 June 1988 Ro:On 880527,emergency Feedwater Piping Near Flow Control Valve Found to Be Warmer than Normal.Cause Not Stated.Water Added to Reseat & to Cool Emergency Feedwater Line & Valve Closed to Prevent Inadvertent Manual Addition of Water 3F0388-02, Special Rept 87-05-01:two Steam Generator Tubes Found Defective,Having Indications Greater than 40% Through Wall & One Tube Identified as Degraded W/Indication of 38% Through Wall.Tubes Removed from Svc by Plugging.Addl Info Encl1988-03-21021 March 1988 Special Rept 87-05-01:two Steam Generator Tubes Found Defective,Having Indications Greater than 40% Through Wall & One Tube Identified as Degraded W/Indication of 38% Through Wall.Tubes Removed from Svc by Plugging.Addl Info Encl 3F0188-19, Special Rept 87-05:during Refuel IV Outage,Steam Generator Eddy Current Testing Revealed Defective tubes.Pre-outage Planning Developed 6% Random Sampling Insp Plan That Would Be Completed on a Steam Generator1988-01-20020 January 1988 Special Rept 87-05:during Refuel IV Outage,Steam Generator Eddy Current Testing Revealed Defective tubes.Pre-outage Planning Developed 6% Random Sampling Insp Plan That Would Be Completed on a Steam Generator 3F1187-09, Special Rept 87-04:Tech Spec 3.3.3.10 Violated.Caused by Removal of Waste Gas Decay Tank Hydrogen & Oxygen Monitors from Svc & Not Returned to Operable Status within 14 Days. Monitors Will Remain Out of Svc Until Sys in Use1987-11-18018 November 1987 Special Rept 87-04:Tech Spec 3.3.3.10 Violated.Caused by Removal of Waste Gas Decay Tank Hydrogen & Oxygen Monitors from Svc & Not Returned to Operable Status within 14 Days. Monitors Will Remain Out of Svc Until Sys in Use ML20236L4161987-11-0505 November 1987 Special Rept 87-03:on 870927,during Channel Check of Triaxial Peak,Monitor SI-005-MEI Found Inoperable.Cause Undetermined.Replacement Monitor to Be Installed After Refueling & Engineering Design Evaluation to Be Made 3F0887-01, Special Rept 87-02:on 870702,reactor Trip Occurred from 88% Rated Thermal Power.Primary Coolant Sample Obtained & Analyzed for Iodine Activity,Per Tech Spec 4.4.8.Reactor Power History,Fuel Burnup & Time Duration Provided1987-08-0303 August 1987 Special Rept 87-02:on 870702,reactor Trip Occurred from 88% Rated Thermal Power.Primary Coolant Sample Obtained & Analyzed for Iodine Activity,Per Tech Spec 4.4.8.Reactor Power History,Fuel Burnup & Time Duration Provided 3F0687-06, Special Rept 87-01:on 870519,auxiliary Bldg & Fuel Handling Area Exhaust Duct mid-range Noble Gas Monitor Inoperable for Seven or More Days.Caused by Alarm Circuit Repair Altering Calibr of Monitor,Rendering Instrument Inoperable1987-06-0909 June 1987 Special Rept 87-01:on 870519,auxiliary Bldg & Fuel Handling Area Exhaust Duct mid-range Noble Gas Monitor Inoperable for Seven or More Days.Caused by Alarm Circuit Repair Altering Calibr of Monitor,Rendering Instrument Inoperable 3F0287-17, Ro:On 870121,during Performance of Surveillance Procedure SP-110,CRD a Ac Breaker Failed to Trip.Caused by Malfunction of Undervoltage Trip Coil Device.Breaker Replaced & Successfully Tested1987-02-23023 February 1987 Ro:On 870121,during Performance of Surveillance Procedure SP-110,CRD a Ac Breaker Failed to Trip.Caused by Malfunction of Undervoltage Trip Coil Device.Breaker Replaced & Successfully Tested 3F0287-21, Ro:On 870112,6-month Interval for Containment Air Lock Type B Tests Found Exceeded Due to Inconsistencies Between Tech Specs & App J.Caused by Failure to Recognize Dual Sources of Requirements.Test Schedule Revised & Tech Spec Changed1987-02-23023 February 1987 Ro:On 870112,6-month Interval for Containment Air Lock Type B Tests Found Exceeded Due to Inconsistencies Between Tech Specs & App J.Caused by Failure to Recognize Dual Sources of Requirements.Test Schedule Revised & Tech Spec Changed 3F0187-07, Special Rept 86-01:on 861223,during Startup from Mode 5,unit Entered Mode 4 W/Noble Gas Activity Monitor & Fuel Handling Bldg Area Exhaust Duct Inoperable.Caused by Poor Response of Installed Check Source.Instrument Recalibr1987-01-0909 January 1987 Special Rept 86-01:on 861223,during Startup from Mode 5,unit Entered Mode 4 W/Noble Gas Activity Monitor & Fuel Handling Bldg Area Exhaust Duct Inoperable.Caused by Poor Response of Installed Check Source.Instrument Recalibr ML20211E0731986-09-26026 September 1986 Unplanned Operating Event Rept 85-6, Manual Reactor Trip Following MSIV Closure Due to Inadvertent Emergency Feedwater Actuation,851009 3F0786-22, Ro:On 860629,reactor Bldg Vented to Relieve Long Term Pressure Buildup.Initially Reported on 860630.Draft LER Under Mgt Review to Determine Reporting Requirements.Next Rept Will Be Submitted by 8608081986-07-29029 July 1986 Ro:On 860629,reactor Bldg Vented to Relieve Long Term Pressure Buildup.Initially Reported on 860630.Draft LER Under Mgt Review to Determine Reporting Requirements.Next Rept Will Be Submitted by 860808 3F0386-13, Suppl 2 to Special Rept 85-03:on 850807,intermediate- & high-range Channels of Gaseous Release Monitors RM-A1 & RM-A2 Not Calibr.Mod to Achieve Overlap & Calibr of Intermediate Channels Completed1986-03-31031 March 1986 Suppl 2 to Special Rept 85-03:on 850807,intermediate- & high-range Channels of Gaseous Release Monitors RM-A1 & RM-A2 Not Calibr.Mod to Achieve Overlap & Calibr of Intermediate Channels Completed 3F0286-08, RO:86-002-00:on 860110,incident Occurred in Seawater Intake Area Resulting in Deaths of Two Divers.All Seawater Pumps Shut Off for 2 H 45 Minutes.Ler Will Be Sent by 8602211986-02-10010 February 1986 RO:86-002-00:on 860110,incident Occurred in Seawater Intake Area Resulting in Deaths of Two Divers.All Seawater Pumps Shut Off for 2 H 45 Minutes.Ler Will Be Sent by 860221 3F0885-10, Special Rept 85-03:on 850807,discovered That mid- & high- Range Channels of Gaseous Effluent Release Monitors Out of Calibr.Caused by Delays in Contractual Arrangements for Supplies.Action Taken to Obtain Matls.Part 21 Related1985-08-21021 August 1985 Special Rept 85-03:on 850807,discovered That mid- & high- Range Channels of Gaseous Effluent Release Monitors Out of Calibr.Caused by Delays in Contractual Arrangements for Supplies.Action Taken to Obtain Matls.Part 21 Related 3F0785-29, Revised Special Rept 85-01 Re Inoperability of Three Triaxial Peak Accelographs.Installation of Reactor Vessel Head Device Delayed Until 850714 Due to Unforeseen Activities1985-07-24024 July 1985 Revised Special Rept 85-01 Re Inoperability of Three Triaxial Peak Accelographs.Installation of Reactor Vessel Head Device Delayed Until 850714 Due to Unforeseen Activities 3F0785-30, Suppl 1 to Special Rept 85-02:Halon Portion of Fire Suppression Sys Inoperable.Addl Work Required Due to Unanticipated Difficulties W/Asbestos Matls in Areas Being Modified.Operability Not Restored by 8507011985-07-24024 July 1985 Suppl 1 to Special Rept 85-02:Halon Portion of Fire Suppression Sys Inoperable.Addl Work Required Due to Unanticipated Difficulties W/Asbestos Matls in Areas Being Modified.Operability Not Restored by 850701 3F0585-14, Special Rept 85-02:on 850408,during Refueling & Mod Outage, Halon Portion of Fire Suppression Sys for Cable Spreading Room Taken Out of Svc in Support of Various Mods to Cable Spreading Room.Fire Watches Immediately Established1985-05-22022 May 1985 Special Rept 85-02:on 850408,during Refueling & Mod Outage, Halon Portion of Fire Suppression Sys for Cable Spreading Room Taken Out of Svc in Support of Various Mods to Cable Spreading Room.Fire Watches Immediately Established 3F0485-20, Special Rept 85-01:on 850317,during Performance of Refueling Interval Surveillance,All Three Triaxial Peak Accelographs Required by Tech Specs Discovered Inoperable.Cause Undetermined.Accelographs Will Be Repaired1985-04-26026 April 1985 Special Rept 85-01:on 850317,during Performance of Refueling Interval Surveillance,All Three Triaxial Peak Accelographs Required by Tech Specs Discovered Inoperable.Cause Undetermined.Accelographs Will Be Repaired 3F1284-04, Special Rept SR-84-5:on 841029,Halon Portion of Fire Suppression Sys for Cable Spreading Room Taken Out of Svc for Mod to Cable Spreading Room.Sys Will Be Out of Svc Periodically During Mod for Welding in Cable Spreading Room1984-12-0707 December 1984 Special Rept SR-84-5:on 841029,Halon Portion of Fire Suppression Sys for Cable Spreading Room Taken Out of Svc for Mod to Cable Spreading Room.Sys Will Be Out of Svc Periodically During Mod for Welding in Cable Spreading Room 3F1084-05, Special Rept 84-04:on 840919,fire Suppression Water Sys Placed in Degraded Mode of Operation to Support Sys Mod. Caused by Redundant Loop Flow Path Allowing Only One Fire Svc Sys Header to Remain in Svc.Loop Removed1984-10-0303 October 1984 Special Rept 84-04:on 840919,fire Suppression Water Sys Placed in Degraded Mode of Operation to Support Sys Mod. Caused by Redundant Loop Flow Path Allowing Only One Fire Svc Sys Header to Remain in Svc.Loop Removed 3F0784-16, Special Rept 84-03:on 840424,false Low & High Pressure Injection Occurred When Channel 2 Low Pressure Bistable Inadvertently Actuated.Cause Not Stated.Bistable Replaced. Plant Stabilized W/Injection of Borated Water Into RCS1984-07-23023 July 1984 Special Rept 84-03:on 840424,false Low & High Pressure Injection Occurred When Channel 2 Low Pressure Bistable Inadvertently Actuated.Cause Not Stated.Bistable Replaced. Plant Stabilized W/Injection of Borated Water Into RCS 3F0684-09, Special Rept 84-02:on 840312,engineered Safeguards High Pressure Injection Initiated.Caused by Spurious Actuation of Channel 2 Train B.Equipment Replaced & Borated Water Injected1984-06-13013 June 1984 Special Rept 84-02:on 840312,engineered Safeguards High Pressure Injection Initiated.Caused by Spurious Actuation of Channel 2 Train B.Equipment Replaced & Borated Water Injected 3F0184-22, RO Special Rept 84-01:on 840114,leak Found in Fire Suppression Water Sys,Rendering Sys Inoperable.Caused by Compliance W/Tech Spec Limiting Condition for Operation.Sys Returned to Operable Status on 8401151984-01-23023 January 1984 RO Special Rept 84-01:on 840114,leak Found in Fire Suppression Water Sys,Rendering Sys Inoperable.Caused by Compliance W/Tech Spec Limiting Condition for Operation.Sys Returned to Operable Status on 840115 3F1283-06, RO 83-050:on 830719,mode Ascension Occurred (Mode 5 to 4) Prior to Completing Required Surveillance.Occurrence Analysis Continuing.Ler Will Be Submitted by 8312221983-12-0505 December 1983 RO 83-050:on 830719,mode Ascension Occurred (Mode 5 to 4) Prior to Completing Required Surveillance.Occurrence Analysis Continuing.Ler Will Be Submitted by 831222 1999-03-10
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair 3F1099-19, Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated1999-10-13013 October 1999 Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated ML20217B0931999-10-0606 October 1999 Part 21 Rept Re Damaged Safety Grade Electrical Cabling Found in Supply on 990831.Damage Created During Cabling Process While Combining Three Conductors Just Prior to Closing.Vendor Notified of Reporting of Issue ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212E9031999-09-30030 September 1999 FPC Crystal River Unit 3 Plant Reference Simulator Four Year Simulator Certification Rept Sept 1995-Sept 1999 3F1099-02, Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20211L1321999-08-31031 August 1999 EAL Basis Document 3F0999-02, Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With ML20212C1501999-08-31031 August 1999 Non-proprietary Version of Rev 0 to Crystal River Unit 3 Enhanced Spent Fuel Storage Engineering Input to LAR Number 239 ML20211B7291999-08-16016 August 1999 Rev 2 to Cycle 11 Colr ML20210P1111999-08-0505 August 1999 SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3 ML20210U5341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Crystal River,Unit 3 ML20209F5601999-07-31031 July 1999 EAL Basis Document, for Jul 1999 3F0799-01, Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With ML20210U5411999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Crystal River,Unit 3 3F0699-07, Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With ML20210U5601999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Crystal River,Unit 3 ML20195C6271999-05-28028 May 1999 Non-proprietary Rev 0 to Addendum to Topical Rept BAW-2346P, CR-3 Plant Specific MSLB Leak Rates ML20196L2031999-05-19019 May 1999 Non-proprietary Rev 0 to BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs 3F0599-04, Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With ML20210U5631999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Crystal River,Unit 3 3F0499-04, Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With ML20204D9661999-03-31031 March 1999 Non-proprietary Rev 1,Addendum a to BAW-2342, OTSG Repair Roll Qualification Rept 3F0399-04, Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease1999-03-10010 March 1999 Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease 3F0399-03, Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With ML20203A4381999-02-0303 February 1999 Safety Evaluation Supporting EAL Changes for License DPR-72, Per 10CFR50.47(b)(4) & App E to 10CFR50 ML20206E9891998-12-31031 December 1998 Kissimmee Utility Authority 1998 Annual Rept ML20206E9021998-12-31031 December 1998 Florida Progress Corp 1998 Annual Rept ML20206E9701998-12-31031 December 1998 Ouc 1998 Annual Rept. with Financial Statements from Seminole Electric Cooperative,Inc 3F0199-05, Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With ML20206E9261998-12-31031 December 1998 Gainesville Regional Utilities 1998 Annual Rept 3F1298-13, Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With 3F1198-05, Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With ML20155F4071998-10-31031 October 1998 Rev 2 to Pressure/Temp Limits Rept ML20155J2701998-10-28028 October 1998 Second Ten-Year Insp Interval Closeout Summary Rept 3F1098-06, Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With ML20206E9461998-09-30030 September 1998 Utilities Commission City of New Smyrna Beach,Fl Comprehensive Annual Financial Rept Sept 30,1998 & 1997 ML20206E9561998-09-30030 September 1998 City of Ocala Comprehensive Annual Financial Rept for Yr Ended 980930 ML20206E9101998-09-30030 September 1998 City of Bushnell Fl Comprehensive Annual Financial Rept for Fiscal Yr Ended 980930 ML20206E9811998-09-30030 September 1998 City of Tallahassee,Fl Comprehensive Annual Financial Rept for Yr Ended 980930 ML20195E3121998-09-30030 September 1998 Comprehensive Annual Financial Rept for City of Leesburg,Fl Fiscal Yr Ended 980930 3F0998-07, Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With ML20236W6501998-07-31031 July 1998 Emergency Action Level Basis Document 3F0898-02, Monthly Operating Rept for Jul 1998 for Crystal River,Unit 11998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Crystal River,Unit 1 ML20236V8801998-07-30030 July 1998 Control Room Habitability Rept 3F0798-01, Monthly Operating Rept for June 1998 for Crystal River Unit 31998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Crystal River Unit 3 ML20236Q4611998-06-30030 June 1998 SER for Crystal River Power Station,Unit 3,individual Plant Exam (Ipe).Concludes That Plant IPE Complete Re Info Requested by GL 88-20 & IPE Results Reasonable Given Plant Design,Operation & History 3F0698-02, Monthly Operating Rept for May 1998 for Crystal River Unit 31998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Crystal River Unit 3 1999-09-30
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O Florida Power CORPORATION UM*
October 28, 1996 3F1096-22 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001
Subject:
Crystal River Unit 3 Forced Outage
Dear Sir:
On September 2,1996, Florida Power Corporation (FPC) shut down the Crystal River Unit 3 (CR-3) nuclear plant due to a leak in the turbine lube oil system. During this forced outage, FPC determined that a modification had been made to the plant during the Spring, 1996 Refuel 10 outage which created an Unreviewed Safety Question (USQ) regarding emergency diesel generator (EDG) loading. This USQ involved a reduction in the margin of safety described in portions of the Technical Specification Bases.
On October 4,1996, while still shut down, FPC was preparing a submittal to request NRC approval of a license amendment to change the affected EDG Technical Specification Bases when additional questions arose regarding the change to the emergency feedwater (EFW) system which created the diesel loading USQ. These questions involved failure modes with the EFW system which needed to be evaluated to ensure the system could perform its safety function and reliance on the turbine-driven, "B" train emergency feedwater pump for "A" train EDG load management. Due to the EFW/EDG issues, and some other design-related issues, FPC management made a decision to keep CR-3 shut down until these issues are adequately addressed. The purpose of this letter is to inform the NRC of our plans to address these issues prior to restarting the plant.
t 9610310071 961028 !
PDR ADOCK 05000302 S PDR I CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power Line St . Crystal River, Florida 344284708 . (352) 795-6486 A Florida Progress Company
U. S. Nuclear Regulatory Commission 3F1096.-22 Page 2 of 7 The issues described in the attached list were identified through a review conducted by a multi-discipline team involved in reviewing the Emergency Operating Procedures (E0Ps) and through design reviews by the engineering organization. The list was reviewed by CR-3 senior management and the items are considered necessary to ensure safety system operability or to increase design margins. Each issue has been documented in the CR-3 corrective action system and will be tracked to closure. Several of the issues have been determined to be reportable and Licensee Event Reports are being processed.
FPC will ensure the safety systems in question are capable of performing their design basis functions prior to restart from this outage. As an added level of assurance, FPC will be establishing an internal restart panel which will function similar to an NRC restart panel using NRC Inspection Manual 0350 as a guideline for conducting the restart readiness review. Upon completion of the work to resolve the issues, the panel will conduct a final review to confirm that all issues have been resolved adequately. When satisfied, restart of the unit will be recommended to the Senior Vice President, Nuclear Operations. In addition, ,
the Nuclear General Review Committee (NGRC) will conduct an independent review '
prior to restart.
Project teams or individual lead responsibility have been established for each l issue to support the design, licensing and installation activities necessary to ;
complete the outage work scope. Final resolutions for some of the issues on the I list have not yet been determined. Other resolutions require relatively long lead procurement activities. Therefore, an integrated outage schedule is not l available at this time. However, we expect the unit to remain shutdown until at least mid-January, 1997. This will also likely move our next refueling outage, .
Refuel 11, to the fall of 1998 rather than the spring of 1998, as currently I scheduled. The NRC will be kept abreast of the schedule and progress on these l issues as the outage continues.
l Sincerely, l
. M. Beard, Jr.
Senior Vice President Nuclear Operations '
PMB/BG Attachment xc: Regional Administrator, Region 11 Senior Resident Inspector NRR Project Manager
~i U. S. Nuclear Regulatory Commiusion i 3F1096 Attach' ment-
- l. Page 3 of 7 CR-3 Design Margin Improvement Outage Scope of Work
- 1. Hiah Pressure Injection (HPI) Pumo Recirculation to the Makeup Tank
[
Concern: The HPI pumps draw suction frem the Borated Water Storage Tank (BWST) during the initial phase of emergency core cooling system
- - -(ECCS) injection. Once BWST level has reached a pre-determined
, level, suction is switched to the reactor building sump with the HPI f pumps taking suction from the discharge of the low pressure injection (LPI) pumps (piggyback operation). During piggyback
! operation, LPI pump discharge pressure keeps the check valve in the 1, suction line from the makeup tank (MUT) to the HPI pumps closed j (MUV-65). During long term small break LOCA (SBLOCA) cooling, HPI flow may require throttling due to lower required ECCS flow. If
! throttling continues, procedures will eventually direct the
!. operators to increase total HPI pump flow by opening the HPI .
recirculation valves at a pre-determined flow rate to divert some l flow to the MUT. Since no flow is exiting the MUT, the tank could j j fill up with recirculation flow and lift the relief valves, dumping i
! fluid onto the auxiliary building floor. This would result in the transfer of RB sump fluid to the auxiliary building sump, which reduces the amount of water available in the RB sump from which the J LPI and reactor building spray pumps take suction during the later ,
stages of core and containment cooling. This could also create a . !
release path for post accident radioactive fluid outside l containment.
Resolution: FPC is consulting with Framatome Technologies, Inc. (FTI) to confirm 1 whether the scenario is valid and within the CR-3 design basis.
Although the resolution of this issue is still undetermined at this time, preliminary indications are that opening of a high point vent valve may preclude the need to open the HPI recirculation valves in the SBLOCA scenarios of concern.
Schedule: This issue will be resolved prior to startup from the current outage.
l
- 2. HPI System Modifications to Improve SBLOCA Maraiqi ,
i Concern: The CR-3 HPI system currently meets all design and licensing basis functional requirements. However, the CR-3 configuration is not consistent with the designs at other Babcock and Wilcox (B&W) pl ants . As.a result,-HPI. minimum and maximum flow limits are more i restrictive and peak cladding temperatures for certain SBLOCA scenarios are higher. In addition, the reduced system design margin has created the need for several snanual operator actions to ensure adequate core cooling. FPC intends to reduce the operator burden created by these actions and the system margin deficit through hardware modifications. These modifications would also make the CR-3 HPI system design more like other B&W plants.
- l. .
U. S. Nuclear Regulatory Commission 3F1096-22 httachment
- Page 4 of 7
! Resolution: At this time, the following modifications are being considered:
l a. Installing cavitating venturis to limit flow through any '
single injection leg due to a postulated break in that leg.
- b. Installing cross-tie piping downstream of the HPI injection control valves to deliver increased core cooling flow should l a failure prevent one or more of the injection valves from opening. ;
, c. Modifying the normal makeup line to ensure automatic isolation l_ occurs upon ES actuation to eliminate the operator action now 3 required to perform this function. This involves modifying l the power supply to the existing isolation valve (MUV-27) and adding another isolation valve powered from the opposite train in series with MUV-27. (Note: the proposed installation of the cavitating venturis could preclude the need for this
, modification).
l Schedule: Since the HPI system is fully capable its design
- function, these modifications are not of meetingne'cessary to considered l complete during the current outage. However, FPC is developing the 4 l design package., and determining whether equipment can be procured in l l a time frame to install in the current outage given the schedules I l for other activities.
I
- 3. LPI Pumo Mission Time ,
Concern: During the IPAP inspection, an issue was raised regarding the need to establish flow through the decay heat remova' (JH) drop line to ,
the decay heat removal (LPI) pumps as part of small break LOCA 1 l mitigation. CR-3 has two redundant, independent LPI trains which I j can take section from the RB sump during lcnq term recirculation !
- core cooling. However, certain small break LOCAs could result in i long-lasting, elevated RCS pressures such that the LPI pumps would i have to operate in the piggyback mode at low flow rates for an extended period of time. As that period of time approaches the current low flow mission time for the LPI pumps, plant procedures direct the operators to trip one pump and open the DH drop line valves to the RB sump to provide additional flow through the remaining running LPI pump. There is only one DH drop line at CR-3 (and many other pressurized water reactors) which has three motor-operated valves in series. Failure of any one of the drop line valves to open would prevent now through the line. If the DH drop line was necessary to ful' . . the ECCS long term core cooling function for small break LOCA mitigation, this would violate the single failure design criterion. j i Msolution: The concern described above is time-dependent. If the time frame is long enough aftor the event, opening of the DH drop line could be considered a long-term recovery action as opposed to an emergency core cooling function. FPC considers the long term recovery phase
- beyond the time frame implied by the regulations where applying the single failure design criterion is necessary. At the time of the l
4 4
l, .
~
4 -
V. S. Nuclear Regulatory Commission !
3F.1096-22 Attachment Page 5 of 7 i IPAP inspection, the low flow mission time for the LPI pumps was 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which was questionable from an ECCS versus long term recovery perspective. FPC is currently low-flow testing a pump which is i l identical to the CR-3 LPI pumps. The test flow rate is '
j approximately 100 gallons per minute (gpm). The design flow rate of ,
the LPI pumps is 3000 gpm. The results of this test are expected to l prove that the pumps could run for an extended poriod at very low ;
flows without damage. If the test is successful, procedures will be i revised to characterize opening the DH drop line in this scenario as ;
a long term recovery action rather than an ECCS function. ;
Schedule: This issue will be resolved before startup from the current outage. !
As of 3:30 p.m. on October 25, 1996, the pump Md completed 18 days i of continuous low-flow testing with no performance (head curve) l degiadation, no mechanical seal leakage, no indication of unexpected i bearing wear, and all vibration parameters stable and well below the l action levels specified in the surveillance procedure. The testing '
! is continuing beyond 18 days.
- 4. Reactor Buildino Soray Pump 18 NPSH Concern: During the long term recirculation phase of core and containment cooling, the reactor building spray pumps (BSPs) take. suction from the reactor building sump. Calculations have shown BSP-1B to have I little margin between required and available net positive suction head (NPSH) during this phase of operaticn. A recent revision of ;
the calculation shows the margin to be approximately one foot of water. It is desired tc W rease this margin.
l Resolution: FPC currently plans to conduct factory testing and/or modify the pump impeller to improve the margin between required and available NPSH. !
l Schedule: This issue will be resolved before startup from the current outage.
I
- 5. Emeroency Feedwater System Uparades and Diesel Generator Load Impact !
Concern 5.1: The CR-3 EFW system is comprised of two 100% capacity trains, with the "A" train pump (EFP-1) being motor driven and the "B" train pump (EFP-2) being steam driven. The steam for the EFP-2 turbine driver is fed through redundant inlet valves (ASV-5
, and ASV-204) to ensure the availability of steam given. a i failure of one of the inlet valves to open. Each pump feeds both steam generators. For a portion of the flow path from the emergency feedwater tank' (EFT-2), the two pumps share a
', common suction line. Under certain accident scenarios, there are failure modes whicn can cause the calculated NPSH l available to both ~ pumps to be less than reqeired. For !
i example, a failure of the DC control power ,surce for the j injection control valves in one train of EFW can result in the '
pump in that train producing high flows which result in excessive friction head losses through the common suction ,
- line. !
l I _. . . ..
l
1 -
l
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U. S. Nuclear Regulatory Commission 3F1096-22 Attachment Page 6 of 7 Concern 5.2: Motor-driven EFP-1 is powered from the "A" train ES bus and is connected to the "A" emergency diesel generator (EGDG-1A).
EFP-2 is steam driven and therefore does not affect "B" train l EDG loading. However, portions of the load management schenie l for EGDG-1A depend on the availability of EFP-2 to: 1) limit the total flow produced by EFP-1 during the early stages of diesel loading and 2) permit EFP-1 to be shut down and the "A" train LPI pump (and other engineered safeguards features) to be started in the later stages of accident mitigation.
Therefore, some postulated failure modes which cause EFP-2 to ;
i be unavailable invalidate assumptions made in EGDG-1A loading !
I
~
cal' 'ations and some accident analyses which may have taken cres . for flow from EFP-2 after EFP-1 was shut down.
Resolution: At this time, the following modifications are being considered:
i a. Installing cavitating venturis in the EFW pump discharge lines to limit flow during the postulated failures which result in i the loss of flow control for an EFW train. This will eliminate the NPSH concern.
- b. Re-enabling "A" train Emergency Feedwater Initiation and Control (EFIC) system actuation of EFP-2 via automatir;a. .y opening steam turbine inlet valve ASV-204. This feature was ;
disabled by a modification in Refuel 10 and will be restored i to ensure EFP-2 auto-starts given a failure of the "B" side initiate logic or ASV-5,
- c. Installing motor operators on cross-tie valves EFV-12 and EFV- ,
13 to allow remote manual opening of these valves. Opening !
these valves ' establishes a flow path allowing the pump from !
one train to feed the steam generators through the injection lines of the other train. This is desirable to ensure the operators can maintain EFW flow control and indication in certain single failure scenarios without requiring local manual valve operation.
Schedule: This issue will be resolved before startup from the current outage.
We expect this issue to require additional interaction with the NRC prior to restart.
- 6. Emeraency Diesel Generator Loadina
- Concern: The rated capacity of EGDG-1A is challenged by the continuous,
- automatically connected loads as well as the loads that are manually connected in the later stages of accident mitigation. Three concerns were created by the Refuel 10 modification which removed
- the "A" train EFIC automatic actuation of ASV-204. Calculated peak
- transient diesel loads were above the 3500 kW maximum engine rating i documented in the FSAR and the ITS basis background for LC0 3.8.1, "AC Sources"; calculated peak diesel load at one minute was above the 3100 kW rating discussed in the basis for Surveillance Requirement 3.8.1.11; and the highest single rejected diesel load f
l l U. S. Nuclear Regulatory Commission 3F1Q96-22 Attachment Page 7 of 7 i
discussed in the basis for Surveillance Requirement 3.8.1.8 increased.
Resolution: A combination of three efforts is being pursued to increase the load capability of EGDG-1A. They include an engine power upgrade to increase one or more of the load ratings; removal and/or reduction of connected loads; and improving the accuracy of the kW meters used I to display the generators' output. We expect this issue to also require additional NRC interaction prior to restart.
Schedule: This issue will be resolved before startup from the current outage.
l l
- 7. Failure Modes and Effects of Loss of DC Power Concern: A number of CR-3 design and operating vulnerabilities have been '
identified on a case-by-case basis through design and E0P reviews '
postulating the effects of a loss of DC power. The loss of DC power also causes a consequential loss of emergency AC power since DC power is required for emergency diesel generator field flashing and bus breaker closure. A Failure Modes and Effects Analysis (FMEA) was performed for the CR-3 Class lE electrical distribution system l (including DC) as part of the original plant design. However, it may not have fully considered system interactions, including effects on redundant trains and components.
Resolution: FPC will perform a DC power FMEA which includes evaluations of system interactions.
, Schedule: The FMEA review will be completed to the extent that FPC is
! satisfied that we have identified any safety significant problems.
Such problems will be addressed prior to startup from the current outage.
! 8. Generic letter 96-06 l Concern: This Generic Letter (GL) identifies three issues regarding the effect of post-accident containment heatup on containment coolers, piping, and penetrations. CR-3 is susceptible to the piping overpressurization phenomenon and is evaluating the water hammer and j
two-phase heat transfer problems.
l Resolution: FPC is installing thermal overpressure protection devices on containment penetrations affected by this phenomenon. Actions to address the impact of the other two issues, if any, will be determined after the review is completed.
Schedule: The overpressure protection devices will be installed prior to startup from the current outage. Actions to address the impact of
, the other two issues, if any, will be scheduled according to the safety significance of the findings.