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l Ibtruary 9,1990 i
MP-90-149 l
Re: 10CFR50.73(a)(2)(i)(B).
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U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Reference:
Facility Operating License No. DPR-65
' Docket No. 50-336 Licensee Event Report 90-001-00 Gentlemen:
- - l This letter forwaras Licensee Event Report 90-001-00 required to be submitted within thirty (30) days pursuant to 10CFR50.73(a)(i)(B).
Very truly yours, NORTHEAST NUCLEAR ENERGY COhiPANY i
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Sth eh. Scace
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Director, Millstone Station 4
SES/RWB:mo Attachment: - LER 90-001-00 cc:
W. T. Russell, Recion I Administrator W. J. Raymond, S'enior Resident inspector, Millstone Unit Nos,1, 2 and 3 G. S. Vissing, NRC Project Manager, Millstone Unit No. 2
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On January 10,1990 at 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br />, with the plant at 100rc power and normal operation, a review of a Technical Specification change request to the fire protection sections noted that the surveillances specified in the existing Technical Specifications for two listed areas were not being performed. The fire suppression Technical Specification surveillance procedures had been revised to delete the removed smoke detectors and add the testing of the new suppression systems for these areas. The Technical Specifications had not been revised to reflect the plant changes. This was considered to be a condition prohibited by the plant's Technical Specifications, since the surveillance requirement listed in the LCO was not performed. There were no safety consequences because the arent in question were protected at all times by the new systems installed to satisfy Appendix "R" requirements. The old detectors were removed durmp the 1967 and 1966 refuehng outapes, and no change to the Technical Specifications was generated due an administrative oversight. The new systems were installed in accordance with approved designs that were found acceptable by the NRC in subsequent reviews. Smce full protection for the areas in question was provided, no other actions are required, with the exception of the submittal of the required Technical Specification changes.
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On January 10,1990 at 074 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />, with the plant at 1009 power and normal operation, a resiew of a Technical Specification change request to the hre protection sections noted that the surveillances specihed in the existing Technical Specihcations for two hsted areas were not being performed. The fire suppression Technical Specification surveillance procedures had been revised to delete the removed smoke detectors and add the testing of the new suppression systems for these areas. The Technical Specifications had not been revised to reflect the plant changes. This was considered to be a condnion prohibned by the plant's Technical Specihcations, smce the surveillance requirement hsted in the LCO was not performed. There were no safety consequences because the areas in question were protected at all times by the new systems installed to sausly Appendix
- R* requirements. The old detectors were removed durmg the 1967 and 1966 refuehng outages, and no change to the Technical Specifications was generated due an administrative oversight. The new systems were installed in accordance wnh approved designs that were found acceptable by the NRC in subsequent reviews. Since full protecuon for the areas in question was provided, no other actions are required, with the exception of the submittal of the required Technical Specification changes. There were no operator actions or automatic safety system responses required by this event.
II, rue of Frent The root cause of the event is personnel error. The omission occurred during the planning and scheduling of events that were part of the modifications to the plant hre protection systems to meet the Appendix *R* requirements. The normal sequence for a known change to a system addressed in the Technical Specifications is to change the Technical Specification requirement such that it becomes effective when the new system will be placed in semce. In this case, the oversight occurred because hre protection is essentially required to be in service at all umes with appropriate compensatory action taken in case of a component or system failure or inoperabihty. Dunnp the transition between the old and new systems, the appropriate compensatory measures (hre watches and backup suppression systems) were in effect to assure the operabihty of the equipment within the areas. Following the completion of work the surveillances on the new systems were staned and the old systems surveillances were never resumed because the items to be checked (smoke detectors) had been remosed. The requirement to perform the old surveillances was overlooked because the new surveillances had taken their place through approved plant procedure revisions.
111.
Analvsk of Event This report is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(i)(B), *Any operation or condition prohibited by the plants Technical Specificauons* There were no safety consequences resulung from this event smce the areas m question were protected at all times by operable fire suppression systems as oemonstrated by the surveillances performed under procedures l
SP 2616D and SP 2412, on the new systems. The new systems were found to be acceptable to the NRC in a post installation audit. Due to the scheduling of the implementation of the new fire suppression systems to meet the Appendtx "R" requirement deadhnes, the requirement to update the technical specihcations before the changes actually occurred, was inadvertently overlooked.
During the transition penod between the old and new systems, the apptopriate compensatory c
measures consisting of fire watches and backup suppression systems (fire hoses) were in effect to l
assure the operabihty of the equipment wahm the areas.
IV.
Corrective Action
The missed surveillances were noted dunng the review of the Technical Specification change package being made to consolidate and update the hre protection sections of the Technical Specihcations. Since the areas in question are properly protected by the new systems and the surveillances being performed demonstrate the operability of the new systems, no other action is required, with the exception of the submittal of the required Technical Specihcation changes. The estimated schedule for submission of the required changes to the Technical Specihcanons is March 31.1490.
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F AcLITV NAME 111 DOCKET NJMBER (2) t FP Mupro eri PAGE 13 i
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In ' addition, it k expected that plans to follow the NRC recommendation for removal of fire protection items frorn the Technical Specihcations when implemented would prevent a recurrence of the events as reported herein. This is based on the fact that the report identifies requirements 5
t not being performed that were listed in the Technical Specifications for specific smoke detectors for.
i specibe areas. Because these specific areas were being protected from hre concerns by new systems and the new systems were being verihed as operable under new or revised procedures, the omission was only of an administrative nature, and removal of the administrative requirements from the Technical Specifications will prevent a recurrence of this nature.
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| 05000245/LER-1990-001-02, :on 900126,determined That 1 H Roving Fire Patrol Established on 900125 for Two Nonfunctional Fire Barriers Penetrations Not Consistent W/Requirements of Tech Spec 3.12.F.2.Caused by Personnel Error |
- on 900126,determined That 1 H Roving Fire Patrol Established on 900125 for Two Nonfunctional Fire Barriers Penetrations Not Consistent W/Requirements of Tech Spec 3.12.F.2.Caused by Personnel Error
| 10 CFR 50.73(a)(2)(i) | | 05000423/LER-1990-001, :on 910108,inoperable Fire Hose Stations Due to Insufficient Fire Hose Inventory.Cause Unknown for Fuel Bldg Fire Hose Rack & Procedural Inadequacy for Auxiliary Bldg Fuel Racks.Station Restored to Operable Status |
- on 910108,inoperable Fire Hose Stations Due to Insufficient Fire Hose Inventory.Cause Unknown for Fuel Bldg Fire Hose Rack & Procedural Inadequacy for Auxiliary Bldg Fuel Racks.Station Restored to Operable Status
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000336/LER-1990-001, :on 900110,discovered That Surveillances in Tech Specs for New Smoke Detectors & Suppression Sys Not Performed.Caused by Personnel Error.Fire Protection Sections of Tech Specs Updated |
- on 900110,discovered That Surveillances in Tech Specs for New Smoke Detectors & Suppression Sys Not Performed.Caused by Personnel Error.Fire Protection Sections of Tech Specs Updated
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000423/LER-1990-001-01, :on 900108,discovered That Fire Hose Racks in Fuel & Auxiliary Blgs Had Insufficient Fire Lengths to Fulfill Tech Spec Operability Requirements.Caused by Procedural Inadequacy.Caution Signs Posted |
- on 900108,discovered That Fire Hose Racks in Fuel & Auxiliary Blgs Had Insufficient Fire Lengths to Fulfill Tech Spec Operability Requirements.Caused by Procedural Inadequacy.Caution Signs Posted
| 10 CFR 50.73(a)(2)(i) | | 05000245/LER-1990-002-01, :on 900305,main Steam Line High Flow Setpoint Determined to Be Incorrect & Nonconservative.Caused by Use of Incorrect Assumptions Based on Original Plant Design Conditions.Setpoint Corrected |
- on 900305,main Steam Line High Flow Setpoint Determined to Be Incorrect & Nonconservative.Caused by Use of Incorrect Assumptions Based on Original Plant Design Conditions.Setpoint Corrected
| 10 CFR 50.73(a)(2)(i) | | 05000336/LER-1990-002-02, :on 900322,Tech Spec Action Statement 3.3.3.10 Not Entered for Out of Svc Stack Gas & Particulate Radiation Monitor.No Particulate Radiation Increases Detected.Caused by Personnel Error |
- on 900322,Tech Spec Action Statement 3.3.3.10 Not Entered for Out of Svc Stack Gas & Particulate Radiation Monitor.No Particulate Radiation Increases Detected.Caused by Personnel Error
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000423/LER-1990-002, :on 900109,partial Train a Containment Depressurization Actuation Signal Generated.Caused by Personnel Error.Warnings Added to Production Maint Mgt Sys for Recirculation Spray Sys Pump Breakers |
- on 900109,partial Train a Containment Depressurization Actuation Signal Generated.Caused by Personnel Error.Warnings Added to Production Maint Mgt Sys for Recirculation Spray Sys Pump Breakers
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000245/LER-1990-003, :on 900302,determined That 4-day Limiting Condition for Operation for One Emergency Power Sources, Gas Turbine Generator Exceeded.Caused by Lack of Verification of Load Requirement |
- on 900302,determined That 4-day Limiting Condition for Operation for One Emergency Power Sources, Gas Turbine Generator Exceeded.Caused by Lack of Verification of Load Requirement
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(0)(i) | | 05000423/LER-1990-003, :on 900115,discovered That Fire Watches Not Established Prior to Removing Deluge Sys for Trains a & B Reserve Station Svc Transformers from Svc.Caused by Personnel Error.Personnel Counseled |
- on 900115,discovered That Fire Watches Not Established Prior to Removing Deluge Sys for Trains a & B Reserve Station Svc Transformers from Svc.Caused by Personnel Error.Personnel Counseled
| 10 CFR 50.73(a)(2)(i) | | 05000336/LER-1990-004-01, :on 900228,control Room Ventilation Sys Operated Outside Tech Specs.Caused by Inconsistency in Tech Specs.Proposed Tech Spec Change Request Submitted Re Emergency Diesel Generator Operability |
- on 900228,control Room Ventilation Sys Operated Outside Tech Specs.Caused by Inconsistency in Tech Specs.Proposed Tech Spec Change Request Submitted Re Emergency Diesel Generator Operability
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000423/LER-1990-004, :on 900115,nuclear Instrument Power Range Channel N43 Became Inoperable Due to Power Supply Failure. Caused by Computer Program 3R5 Design Inadequacy.Night Order Issued to Enter Tech Spec 4.2.1.1.1.b |
- on 900115,nuclear Instrument Power Range Channel N43 Became Inoperable Due to Power Supply Failure. Caused by Computer Program 3R5 Design Inadequacy.Night Order Issued to Enter Tech Spec 4.2.1.1.1.b
| 10 CFR 50.73(a)(2)(i) | | 05000245/LER-1990-004-02, :on 900406,calculation Which Was Performed to Verify Reactor High Pressure Scram Setpoint Demonstrated That Existing Head Correction Nonconservative.Caused by Lack of Independent Verification of Setpoint |
- on 900406,calculation Which Was Performed to Verify Reactor High Pressure Scram Setpoint Demonstrated That Existing Head Correction Nonconservative.Caused by Lack of Independent Verification of Setpoint
| 10 CFR 50.73(a)(2)(i) | | 05000336/LER-1990-004, :on 900228,Unit Operated Outside of Tech Specs for Control Room Air Conditioning Sys.Caused by Inconsistency in Tech Specs.Tech Spec Change Request Approved |
- on 900228,Unit Operated Outside of Tech Specs for Control Room Air Conditioning Sys.Caused by Inconsistency in Tech Specs.Tech Spec Change Request Approved
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) | | 05000336/LER-1990-005, :on 900503,notified of Potential High Energy Line Break Via as Sys in Safety Related Areas.Caused by Inaccurate Conclusions Drawn from 1973 Rept.As Sys Removed from Svc & Plant Mods Initiated |
- on 900503,notified of Potential High Energy Line Break Via as Sys in Safety Related Areas.Caused by Inaccurate Conclusions Drawn from 1973 Rept.As Sys Removed from Svc & Plant Mods Initiated
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000336/LER-1990-005-02, :on 900503,identified Potential for High Energy Line Break in Auxiliary Steam Sys That Could Degrade Plant Areas Determined as Mild Environs.Probably Caused by Incorrect Conclusions from Analysis in 1973 |
- on 900503,identified Potential for High Energy Line Break in Auxiliary Steam Sys That Could Degrade Plant Areas Determined as Mild Environs.Probably Caused by Incorrect Conclusions from Analysis in 1973
| 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vi) | | 05000423/LER-1990-005, :on 900118,manual Plant Trip Initiated in Anticipation of Automatic Trip on lo-lo Level in All Four Steam Generators.Caused by Loss of Preload on Coupling Blocks from Personnel Error.Procedure Revised |
- on 900118,manual Plant Trip Initiated in Anticipation of Automatic Trip on lo-lo Level in All Four Steam Generators.Caused by Loss of Preload on Coupling Blocks from Personnel Error.Procedure Revised
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000245/LER-1990-005-01, :on 900411,determined That Daily Surveillance Greater than 6 H from Previous Days Surveillance.Caused by Combination of Factors Including Administrative Deficiency. Operations Dept Logs Revised |
- on 900411,determined That Daily Surveillance Greater than 6 H from Previous Days Surveillance.Caused by Combination of Factors Including Administrative Deficiency. Operations Dept Logs Revised
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000245/LER-1990-006-01, :on 900424,determined That Gas Turbine Encl Not Included in Monthly Surveillance for Nonsupervised Circuits. Caused by Personnel Error.Gas Turbine Fire Detection Surveillance Changed to Monthly |
- on 900424,determined That Gas Turbine Encl Not Included in Monthly Surveillance for Nonsupervised Circuits. Caused by Personnel Error.Gas Turbine Fire Detection Surveillance Changed to Monthly
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000423/LER-1990-006, :on 900118,chemistry Dept Determined 6-month Interval Tech Spec for Average Disintegration Energy Determination Not Met.Caused by Inappropriate Appliance of Surveillance Interval.Addl Controls Placed |
- on 900118,chemistry Dept Determined 6-month Interval Tech Spec for Average Disintegration Energy Determination Not Met.Caused by Inappropriate Appliance of Surveillance Interval.Addl Controls Placed
| 10 CFR 50.73(a)(2)(i) | | 05000336/LER-1990-006-02, :on 900508,reactor Manually Tripped When Decreasing Levels Noted in Steam Generator 1 & Feedwater Regulating Valve Indicated Full Open.Caused by Valve Stem Separating from Plug.Feedwater Flow Restored |
- on 900508,reactor Manually Tripped When Decreasing Levels Noted in Steam Generator 1 & Feedwater Regulating Valve Indicated Full Open.Caused by Valve Stem Separating from Plug.Feedwater Flow Restored
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) | | 05000245/LER-1990-007-01, :on 900512,determined That One Emergency Power Source Did Not Have Sufficient Capacity for Accident Loading Conditions.Gas Turbine Generator Declared Inoperable & Orderly Reactor Shutdown Initiated |
- on 900512,determined That One Emergency Power Source Did Not Have Sufficient Capacity for Accident Loading Conditions.Gas Turbine Generator Declared Inoperable & Orderly Reactor Shutdown Initiated
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) | | 05000336/LER-1990-007, :on 900611,discovered Missed Surveillance Prior to Entering Mode 4 |
- on 900611,discovered Missed Surveillance Prior to Entering Mode 4
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000336/LER-1990-007-02, :on 900611,discovered That Surveillance Procedure 2609E Re Encl Bldg Filtration Sys Testing - Refueling Not Performed Prior to Entering Mode 4.Caused by Personnel Error.Missed Surveillance Performed |
- on 900611,discovered That Surveillance Procedure 2609E Re Encl Bldg Filtration Sys Testing - Refueling Not Performed Prior to Entering Mode 4.Caused by Personnel Error.Missed Surveillance Performed
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000423/LER-1990-007, :on 900125,discovered That Tech Spec Surveillance Did Not Verify Load Shedding of Five Compressors.Caused by Procedural Inadequacy & Personnel Error.Surveillance Procedure Revised |
- on 900125,discovered That Tech Spec Surveillance Did Not Verify Load Shedding of Five Compressors.Caused by Procedural Inadequacy & Personnel Error.Surveillance Procedure Revised
| 10 CFR 50.73(a)(2)(i) | | 05000245/LER-1990-008-01, :on 900531,determined That Fuel Thermal Limit Exceeded Tech Spec Limit.Caused by Underestimation of Xenon Transient That Resulted from Power Reduction.Refresher Course Planned for Personnel |
- on 900531,determined That Fuel Thermal Limit Exceeded Tech Spec Limit.Caused by Underestimation of Xenon Transient That Resulted from Power Reduction.Refresher Course Planned for Personnel
| 10 CFR 50.73(a)(2)(i) | | 05000336/LER-1990-008-02, :on 900620,determined That Grab Sample of Unit Stack Gas Not Taken.Caused by Lack of Communication Between Personnel.Grab Sample Obtained & Analyzed |
- on 900620,determined That Grab Sample of Unit Stack Gas Not Taken.Caused by Lack of Communication Between Personnel.Grab Sample Obtained & Analyzed
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000245/LER-1990-009-01, :on 900511,determined That House Heating Steam Sys Could Potentially Degrade Environ Classified, Eeq Mild Environs. Caused by Incorrect Conclusion Drawn from 1973 Study.Plant Mods Implemented |
- on 900511,determined That House Heating Steam Sys Could Potentially Degrade Environ Classified, Eeq Mild Environs. Caused by Incorrect Conclusion Drawn from 1973 Study.Plant Mods Implemented
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000336/LER-1990-009-02, :on 900617,inadvertent Partial Actuation of Train B of Enclosure Bldg Filtration Sys Occurred.Root Cause Not Determined.No Corrective Actions Recommended Until Further Testing & Troubleshooting Performed |
- on 900617,inadvertent Partial Actuation of Train B of Enclosure Bldg Filtration Sys Occurred.Root Cause Not Determined.No Corrective Actions Recommended Until Further Testing & Troubleshooting Performed
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-009-01, :on 900617,facility Experienced Inadvertent Partial Actuation of Train B of Auxiliary Exhaust Actuation Sys.Cause Undetermined.Two Addl Actuations Occurred on 900711 |
- on 900617,facility Experienced Inadvertent Partial Actuation of Train B of Auxiliary Exhaust Actuation Sys.Cause Undetermined.Two Addl Actuations Occurred on 900711
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000423/LER-1990-009, :on 900309,automatic Turbine Trip W/Subsequent Reactor Trip Occurred Due to High Stator Cooling Water Temp. Caused by Failure of Mechanical Linkage on Fisher & Portor Controller.Controller Replaced W/Spare |
- on 900309,automatic Turbine Trip W/Subsequent Reactor Trip Occurred Due to High Stator Cooling Water Temp. Caused by Failure of Mechanical Linkage on Fisher & Portor Controller.Controller Replaced W/Spare
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(viii)(B) | | 05000245/LER-1990-010-01, :on 891223,EEQ Barriers Violated W/Switchgear Area & Heating & Ventilation Room Mild Environ & Turbine Deck Had Potential Harsh Environ.Caused by Lack of Formal guidance.Long-term Program Developed |
- on 891223,EEQ Barriers Violated W/Switchgear Area & Heating & Ventilation Room Mild Environ & Turbine Deck Had Potential Harsh Environ.Caused by Lack of Formal guidance.Long-term Program Developed
| | | 05000423/LER-1990-010, :on 900319,both Trains of Auxiliary Bldg Filters Became Inoperable When Train B Circuit Breaker Motor Failed & Train a Filter Removed from Svc.Caused by Fatigue Failure in Breaker Spring.Filter Work Stopped |
- on 900319,both Trains of Auxiliary Bldg Filters Became Inoperable When Train B Circuit Breaker Motor Failed & Train a Filter Removed from Svc.Caused by Fatigue Failure in Breaker Spring.Filter Work Stopped
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000336/LER-1990-010-02, :on 900621,door Identified in Configuration Not Consistent W/Bechtel Design Drawings During High Energy Line Review.Caused by Lack of Knowledge of Requirements.Double Door Reinforced |
- on 900621,door Identified in Configuration Not Consistent W/Bechtel Design Drawings During High Energy Line Review.Caused by Lack of Knowledge of Requirements.Double Door Reinforced
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000336/LER-1990-010-01, :on 900621,door Identified as Being in Configuration Not Consistent W/Bechtel Design Drawings Due to Lack of Knowledge of HELB Requirements for Area |
- on 900621,door Identified as Being in Configuration Not Consistent W/Bechtel Design Drawings Due to Lack of Knowledge of HELB Requirements for Area
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000245/LER-1990-011-01, :on 900720,Tech Spec Fire Door Found Blocked Open & Unguarded by Fire Watch.Caused by Personnel Error. Personnel Cautioned to Be Aware of Potential for Impacting Tech Spec Barrier Requirements |
- on 900720,Tech Spec Fire Door Found Blocked Open & Unguarded by Fire Watch.Caused by Personnel Error. Personnel Cautioned to Be Aware of Potential for Impacting Tech Spec Barrier Requirements
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) | | 05000423/LER-1990-011, :on 900330,manual Reactor Trip Initiated Due to Anticipated Turbine Trip from Loss of Condenser Vacuum. Caused by Failure to Collect Debris from Manual Screen Washing.Elbow Replaced & Screen Wash Restored |
- on 900330,manual Reactor Trip Initiated Due to Anticipated Turbine Trip from Loss of Condenser Vacuum. Caused by Failure to Collect Debris from Manual Screen Washing.Elbow Replaced & Screen Wash Restored
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-012-02, :on 900827,automatic Reactor Trip Occurred During Bypass Switch Operations.Caused by Operator Error. Procedure Sp 2601D Revised to Incorporate Separate Section on Performing Calibrs |
- on 900827,automatic Reactor Trip Occurred During Bypass Switch Operations.Caused by Operator Error. Procedure Sp 2601D Revised to Incorporate Separate Section on Performing Calibrs
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000245/LER-1990-012-01, :on 900813,determined That hi-hi Trip Settings on Both Offgas Instrument Drawers Set in Nonconservative Direction & Exceeded Tech Spec 3.8.B.1.Caused by Failure to Recognize Significance of Response Factor |
- on 900813,determined That hi-hi Trip Settings on Both Offgas Instrument Drawers Set in Nonconservative Direction & Exceeded Tech Spec 3.8.B.1.Caused by Failure to Recognize Significance of Response Factor
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) | | 05000423/LER-1990-012, :on 900406,review of Steam Generator Blowdown Monitor High Radiation Alarm Setpoint Revealed That Setpoint Was Nonconservative.Caused by Administrative Deficiency. Correct Setpoint Installed |
- on 900406,review of Steam Generator Blowdown Monitor High Radiation Alarm Setpoint Revealed That Setpoint Was Nonconservative.Caused by Administrative Deficiency. Correct Setpoint Installed
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000245/LER-1990-013, :on 900806,standby Gas Treatment Sys Initiation Occurred Due to Inadvertent de-energization of Power Supply to Channel 2 Process Radiation Monitoring Sys.Power Restored to Channel 2 & Initiation Trip Logic Reset |
- on 900806,standby Gas Treatment Sys Initiation Occurred Due to Inadvertent de-energization of Power Supply to Channel 2 Process Radiation Monitoring Sys.Power Restored to Channel 2 & Initiation Trip Logic Reset
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000423/LER-1990-013, :on 900416,manual Reactor Trip Occurred Due to Imminent Loss of Condenser Vacuum.Caused by Inadequate Administrative Guidance.Personnel Instructed to Closely Monitor Trash Rack Water Levels |
- on 900416,manual Reactor Trip Occurred Due to Imminent Loss of Condenser Vacuum.Caused by Inadequate Administrative Guidance.Personnel Instructed to Closely Monitor Trash Rack Water Levels
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-014-02, :on 901010,main Steam Safety Valve Setpoint Drift Discovered During as-found Simmer Test |
- on 901010,main Steam Safety Valve Setpoint Drift Discovered During as-found Simmer Test
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000245/LER-1990-014-01, :on 900907,inconsistency Between Procedural & Design Parameters Associated W/Lpci HX Flow Rates Identified.Caused by Inadequate Evaluation of Original Plant Design Documentation |
- on 900907,inconsistency Between Procedural & Design Parameters Associated W/Lpci HX Flow Rates Identified.Caused by Inadequate Evaluation of Original Plant Design Documentation
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000423/LER-1990-014, :on 900519,manual Reactor Trip Initiated as Result of Anticipated Turbine Trip Due to Condenser Vacuum. Caused by Design Deficiency in That Traveling Screen Capacity Inadequate.Traveling Screen Modified |
- on 900519,manual Reactor Trip Initiated as Result of Anticipated Turbine Trip Due to Condenser Vacuum. Caused by Design Deficiency in That Traveling Screen Capacity Inadequate.Traveling Screen Modified
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000423/LER-1990-015, :on 900512,feedwater Isolation Occurred While Opening Msivs.Caused by MSIV 2 Opening Faster than Other Msivs,Resulting in Swell in Steam Generator 2.Steam Generator Level Restored to Normal |
- on 900512,feedwater Isolation Occurred While Opening Msivs.Caused by MSIV 2 Opening Faster than Other Msivs,Resulting in Swell in Steam Generator 2.Steam Generator Level Restored to Normal
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-015, :on 900919,unit Experienced Inadvertent Isolation of Containment Purge Valves.On 900920,unit Experienced Actuation of SI Actuation Sys.Caused by Operator Error.Technician Counseled |
- on 900919,unit Experienced Inadvertent Isolation of Containment Purge Valves.On 900920,unit Experienced Actuation of SI Actuation Sys.Caused by Operator Error.Technician Counseled
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-015-02, :on 900919,containment Purge Valves 2-AC-4, 2-AC-5,2-AC-6 & 2-AC-7 Inadvertently Isolated |
- on 900919,containment Purge Valves 2-AC-4, 2-AC-5,2-AC-6 & 2-AC-7 Inadvertently Isolated
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) | | 05000245/LER-1990-015-01, :on 900914,reactor Scram Occurred on Low Reactor Water Level After Feedwater Regulating Valves Began to Close |
- on 900914,reactor Scram Occurred on Low Reactor Water Level After Feedwater Regulating Valves Began to Close
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000336/LER-1990-016, :on 901009,inadvertent ESAS Actuation of Containment Purge Isolation Sys Occurred.Caused by Loose Ground Wire in Control Room Cabinet S14D.Ground Wire Termination Tightened |
- on 901009,inadvertent ESAS Actuation of Containment Purge Isolation Sys Occurred.Caused by Loose Ground Wire in Control Room Cabinet S14D.Ground Wire Termination Tightened
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000336/LER-1990-016-02, :on 901009,inadvertent ESAS Actuation Occurred in Violation of Tech Specs |
- on 901009,inadvertent ESAS Actuation Occurred in Violation of Tech Specs
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) |
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