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U.S. Nuclear Regulatory Commission L
Mail Station P1-137 i
Washington, D.C.
20555 i
Attention: Document Control Desk Gentlemen:
SUBJECT:
Grand Gulf Nuclear Station Unit 1 Doclet.No. 50-416 License No. NPF-29
. Uncontrolled Lowering of
- - i Fuel' Bundle-LER 90-021 AECM-90/0200 Attached is Licensee Event' Report (LER)90-021 which is a final report.
Yours.truly, l
MY W WTC/RR:cg Attachment cc:
Mr. D. C. Hintz (w/a)-
Mr.R.B.McGehee(w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr.H.O.Christensen(w/a)
Mr. Stewart D. Ebneter (w/a)
Regional Administrator U.S. Nuclear Regulatory Commission Region-11~
- - 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323
.i Mr. L. L. Kintner, Project Manager (w/a)
Office of Nuclear Reactor Regulation U.S. ' Nuclear Regulatory Commission Mail Stop 11D21 Washington, D.C.
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On October 24, 1990 during fuel moves, a fully grappled fuel bundle was lowered into the vessel in an uncontrolled manner due to independent failures of redundant refueling equipment brake systems.
The bundle entered it's targeted core location striking no adjacent structures.
The failure of the normal safety brake is attributed to the safety brake manual disengage lever binding in the disengaged position due to close proximity with the brake housing cover. The emergency brake also failed due to the emergency brake manual disengage lever binding in the disengaged position due to a deformed spring.
1 Chemistry analyses and radiation surveys indicated no detectable fission product or gas release occurred.
The motor / brake assembly was replaced and tested satisfactorily.
The emergency brake disengage rod.was removed from both the refueling and fuel handling bridges.
Even though inspection revealed no apparent physical damage, the subject fuel-bundle was replaced in the core configuration with a slightly less reactive bundle.
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Reportable Occurrence i
On October 24, 1990 during fuel moves, a failure of the refueling equipment brake systems caused an uncontroiled lowering of a fuel bundle into it's respective core location. This event is not reportable per 50.73 but is reported as a valuntary report'.
B.
Initial Cause The plant was in Operational Condition 5, Refueling, at the time of the occurrence.
C.
Description of Occurrence On October 24, 1990 at approximately 2135, fuel bundle XNB-487 was being moved to a new location in the core.
The Refueling Platform Laser Track System was being used to position the bundle over it's targeted core location and to lower it to a preset mast height where it could be manually lowered into the core. The system started its automatic descent. Upon reaching the preset mast height,-which would automatically stop the bundle's descent, the bundle failed to stop and appeared to accelerate during its descent.
During the accelerated descent, the emergency stop button was depressed, which should have engaged the emergency brake, with no apparent affect. The bundle continued to descend until it was seated l
in its support piece without striking any adjacent structures.
It is estimated that the bundle descended.at an approximate average speed of-3 feet per second.
There was no evidence of gases being released from l
the bundle after it was seated.
An investigation was performed prior to restoring the refueling bridge to service.
D.
Apparent-Cause During the investigation, two abnormalities were observed.
The emergency brake manual disengage rod was found in the disengaged position and the manual safety brake.. lever was also found in the disengaged position.
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Further investigations were performed in a lab environment. Following l
the testing of the motor and brake, it was determined that the normal safety brake manual disengage lever was susceptible to binding on the brake housing cover and remain disengaged when the brake solenoid was deenergized (normally deenergizing the brake solenoid would cause the
.I brake to engage).
Additionally, one of the brake adjustment screws was loose and the solenoid air gap was found out of tolerance.
Investigation of the emergency brake revealed that the manual disengage rod was susceptible to binding due to a deformed spring which was normally in contact with the manual disengage rod Demonstrations indicated that the rod bound at a high enough posiiton to disengage the brake.
Based on the results of the investigation, the probable causes of the normal and emergency brake failures, as determined by olant personnel, were the binding of normal brake lever on the motor housing due to close tolerances and the binding of the emergency brake lever in the up position due to a deformed spring.
E.
Supplementai Corrective Actions o
The motor / brake assembly was replaced.and was tested to verify proper logic functions. The brake adjustment was verified per the vendor requirements.
Additionally, the housing cover was milled out so that sufficient clearance existed between the safety brake's manual disengage lever and the housing cover.
The manual disengage rod for the emergency brake was also removed from the refueling and fuel handling bridges as a part of the corrective actions leaving energization of the emergency brake soelnoid as a means for releasing this brake.
During the remaining fuel moves, frequent inspections were performed on both the refueling and-fuel handling bridges to ensure proper operation, o
An inspection and evaluation was performed on the fuel bundle components, no apparent structural damage was identified.
Even though the bundle appeared fully intact, it was removed from the core and discharged to the spent fuel pool.
The bundle was replaced in the core configuration by a suitable bundle l
scheduled for discharge.
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Chemistry sample. analyses and radiation surveys were performed l
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release occurred. Therefore, there was no indication of a loss l
of cladding integrity.
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F. Safety Assessment
The fuel bundle was initially inserted into the core during Cycle 3 and reinserted during Cycle 4.
Based on the results of the surveys and analyses performed by plant personnel, the event did not adversely impact plant or public safety, i
The control rod which is associated with the subject bundle was stroked and no abnormal operation was observed.
The revised core configuration replaced bundle (XNB-487)'with a' slightly less reactive bundle which will result.in a slight decrease in reactivity.
Therefore, the shutdown margin analysis bounds the revised core configuration.
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| | | Reporting criterion |
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| 05000416/LER-1990-001, :on 900112,determined That Tech Spec Reactor Coolant Cooldown Rate Limit of 100 Degrees F Exceeded in Reactor Bottom Head Drain Pipe.Caused by Programmatic Deficiencies.Incident Rept Form Improved |
- on 900112,determined That Tech Spec Reactor Coolant Cooldown Rate Limit of 100 Degrees F Exceeded in Reactor Bottom Head Drain Pipe.Caused by Programmatic Deficiencies.Incident Rept Form Improved
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-002, :on 900124,discovered That Div II Solenoid Associated W/Secondary Containment Isolation Valve Did Not Operate During Quarterly Valve Stroke Time Tests.Caused by Procedure Inadequacy.Procedures Changed |
- on 900124,discovered That Div II Solenoid Associated W/Secondary Containment Isolation Valve Did Not Operate During Quarterly Valve Stroke Time Tests.Caused by Procedure Inadequacy.Procedures Changed
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-003, :on 900215,identified Potential Single Failure Scenario That Could Result in Unavailability of Both Core Spray Sys for Long Term post-LOCA Core Cooling.Detailed Heatup Evaluation Performed for Power Bundle |
- on 900215,identified Potential Single Failure Scenario That Could Result in Unavailability of Both Core Spray Sys for Long Term post-LOCA Core Cooling.Detailed Heatup Evaluation Performed for Power Bundle
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000416/LER-1990-003-01, Corrected LER 90-003-01:on 900215,potential Single Failure Scenario Identified That Could Result in Unavailability of Both Core Spray Sys for post-LOCA Core Cooling.Detailed Heatup Evaluation Performed | Corrected LER 90-003-01:on 900215,potential Single Failure Scenario Identified That Could Result in Unavailability of Both Core Spray Sys for post-LOCA Core Cooling.Detailed Heatup Evaluation Performed | | | 05000416/LER-1990-004, :on 900406,discovered That Surveillance Procedure for Verifying Air Pressure in Containment Airlock Seal Flasks Did Not Fully Implement Tech Spec 4.6.1.3.d.2. Caused by Deficient Procedures |
- on 900406,discovered That Surveillance Procedure for Verifying Air Pressure in Containment Airlock Seal Flasks Did Not Fully Implement Tech Spec 4.6.1.3.d.2. Caused by Deficient Procedures
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-005, :on 900406,discovered That Current Surveillance Practices for Stroke Testing Fresh Air Makeup Intake Valves Could Have Resulted in Unfiltered Pathway.Administrative Controls Placed on Valves |
- on 900406,discovered That Current Surveillance Practices for Stroke Testing Fresh Air Makeup Intake Valves Could Have Resulted in Unfiltered Pathway.Administrative Controls Placed on Valves
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000416/LER-1990-006, :on 900502,review of Effluent Sample Analysis Revealed That Turbine Bldg Ventilation Exhaust Had Not Been Analyzed Exceeding Time Limit.Caused by Personnel Error. Training & Counseling Conducted |
- on 900502,review of Effluent Sample Analysis Revealed That Turbine Bldg Ventilation Exhaust Had Not Been Analyzed Exceeding Time Limit.Caused by Personnel Error. Training & Counseling Conducted
| | | 05000416/LER-1990-007, :on 900516,discovered That Actions Taken for Inoperable Reactor Water Level Transmitter Not Adequate to Comply W/Requirements for Tech Spec 3.3.2.Caused by Personnel Error.Meetings Held W/Personnel |
- on 900516,discovered That Actions Taken for Inoperable Reactor Water Level Transmitter Not Adequate to Comply W/Requirements for Tech Spec 3.3.2.Caused by Personnel Error.Meetings Held W/Personnel
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-008, :on 900521,discovered That Fire Rated Door Required by Tech Spec 3/4.7.7 Not Designated as Tech Spec Door in Ssurveillance Procedures.Caused by Incorrect Interpretation of Tech Spec Requirements |
- on 900521,discovered That Fire Rated Door Required by Tech Spec 3/4.7.7 Not Designated as Tech Spec Door in Ssurveillance Procedures.Caused by Incorrect Interpretation of Tech Spec Requirements
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-009, :on 900526,nonlicensed Personnel Performed Step Out of Sequence During Breaker rack-out & Caused LPCS Pump Breaker to Close.Operator Involved Counseled on Failure to Adhere to Breaker Operation Procedure |
- on 900526,nonlicensed Personnel Performed Step Out of Sequence During Breaker rack-out & Caused LPCS Pump Breaker to Close.Operator Involved Counseled on Failure to Adhere to Breaker Operation Procedure
| | | 05000416/LER-1990-010-01, :on 900706,error Discovered in Evaluation Used to Demonstrate Adequacy of Svc Water Flow to HPCS Pump Room Cooler.Matls Nonconformance Rept Generated to Document Discrepancy |
- on 900706,error Discovered in Evaluation Used to Demonstrate Adequacy of Svc Water Flow to HPCS Pump Room Cooler.Matls Nonconformance Rept Generated to Document Discrepancy
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000416/LER-1990-011-01, :on 900724,reactor Tripped on High Reactor Water Level W/Reactor Power in Process of Being Reduced in Attempt to Control Turbine Pump B Oscillations.Caused by Malfunction of Controller.Controller Calibr |
- on 900724,reactor Tripped on High Reactor Water Level W/Reactor Power in Process of Being Reduced in Attempt to Control Turbine Pump B Oscillations.Caused by Malfunction of Controller.Controller Calibr
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-013, :on 900725,instrument Cable from B Detector Disconnected Rather than C Detector While Performing Maint on Monitoring Sys.Caused by Personnel Error.Personnel Counseled on Methods for Self Verification |
- on 900725,instrument Cable from B Detector Disconnected Rather than C Detector While Performing Maint on Monitoring Sys.Caused by Personnel Error.Personnel Counseled on Methods for Self Verification
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-014, :on 900810,retest Not Performed Prior to Returning Secondary Containment Isolation Valve to Operable Status.Caused by Programmatic Weaknesses in Administrative Controls.Plant Procedures Changed |
- on 900810,retest Not Performed Prior to Returning Secondary Containment Isolation Valve to Operable Status.Caused by Programmatic Weaknesses in Administrative Controls.Plant Procedures Changed
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-015, :on 900823,discovered That Fire Rated Assembly Penetration Not Properly Sealed.Cause Not Determined. Nonconformance Rept Written & Work Order Initiated to Seal Penetration |
- on 900823,discovered That Fire Rated Assembly Penetration Not Properly Sealed.Cause Not Determined. Nonconformance Rept Written & Work Order Initiated to Seal Penetration
| | | 05000416/LER-1990-016-01, :on 900912,cooling Water Outlet Valve Failed to Stroke to Full Open Position & Div II Purge Sys Declared Inoperable |
- on 900912,cooling Water Outlet Valve Failed to Stroke to Full Open Position & Div II Purge Sys Declared Inoperable
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-017, :on 900916,reactor Scram Occurred Due to Loss of Balance of Plant Busses |
- on 900916,reactor Scram Occurred Due to Loss of Balance of Plant Busses
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-018-01, :on 901008,secondary Containment Doors Found Open During Refueling Outage |
- on 901008,secondary Containment Doors Found Open During Refueling Outage
| 10 CFR 50.73(a)(2)(1) | | 05000416/LER-1990-019-01, :on 901014,during Refueling Outage Four,Three Events Occurred in Which Same Power Supply Breaker Inadvertently Opened |
- on 901014,during Refueling Outage Four,Three Events Occurred in Which Same Power Supply Breaker Inadvertently Opened
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-020-01, :on 901016,containment Cooling Sys Found on High Radiation Level |
- on 901016,containment Cooling Sys Found on High Radiation Level
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-021-01, :on 901024,uncontrolled Lowering of Fuel Bundle Occurred |
- on 901024,uncontrolled Lowering of Fuel Bundle Occurred
| | | 05000416/LER-1990-022, :on 901026,loss of Shutdown Cooling Due to Inadequate Procedure |
- on 901026,loss of Shutdown Cooling Due to Inadequate Procedure
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | | 05000416/LER-1990-022-01, :on 901026,breaker 52-15309,which Supplies Power to Common Suction Isolation Valve E12-F008,closed, Tripping RHR Pump B & Isolating Shutdown Cooling Sys.Caused by Inadequate Procedure.Standing Order Issued |
- on 901026,breaker 52-15309,which Supplies Power to Common Suction Isolation Valve E12-F008,closed, Tripping RHR Pump B & Isolating Shutdown Cooling Sys.Caused by Inadequate Procedure.Standing Order Issued
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | | 05000416/LER-1990-023, :on 901105,shutdown Cooling Isolation Occurred Due to Blown Fuse |
- on 901105,shutdown Cooling Isolation Occurred Due to Blown Fuse
| 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | | 05000416/LER-1990-024, :on 901114,standby Fresh Air Unit Actuation Occurred Due to Inadequate Test Instruction |
- on 901114,standby Fresh Air Unit Actuation Occurred Due to Inadequate Test Instruction
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-025, :on 901122,actuation of Reactor Protection Sys Occurred During Surveillance of Reactor Mode Switch.Caused by Diminishing Power Supply of Test Equipment.Evaluation Will Be Conducted |
- on 901122,actuation of Reactor Protection Sys Occurred During Surveillance of Reactor Mode Switch.Caused by Diminishing Power Supply of Test Equipment.Evaluation Will Be Conducted
| | | 05000416/LER-1990-026, :on 901124,manual Scram Inserted Following Lockup of Rod Pattern Control Sys During Reactor Startup. Caused by Transient of Reactor Water Level Attributed to Open Drain Valves.Startup Procedure Amended |
- on 901124,manual Scram Inserted Following Lockup of Rod Pattern Control Sys During Reactor Startup. Caused by Transient of Reactor Water Level Attributed to Open Drain Valves.Startup Procedure Amended
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-028, :on 901210,ESF,RPS & ECCS Hpsc Actuated, Resulting in Reactor Scram from Full Power.Caused by Failed Solder Joint in Instrument Air Sys & Leaking Root Valve. Joint Reinspected & Procedures Revised |
- on 901210,ESF,RPS & ECCS Hpsc Actuated, Resulting in Reactor Scram from Full Power.Caused by Failed Solder Joint in Instrument Air Sys & Leaking Root Valve. Joint Reinspected & Procedures Revised
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000416/LER-1990-029, :on 901218,reactor Protection Sys Actuation Occurred,Resulting in Automatic Plant Shutdown Due to Reactor Feedwater Pump Trip.Caused by Air Supply Valve Not Fully Open.Air Supply Valve Replaced |
- on 901218,reactor Protection Sys Actuation Occurred,Resulting in Automatic Plant Shutdown Due to Reactor Feedwater Pump Trip.Caused by Air Supply Valve Not Fully Open.Air Supply Valve Replaced
| 10 CFR 50.73(a)(2)(iv), System Actuation |
|