Event Notification Report for May 25, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/24/2000 - 05/25/2000
EVENT NUMBERS
37031370323703337034
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37031
Facility: CATAWBA
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEVIN PHILLIPS
HQ OPS Officer: FANGIE JONES
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEVIN PHILLIPS
HQ OPS Officer: FANGIE JONES
Notification Date: 05/25/2000
Notification Time: 09:21 [ET]
Event Date: 05/25/2000
Event Time: 08:32 [EDT]
Last Update Date: 06/22/2000
Notification Time: 09:21 [ET]
Event Date: 05/25/2000
Event Time: 08:32 [EDT]
Last Update Date: 06/22/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
MARK LESSER (R2)
MARK LESSER (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
AUXILIARY BUILDING FILTERED EXHAUST MAY NOT HAVE BEEN ABLE TO PERFORM ITS DESIGN FUNCTION
The licensee discovered both doors on the vestibule to the '1A' Centrifugal Charging Pump Room not fully closed, this may have prevented the Auxiliary Building filtered exhaust system from maintaining a negative pressure in that room. This would render both trains of Auxiliary Building exhaust inoperable per Technical Specifications 3.7.12 and would require entry into Technical Specification 3.0.3. The doors were closed on discovery which returned the unit to compliance with Technical Specifications.
The licensee will followup with an investigation into when and how the doors came to be open.
The licensee notified the NRC Resident Inspector.
* * * UPDATE ON 6/22/00 @ 1610 BY BURGESS TO GOULD * * * RETRACTION
After further engineering evaluation, it has been determined that the plant was not outside the design basis during this event. Therefore, this notification is being retracted.
The NRC Resident Inspector will be notified by the licensee.
The Reg 2 RDO (Landis) was informed by the NRC Operations Officer.
The licensee discovered both doors on the vestibule to the '1A' Centrifugal Charging Pump Room not fully closed, this may have prevented the Auxiliary Building filtered exhaust system from maintaining a negative pressure in that room. This would render both trains of Auxiliary Building exhaust inoperable per Technical Specifications 3.7.12 and would require entry into Technical Specification 3.0.3. The doors were closed on discovery which returned the unit to compliance with Technical Specifications.
The licensee will followup with an investigation into when and how the doors came to be open.
The licensee notified the NRC Resident Inspector.
* * * UPDATE ON 6/22/00 @ 1610 BY BURGESS TO GOULD * * * RETRACTION
After further engineering evaluation, it has been determined that the plant was not outside the design basis during this event. Therefore, this notification is being retracted.
The NRC Resident Inspector will be notified by the licensee.
The Reg 2 RDO (Landis) was informed by the NRC Operations Officer.
Power Reactor
Event Number: 37032
Facility: MCGUIRE
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: H. M. HARRIS
HQ OPS Officer: JOHN MacKINNON
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: H. M. HARRIS
HQ OPS Officer: JOHN MacKINNON
Notification Date: 05/25/2000
Notification Time: 21:47 [ET]
Event Date: 05/25/2000
Event Time: 20:46 [EDT]
Last Update Date: 05/30/2000
Notification Time: 21:47 [ET]
Event Date: 05/25/2000
Event Time: 20:46 [EDT]
Last Update Date: 05/30/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
Person (Organization):
MARK LESSER (R2)
JOHN HANNON (NRR)
JOSEPH GIITTER (IRO)
MARK LESSER (R2)
JOHN HANNON (NRR)
JOSEPH GIITTER (IRO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
TURBINE RUNBACK TO 50% DUE TO LOSS OF A MAIN FEEDWATER PUMP FOLLOWED BY A REACTOR TRIP ON LOW LOW STEAM GENERATOR WATER LEVEL. ONE AUXILIARY FEEDWATER PUMP HAD TO BE MANUALLY STARTED.
Reactor automatic trip occurred at 2046 ET. Initiating event was steam generator low low water level on "C" steam generator. Prior to the trip Channel 1 vital 120 volt AC power was lost due to the tripping open of EVIA (DC/AC ) inverter AC output breaker. No testing was in progress at this time and the cause of the AC breaker trip is under investigation. When the AC output breaker opened it caused a loss of Channel 1 power. Main feedwater pump "1A" control circuitry interfaces with Channel 1 120 volt AC power and when Main Feedwater pump "1A" control circuitry power was lost main feedwater pump "1A" turbine tripped. The loss of Main Feedwater pump "1A" turbine initiated an automatic main turbine runback to 50% power. After the main turbine runback to 50% power the reactor tripped on steam generator "1C" low low water level. Only one of two motor driven auxiliary feedwater pumps automatically started on steam generator "1C" low low water level. Auxiliary feedwater pump "1A" was manually started approximately 2 minutes after the automatic reactor trip. The licensee is investigating why the "1A" auxiliary feedwater pump did not automatically start. All rods fully inserted into the core and reactor coolant temperature is being maintained at Tave no load condition of 557 degrees F. No PORVs or code safety valves on either the primary or secondary side of the plant opened. All the Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. Offsite electrical grid is stable. The licensee's investigation into the initiating event is continuing.
The NRC Resident Inspector was notified of this event by the licensee.
* * * UPDATE ON 05/26/00 AT 1640 EDT BY HEAFNER TAKEN BY MACKINNON * * *
Only the Turbine Driven Auxiliary Feedwater Pump and the Motor Driven Auxiliary Feedwater Pump "1B" automatically started. The "1A" Motor Driven Auxiliary Feedwater Pump did not automatically start since its control circuitry is supplied electrical power from Channel 1 vital 120 volt AC power, which was lost. As stated in the original event report, Auxiliary Feedwater Pump "1A" was manually started approximately 2 minutes after the automatic reactor trip.
Two of three Primary side Power Operated Relief Valves (PORV) opened and subsequently closed. No secondary side PORVs or code safety valves on either the Primary or Secondary side of the plant opened. Investigation into the event is continuing.
R2DO (Tom Decker) notified.
The NRC Resident Inspector was notified of this update by the licensee.
* * * UPDATE ON 05/30/00 @ 1634 BY DIAL TO GOULD * * *
An ongoing evaluation has determined that Main Feedwater Pump "1A" tripped as a result of high discharge pressure which occurred due to the response of the Feedwater Regulating Valves control circuitry. This circuitry was associated with the failed power supply.
The NRC Resident Inspector will be informed by the licensee. The R2DO (McAlpine) was notified by the NRC Headquarters Operations Officer.
Reactor automatic trip occurred at 2046 ET. Initiating event was steam generator low low water level on "C" steam generator. Prior to the trip Channel 1 vital 120 volt AC power was lost due to the tripping open of EVIA (DC/AC ) inverter AC output breaker. No testing was in progress at this time and the cause of the AC breaker trip is under investigation. When the AC output breaker opened it caused a loss of Channel 1 power. Main feedwater pump "1A" control circuitry interfaces with Channel 1 120 volt AC power and when Main Feedwater pump "1A" control circuitry power was lost main feedwater pump "1A" turbine tripped. The loss of Main Feedwater pump "1A" turbine initiated an automatic main turbine runback to 50% power. After the main turbine runback to 50% power the reactor tripped on steam generator "1C" low low water level. Only one of two motor driven auxiliary feedwater pumps automatically started on steam generator "1C" low low water level. Auxiliary feedwater pump "1A" was manually started approximately 2 minutes after the automatic reactor trip. The licensee is investigating why the "1A" auxiliary feedwater pump did not automatically start. All rods fully inserted into the core and reactor coolant temperature is being maintained at Tave no load condition of 557 degrees F. No PORVs or code safety valves on either the primary or secondary side of the plant opened. All the Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. Offsite electrical grid is stable. The licensee's investigation into the initiating event is continuing.
The NRC Resident Inspector was notified of this event by the licensee.
* * * UPDATE ON 05/26/00 AT 1640 EDT BY HEAFNER TAKEN BY MACKINNON * * *
Only the Turbine Driven Auxiliary Feedwater Pump and the Motor Driven Auxiliary Feedwater Pump "1B" automatically started. The "1A" Motor Driven Auxiliary Feedwater Pump did not automatically start since its control circuitry is supplied electrical power from Channel 1 vital 120 volt AC power, which was lost. As stated in the original event report, Auxiliary Feedwater Pump "1A" was manually started approximately 2 minutes after the automatic reactor trip.
Two of three Primary side Power Operated Relief Valves (PORV) opened and subsequently closed. No secondary side PORVs or code safety valves on either the Primary or Secondary side of the plant opened. Investigation into the event is continuing.
R2DO (Tom Decker) notified.
The NRC Resident Inspector was notified of this update by the licensee.
* * * UPDATE ON 05/30/00 @ 1634 BY DIAL TO GOULD * * *
An ongoing evaluation has determined that Main Feedwater Pump "1A" tripped as a result of high discharge pressure which occurred due to the response of the Feedwater Regulating Valves control circuitry. This circuitry was associated with the failed power supply.
The NRC Resident Inspector will be informed by the licensee. The R2DO (McAlpine) was notified by the NRC Headquarters Operations Officer.
Fuel Cycle Facility
Event Number: 37033
Facility: PORTSMOUTH GASEOUS DIFFUSION PLANT
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: JIM MCCLEERY
HQ OPS Officer: FANGIE JONES
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: JIM MCCLEERY
HQ OPS Officer: FANGIE JONES
Notification Date: 05/25/2000
Notification Time: 23:52 [ET]
Event Date: 05/25/2000
Event Time: 18:00 [EDT]
Last Update Date: 05/26/2000
Notification Time: 23:52 [ET]
Event Date: 05/25/2000
Event Time: 18:00 [EDT]
Last Update Date: 05/26/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
JOHN MADERA (R3)
WAYNE HODGES (NMSS)
JOHN MADERA (R3)
WAYNE HODGES (NMSS)
NRC BULLETIN 91-01, 24 HOUR NOTIFICATION
The following is the faxed report from Portsmouth:
On May 16, 2000, PORTS plant personnel initiated a review of an NRC event (# 36993) submitted by Paducah Gaseous Diffusion Plant (PGDP), to assess applicability at the Portsmouth sight. Errors associated with calibration correction factors used to determine uranium masses were found to have not occurred at the PORTS sight.
On May 25, 2000, after further review by Nuclear Criticality Safety dept. personnel, a violation of double contingency was identified when it was determined that the independence of non-destructive assay (NDA) measurements was not maintained. NCSA-PLANT013.A00 indicated that NDA measurements will be maintained independent. However, calibration of instruments was not maintained independent and therefore the resulting measurements were not independent. As a result, the calibration process failed to ensure that a single failure could not effect two independent mass measurements. Double contingency was violated in cases where two independent uranium mass measurements were required to establish double contingency.
A review of NDA calibration data confirmed that no measurement errors actually existed which would have affected the mass readings taken to establish double contingency. This event is being reported because the independence or mass measurement readings was not established and maintained.
SAFETY SIGNIFICANCE OF EVENTS:
The safety significance of this event is very low. All historical calibrations were reviewed (Ref. POEF-38.340.00.086) and no discrepancies were identified with historical measurements. Furthermore, ongoing laboratory Quality Assurance programs and practices help ensure errors of this nature remain sufficiently unlikely to be relied on as criticality controls.
POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW CRITICALITY COULD OCCUR):
An error in NDA measurements plus uncertainty would have to be low by greater than a factor of two or results in a critical mass of uranium going undetected. if this mass were then subject to the correct geometry, moderation and reflection conditions a criticality could error
CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
Mass is the only controlled parameter. Two independent measurements of mass are required to ensure double contingency. Since independence of NDA of NDA mass measurements is called Into question, only one independent estimate of mass is available. The absence of a second independent NDA mass measurement represents a loss of one double contingency control.
ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE OF CRITICAL MASS):
No uranium was involved in the event. NDA measurements are used to classify cascade deposits as either greater than safe mass or less than safe mass. (Safe Mass is 43.5% of minimum critical mass.) NDA measurements are also used in batching operations to ensure the 235U mass in the final container is less than 350 grams 235U which is less half the minimum critical mass at 100% enrichment and optimum moderation, geometry and reflection conditions.
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES
Nuclear criticality safety controls include independent mass measurements of 235U. Failure to ensure the calibration of the NDA instrumentation was independent lead to a potential failure of the mass measurements.
CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED:
NDA measurements performed for criticality safety purposes have been suspended until independent verification of calibration data requirements is flown into applicable procedures and programs.
The certificate holder notified the NRC Resident Inspector and will notify the DOE representative.
* * * UPDATE ON 05/26/00 AT 2012 ET BY McCLEERY TAKEN BY MACKINNON * * *
Update 1, 5/26/00 The NCSE for the above NCSA failed to specifically address failure in calibration or source checks as part of the contingency event "erroneous uranium analysis or operator error." Therefore, this indicates that the NCSE failed to analyze or bound a credible scenario that could lead to a criticality.
A review of past calibration data revealed no suspect calibration or measurements. No known or suspected error in NDA measurements has been discover DOI-344-00-02, Rev A was issued to ensure NDA measurements supporting NCS activities were suspended until requirements for independent verification of NDA calibration data and factors are flowed into applicable procedures.
R3DO (Monty Phillips) & NMSS EO (Hodges) notified.
The NRC Resident Inspector was notified of this event by the certificate holder.
The following is the faxed report from Portsmouth:
On May 16, 2000, PORTS plant personnel initiated a review of an NRC event (# 36993) submitted by Paducah Gaseous Diffusion Plant (PGDP), to assess applicability at the Portsmouth sight. Errors associated with calibration correction factors used to determine uranium masses were found to have not occurred at the PORTS sight.
On May 25, 2000, after further review by Nuclear Criticality Safety dept. personnel, a violation of double contingency was identified when it was determined that the independence of non-destructive assay (NDA) measurements was not maintained. NCSA-PLANT013.A00 indicated that NDA measurements will be maintained independent. However, calibration of instruments was not maintained independent and therefore the resulting measurements were not independent. As a result, the calibration process failed to ensure that a single failure could not effect two independent mass measurements. Double contingency was violated in cases where two independent uranium mass measurements were required to establish double contingency.
A review of NDA calibration data confirmed that no measurement errors actually existed which would have affected the mass readings taken to establish double contingency. This event is being reported because the independence or mass measurement readings was not established and maintained.
SAFETY SIGNIFICANCE OF EVENTS:
The safety significance of this event is very low. All historical calibrations were reviewed (Ref. POEF-38.340.00.086) and no discrepancies were identified with historical measurements. Furthermore, ongoing laboratory Quality Assurance programs and practices help ensure errors of this nature remain sufficiently unlikely to be relied on as criticality controls.
POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW CRITICALITY COULD OCCUR):
An error in NDA measurements plus uncertainty would have to be low by greater than a factor of two or results in a critical mass of uranium going undetected. if this mass were then subject to the correct geometry, moderation and reflection conditions a criticality could error
CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
Mass is the only controlled parameter. Two independent measurements of mass are required to ensure double contingency. Since independence of NDA of NDA mass measurements is called Into question, only one independent estimate of mass is available. The absence of a second independent NDA mass measurement represents a loss of one double contingency control.
ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE OF CRITICAL MASS):
No uranium was involved in the event. NDA measurements are used to classify cascade deposits as either greater than safe mass or less than safe mass. (Safe Mass is 43.5% of minimum critical mass.) NDA measurements are also used in batching operations to ensure the 235U mass in the final container is less than 350 grams 235U which is less half the minimum critical mass at 100% enrichment and optimum moderation, geometry and reflection conditions.
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES
Nuclear criticality safety controls include independent mass measurements of 235U. Failure to ensure the calibration of the NDA instrumentation was independent lead to a potential failure of the mass measurements.
CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED:
NDA measurements performed for criticality safety purposes have been suspended until independent verification of calibration data requirements is flown into applicable procedures and programs.
The certificate holder notified the NRC Resident Inspector and will notify the DOE representative.
* * * UPDATE ON 05/26/00 AT 2012 ET BY McCLEERY TAKEN BY MACKINNON * * *
Update 1, 5/26/00 The NCSE for the above NCSA failed to specifically address failure in calibration or source checks as part of the contingency event "erroneous uranium analysis or operator error." Therefore, this indicates that the NCSE failed to analyze or bound a credible scenario that could lead to a criticality.
A review of past calibration data revealed no suspect calibration or measurements. No known or suspected error in NDA measurements has been discover DOI-344-00-02, Rev A was issued to ensure NDA measurements supporting NCS activities were suspended until requirements for independent verification of NDA calibration data and factors are flowed into applicable procedures.
R3DO (Monty Phillips) & NMSS EO (Hodges) notified.
The NRC Resident Inspector was notified of this event by the certificate holder.
Power Reactor
Event Number: 37034
Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: CHRIS SERATA
HQ OPS Officer: FANGIE JONES
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: CHRIS SERATA
HQ OPS Officer: FANGIE JONES
Notification Date: 05/26/2000
Notification Time: 02:29 [ET]
Event Date: 05/25/2000
Event Time: 22:54 [EDT]
Last Update Date: 05/26/2000
Notification Time: 02:29 [ET]
Event Date: 05/25/2000
Event Time: 22:54 [EDT]
Last Update Date: 05/26/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iii)(D) - ACCIDENT MITIGATION
Person (Organization):
DAN HOLODY (R1)
DAN HOLODY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 15 | Power Operation | 15 | Power Operation |
FOUR FILTRATION RECIRCULATION AND VENTILATION FANS TRIPPED UNEXPECTEDLY
"On 5/25/00 at 22:54 hours, four operating Filtration, Recirculation. and Ventilation System (FRVS) recirculation fans (A, B, D, E) unexpectedly tripped while they were running for a monthly surveillance. An FRVS ventilation fan, that was also in service, continued to run.
"The FRVS consists of two subsystems, the Recirculation System and the Ventilation System. The FRVS Recirculation System is an Engineered Safety Feature (ESF) System, located inside the Secondary Containment, that reduces offsite doses significantly below 10CFR100 guidelines during a LOCA, refueling accident, or high radioactivity in the Secondary Containment. Upon a Secondary Containment isolation, the FRVS Recirculation System is actuated and recirculates the Secondary Containment air through filters for cleanup. This subsystem is the initial cleanup system before discharge is made via the FRVS Ventilation subsystem to the outdoors. The FRVS Ventilation System is an ESF system, located inside the Secondary Containment, that maintains the building at a negative pressure with respect to the outdoors. The system takes suction from the discharge duct of the FRVS Recirculation system and discharges the air through filters to the outdoors.
"Investigation into the cause of the fan trips has identified a manual damper in the ventilation system ductwork that failed to the closed position. This manual damper is normally open during power operation. This damper is repositioned closed during refueling outages to redistribute ventilation through the Secondary Containment. Immediate actions were taken to return the damper to the open position. Recirculation fans have been returned to service and are operating satisfactorily.
"The ventilation system ductwork was reviewed to identify other manual dampers that could have the same or similar affect on system operation. Two other dampers were identified, their positions have been verified to be correct.
"Hope Creek is presently in Operational Condition 1 at 15% power with the turbine/generator off-line awaiting replacement of the 'C' Main Power Transformer. All safety related equipment is available."
The fans had been operating about 10 hours before tripping, it is suspected that the damper was not adequately secured in the open position when the plant exited the refueling outage. The licensee is investigating.
The licensee notified the NRC Resident Inspector and the local township authorities.
"On 5/25/00 at 22:54 hours, four operating Filtration, Recirculation. and Ventilation System (FRVS) recirculation fans (A, B, D, E) unexpectedly tripped while they were running for a monthly surveillance. An FRVS ventilation fan, that was also in service, continued to run.
"The FRVS consists of two subsystems, the Recirculation System and the Ventilation System. The FRVS Recirculation System is an Engineered Safety Feature (ESF) System, located inside the Secondary Containment, that reduces offsite doses significantly below 10CFR100 guidelines during a LOCA, refueling accident, or high radioactivity in the Secondary Containment. Upon a Secondary Containment isolation, the FRVS Recirculation System is actuated and recirculates the Secondary Containment air through filters for cleanup. This subsystem is the initial cleanup system before discharge is made via the FRVS Ventilation subsystem to the outdoors. The FRVS Ventilation System is an ESF system, located inside the Secondary Containment, that maintains the building at a negative pressure with respect to the outdoors. The system takes suction from the discharge duct of the FRVS Recirculation system and discharges the air through filters to the outdoors.
"Investigation into the cause of the fan trips has identified a manual damper in the ventilation system ductwork that failed to the closed position. This manual damper is normally open during power operation. This damper is repositioned closed during refueling outages to redistribute ventilation through the Secondary Containment. Immediate actions were taken to return the damper to the open position. Recirculation fans have been returned to service and are operating satisfactorily.
"The ventilation system ductwork was reviewed to identify other manual dampers that could have the same or similar affect on system operation. Two other dampers were identified, their positions have been verified to be correct.
"Hope Creek is presently in Operational Condition 1 at 15% power with the turbine/generator off-line awaiting replacement of the 'C' Main Power Transformer. All safety related equipment is available."
The fans had been operating about 10 hours before tripping, it is suspected that the damper was not adequately secured in the open position when the plant exited the refueling outage. The licensee is investigating.
The licensee notified the NRC Resident Inspector and the local township authorities.