| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
04/16/2010 - 04/19/2010 ** EVENT NUMBERS ** |
 | Other Nuclear Material | Event Number: 45845 | Rep Org: MONTANA DEPARTMENT OF TRANSPORTATIO
Licensee: MONTANA DEPARTMENT OF TRANSPORTATION
Region: 4
City: HOT SPRINGS State: MT
County:
License #: 25-11498-01
Agreement: N
Docket:
NRC Notified By: REX HOY
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/16/2010
Notification Time: 13:21 [ET]
Event Date: 04/15/2010
Event Time: 14:10 [MDT]
Last Update Date: 04/16/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
| Person (Organization):
BLAIR SPITZBERG (R4DO)
ANGELA MCINTOSH (FSME)
| Event Text DAMAGED TROXLER MOISTURE DENSITY
At 1410 MST on 4/15/10, a gauge operator was conducting a density reading near a culvert on Highway 28 in Hot Springs, MT. After placing the gauge at the target areas, the operator walked about twenty feet away. While he was away, a compactor that was compacting the road bed, slipped off the road near the culvert. When the compactor operator was backing up to get back on the road, he backed over the density gauge.
The gauge operator immediately secured the area and contacted the Radiation Safety Officer (RSO). The RSO visually inspected the gauge and determined that both sources were retracted into the shielded position and appeared to be intact. He then surveyed the area and determined that no radiation release had occurred. The RSO conducted a wipe test and determined that no source leakage occurred. There were no personnel overexposures during this event. The gauge is currently in storage at the Montana Department of Transportation awaiting transport to Troxler for disposal.
The gauge was a model 3440, serial number 36152. It contained a Cs-137 source (serial number 77-3287, 8 mCi on 11/8/04) and an Am/Be-241 source (serial number 78-1047, 40 mCi on 10/13/04). |
| Power Reactor | Event Number: 45847 | Facility: SALEM
Region: 1 State: NJ
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: ERIC POWELL
HQ OPS Officer: VINCE KLCO | Notification Date: 04/16/2010
Notification Time: 17:07 [ET]
Event Date: 04/16/2010
Event Time: 16:49 [EDT]
Last Update Date: 04/16/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
MEL GRAY (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFSITE NOTIFICATION DUE TO DISCOVERY OF DIESEL FUEL OIL IN A MONITORING WELL
"At 1649 [EDT], on April 16, 2010, Salem Unit 2 notified the New Jersey Department of Environmental Protection [NJDEP] via the NJDEP hotline of the discovery of fuel oil from a sample well. At 1645 [EDT], a sample was pulled from a monitoring well on the north east side of the Unit 2 Auxiliary Building which contained approximately 1 pint of diesel fuel oil. This notification to the State of New Jersey is reportable in accordance with 10 CFR 50.72(b)(2)(xi)."
The source of the diesel fuel oil is under investigation by the licensee.
The licensee notified the NRC Resident Inspector and will notify the State of Delaware. |
| Power Reactor | Event Number: 45848 | Facility: SALEM
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MATT MOG
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/16/2010
Notification Time: 22:08 [ET]
Event Date: 04/16/2010
Event Time: 16:41 [EDT]
Last Update Date: 04/16/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
MEL GRAY (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text UNPLANNED EMERGENCY DIESEL GENERATOR START DUE TO LOSS OF A 4160 V BUS
"At 1641 hours [EDT] on April 16, 2010, an [unplanned] automatic start signal was generated for the 1C Emergency Diesel Generator (EDG). Surveillance testing was being performed on the 1C 4160 volt bus to verify the ability to swap between the 13 and 14 Station Power Transformers (SPT) which are the offsite power in-feeds for the 1C 4160 volt bus. The [failure to] swap from the 13 SPT to 14 SPT resulted in a loss of power to the 1C 4160 volt vital bus generating the automatic start signal for the 1C EDG. The 1C EDG started but the output breaker did not close. Abnormal operating procedure S1.OP-AB.4KV-0003 was entered for loss of the 1C 4160 volt vital bus. Salem Unit 1 was in Mode 6 with core reload (fuel movement) in progress. Fuel movement was stopped as a result of the loss of the 1C 4160 volt bus. All fuel assemblies were placed in a safe position. There was no impact to shutdown cooling and no loss of spent fuel pool cooling. Core reload has been suspended until power is restored to the 1C 4160 volt bus. Initial investigations determined the suspected cause to be a failed auxiliary contact associated with a SPT in-feed breaker. This prevented the opposite in-feed and EDG breaker from closing. Extent of condition review and repairs are in progress."
Fuel movement was suspended due to loss of the fuel handling building exhaust fan which is powered by the 1C 4160V bus. There was no further impact on any safety related equipment required in Mode 6.
The license will notify the NRC Resident Inspector, Lower Alloways Creek Township, the State of New Jersey, and the State of Delaware. |
| Power Reactor | Event Number: 45849 | Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: BROCK LINDAU
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/16/2010
Notification Time: 23:35 [ET]
Event Date: 04/16/2010
Event Time: 20:30 [CDT]
Last Update Date: 04/17/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
BLAIR SPITZBERG (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text TECHNICAL SUPPORT CENTER INOPERABLE DUE TO HVAC FAILURE
"At 2030 CDT [on 4/16/10], the Technical Support Center ventilation system was found not running. The cause for the failure of the Technical Support Center ventilation system is not known at this time. This condition renders the Technical Support Center unavailable for emergency planning responses. Alternate facilities are available. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for Loss of Emergency Preparedness Capabilities."
The licensee notified the NRC Resident Inspector.
* * * UPDATE FROM DAVID SPARGO TO HOWIE CROUCH @ 1321 EDT ON 4/17/10 * * *
"As of April 17, 2010 at 12:00 CDT, the Technical Support Center is once again available for Emergency Planning Responses. The cause of the failure was a broken belt on the TSC Air Handling Unit. This belt has been replaced."
The licensee notified the NRC Resident Inspector. Notified R4DO (Spitzberg). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 45850 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: AL MANNING
HQ OPS Officer: DONALD NORWOOD | Notification Date: 04/17/2010
Notification Time: 01:29 [ET]
Event Date: 04/16/2010
Event Time: 19:17 [EDT]
Last Update Date: 04/17/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 86 | Power Operation | Event Text LOSS OF COOLING ACCIDENT SIGNAL DUE TO HIGH DRYWELL PRESSURE SIGNAL
"On April 16, 2010 at 1917 hrs., Unit 1 received an ECCS [emergency core cooling system] loss of cooling accident (LOCA) signal on high drywell pressure. Based on plant data, drywell pressure reached a maximum pressure of approximately 1.25 psig, which is below the LOCA and RPS [reactor protection system] signal actuation pressure of 1.85 psig. At this time, the cause of the drywell pressure increase is under investigation. RPS logic did not initiate due to drywell pressure not reaching the actuation setpoint of 1.85 psig. Although the ECCS logic prematurely actuated, the signal is being treated as 'valid' for the ECCS actuation until further investigation is completed. All expected ECCS actions occurred as a result of the signal. The LOCA logic has been reset and all affected systems have been returned to normal or standby configuration.
"As a result of the LOCA system actuations, several cooling tower fans tripped and condenser vacuum began to decrease. Reactor power was reduced to approximately 86 percent as a result of decreasing condenser vacuum. Power is being maintained at approximately 86 - 88 percent at this time. There are no other plant issues or concerns at this time."
The licensee notified the NRC Resident Inspector.
According to the licensee, normal drywell operating pressure is .5 to 1.2 psig. Prior to the event, drywell pressure had been steady at approximately 1.0 psig. Current drywell pressure is .6 psig. According to the licensee, ECCS systems that started (but did not inject) included: Core Spray pumps, Residual Heat Removal pumps, High Pressure Coolant Injection pump, and the Diesel Generators. An event review team is assessing this event to determine the root cause.
* * * RETRACTION FROM GRIFFIS TO KLCO ON 4/17/2010 AT 1719 EDT* * *
"On April 16, 2010, Hatch Unit 1 received a LOCA ECCS initiation from a high drywell pressure signal. Based on the information available at that time, a notification was made to the NRC assuming the signal to be valid until further investigation could be completed.
"After further review, the determination has been made that the initiation signal originated from a faulted ATTS [Analog Transmitter Trip System] card and not from a valid high drywell pressure condition. Based on this information, this condition did not require an NRC notification in accordance with 10CFR50.72 and, as such, is being retracted through this updated response."
The licensee notified the NRC Resident Inspector.
Notified the R2DO (Hopper) |
| Power Reactor | Event Number: 45851 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [ ] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: TERRY BACON
HQ OPS Officer: DONALD NORWOOD | Notification Date: 04/17/2010
Notification Time: 02:19 [ET]
Event Date: 04/16/2010
Event Time: 22:37 [CDT]
Last Update Date: 04/17/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
STEVE ORTH (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | A/R | Y | 13 | Power Operation | 0 | Hot Standby | Event Text REACTOR TRIP DUE TO TURBINE TRIP
"During a normal reactor shutdown for a refueling outage on Unit 2, a reactor trip occurred at approximately 13% power. This reactor trip was the result of a turbine trip due to high differential pressure between A & B condensers (greater than 2.5 inches). The cause for the vacuum difference between condensers is unknown at this time.
"The reactor trip first actuated indication was a high flux rate trip and the turbine trip first out indication was not received. The reason for this difference is unknown at this time.
"The operating crew responded to the reactor trip utilizing emergency operating procedures for reactor trip and reactor trip recovery and transitioned into normal shutdown procedures.
"All rods inserted as expected and all other systems operated as expected."
The licensee notified the NRC Resident Inspector.
According to the licensee: The plant is in a normal post-trip electrical lineup. No automatic relief valve operations occurred. The motor driven auxiliary feed pump was manually started. The main steam isolation valves are open and decay heat is being removed by steam through the turbine bypass valves to the main condenser. |
| Power Reactor | Event Number: 45852 | Facility: SALEM
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: PATRICK MARTIN0
HQ OPS Officer: STEVE SANDIN | Notification Date: 04/18/2010
Notification Time: 09:30 [ET]
Event Date: 04/18/2010
Event Time: 07:50 [EDT]
Last Update Date: 04/18/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
MEL GRAY (R1DO)
ROBERT NELSON (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text OFFSITE NOTIFICATION DUE TO FATALITY DURING MEDICAL TRANSPORT
"On 4/18/10, Salem management was notified that a contract employee working 1R20 (Salem Unit 1 outage) was complaining of respiratory issues. Site protection was contacted at about 0700 hours and they immediately responded to the control point area where the employee was resting. The employee was evaluated by site protection and immediately sent via ambulance to Salem County hospital (0721 hours). While enroute to the hospital the contract employee died in the ambulance (0750 hours).
"There were no other injuries associated with this event.
"There were no affects on plant systems or safety features as a result of this event.
"A notification has been made to the NRC Resident Inspector in accordance with 11.8.2.a."
The contract employee was not working in a radiological control area (RCA) at the time and there was no personnel contamination associated with this incident.
The licensee will inform local agencies (Lower Alloways Creek Township) and the states of New Jersey and Delaware including the New Jersey State Police. |
| Power Reactor | Event Number: 45853 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: STEVEN INGALLS
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/18/2010
Notification Time: 16:57 [ET]
Event Date: 04/18/2010
Event Time: 13:00 [CDT]
Last Update Date: 04/18/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
STEVE ORTH (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFSITE NOTIFICATION DUE TO LOSS OF CONDENSER TUBE CLEANING (AMERTAP) BALLS
"During the auto shutdown process of TP 1539, Amertap Ball Check, an equipment malfunction occurred when the ball catching flap did not auto reposition to the 'catch' position. This resulted in a loss of amertap balls to the external circulating water system. The quantity of balls lost to the environment is unknown but could potentially be the entire quantity of 1500 balls.
"The Xcel Energy Environmental Service informed the plant that a report will be made to the Minnesota Pollution Control Agency during business hours on 4/19/2010.
"The NRC Resident Inspector has been informed.
"A news release is not planned."
The Amertap system is out of service until repaired. The licensee will also be notifying the Prairie Island Indian Nation and Goodhue County, MN of this incident. |
| Power Reactor | Event Number: 45854 | Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: RANDALL GOLDEN
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/18/2010
Notification Time: 18:04 [ET]
Event Date: 04/18/2010
Event Time: 13:57 [CDT]
Last Update Date: 04/18/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
BLAIR SPITZBERG (R4DO)
ROBERT NELSON (NRR)
WILLIAM GOTT (IRD)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 11 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP DURING REACTOR STARTUP
"This is a 4-hour Non-Emergency, 10CFR 50.72(b)(2)(iv)(B) notification due to an RPS actuation (scram). Arkansas Nuclear One Unit 1 was manually tripped from 11% power at 1357 hrs. CDT. The cause of the trip was operator judgment due to a small fire reported in the high pressure turbine enclosure at governor valve number 3 by a fire watch stationed at that location and an unrelated failure of P-32C Reactor Coolant Pump 3rd stage seal (upper seal) occurring earlier that afternoon. No additional equipment issues were noted. An extinguishing agent (CO2) was applied within approximately 30 seconds. All control rods fully inserted. No primary safeties lifted. No secondary safeties lifted. Emergency feedwater did not actuate and was not needed. No safety systems actuated. The plant will be cooled down to repair the P-32C Reactor Coolant Pump Seal.
"The NRC Resident has been notified."
There was no effect on Unit 2. The grid is stable with Unit 1 in a normal shutdown electrical lineup. Decay heat is being removed via the steam dumps to condenser using normal feed to the steam generators. The licensee will be notifying the Arkansas Department of Health. |
| Power Reactor | Event Number: 45855 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JOHN KEMPKES
HQ OPS Officer: DONALD NORWOOD | Notification Date: 04/19/2010
Notification Time: 03:12 [ET]
Event Date: 04/18/2010
Event Time: 22:25 [CDT]
Last Update Date: 04/19/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
STEVE ORTH (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LCO 3.0.3 ENTRY AND LOSS OF SAFETY FUNCTION DUE TO LOSS OF TURBINE BUILDING HIGH ENERGY LINE BREAK COMPENSATORY MEASURE
"At 2225 CDT on 4/18/2010, Operations discovered that the Unit 1 Turbine Building Truck Aisle Rollup Door Security Fence was closed. This fence was to be maintained open as the truck aisle is a required drainage path from the Unit 1 Turbine Building to outside in the event of flooding resulting from a High Energy Line Break (HELB). With the expanded metal mesh door [fence] closed, turbine building debris could clog the drainage path and result in a higher than calculated water level being reached for this event. As the final water level cannot be predicted, this represented an unanalyzed condition. The higher water levels would be reached at least one hour after the postulated turbine building HELB event. High water levels could result in a Loss of Safety Function for Unit 1 Emergency Diesel Generators. Auxiliary Feedwater (both units) and DC Electrical Power (both units) if water levels exceed critical heights in the associated rooms.
"The doors [fence] were reopened at 2227 CDT. Unit 1 entered LCO 3.0.3 for this two minute period. With the doors [fence] opened, Unit 1 and 2 LCO conditions were again satisfied.
"The Unit 2 truck aisle was in the assumed condition, and Unit 2 is in Mode 5 so loss of AFW or DC power would not result in an LCO 3.0.3 entry. Unit 2 Emergency Diesel Generators are not affected as the Unit 2 truck aisle drain path would prevent water levels from reaching critical heights.
"The initial investigation determined that a Security Officer had closed the gate at approximately 1855 CDT on 4/18/2010.
"The NRC Resident Inspector has been notified. A press release is not planned." | |