| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
02/25/2013 - 02/26/2013 ** EVENT NUMBERS ** |
 | Agreement State | Event Number: 48757 | Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: THERMO FISHER SCIENTIFIC, INC.
Region: 1
City: FRANKLIN State: MA
County:
License #: 55-0447
Agreement: Y
Docket:
NRC Notified By: JOHN SUMARES
HQ OPS Officer: STEVE SANDIN | Notification Date: 02/15/2013
Notification Time: 14:00 [ET]
Event Date: 02/15/2013
Event Time: 13:00 [EST]
Last Update Date: 02/15/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
ANTHONY DIMITRIADIS (R1DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - LEAKING SEALED NI-63 SOURCE IN GAS CHROMATOGRAPH
The following information was received from the Commonwealth of Massachusetts via fax:
"The licensee reported to the Agency (MA Radiation Control Board) that a wipe test survey of an SVAC-G module containing a Ni-63 sealed source yielded 0.0068 microcuries. The licensee gave no further details about the sealed source or the device, and stated that a written report was mailed to the Agency. The SS&D registration certificate indicates the sealed source maximum activity is 5 millicuries. The SS&D certificate indicates the source is contained inside the 'Thermo Fisher Scientific' model SVAC-G module which is contained in a 'Thermo Fisher' model EGIS Defender gas chromatograph instrument." |
| Non-Agreement State | Event Number: 48759 | Rep Org: DEPARTMENT OF VETERANS AFFAIRS
Licensee: CARL VINSON VA MEDICAL CENTER
Region: 1
City: DUBLIN State: GA
County:
License #: 03-23853-01VA
Agreement: Y
Docket:
NRC Notified By: THOMAS HUSTON
HQ OPS Officer: STEVE SANDIN | Notification Date: 02/15/2013
Notification Time: 16:56 [ET]
Event Date: 02/15/2013
Event Time: 07:00 [EST]
Last Update Date: 02/15/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
20.1906(d)(1) - SURFACE CONTAM LEVELS > LIMITS
| Person (Organization):
PATTY PELKE (R3DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text PACKAGE RECEIVED AT VA MEDICAL CENTER WITH OUTSIDE CONTAMINATION EXCEEDING LIMITS
"[The Department of Veterans Affairs, Veterans Health Administration (VHA) National Health Physics Programs] has a master materials license from the Nuclear Regulatory Commission. License number is 03-23853-01VA.
"Per 10 CFR 20.1906(d), [the VA] is reporting receipt of a package of radioactive material with removable surface contamination on the outside of the package greater than NRC reporting limits.
"The package was received today (February 15, 2013) at around 7 AM ET by Carl Vinson VA Medical Center, Dublin, Georgia. The package was checked-in and surveyed around 8:30 AM ET. This medical center holds permit number 10-09569-01 under the VHA master materials license.
"Wipe tests performed on the external surface of the package indicated a removable contamination level of 4190 dpm/cm2 as compared to the regulatory limit of 220 dpm/cm2 for beta-gamma emitters.
"The package contained eight unit dosages of radiopharmaceuticals ranging between 6 and 40 millicuries each of Technetium-99m and was shipped from Cardinal Health in Augusta, Georgia. The inner packaging materials were also found to be contaminated.
"The VA facility Nuclear Medicine Technologist who surveyed the packages immediately notified the vendor/shipper, who serves as the final delivery carrier, about the contaminated package at about 9 AM ET. [The VA National Programs] office was not notified until about 4 PM ET.
"As corrective actions: the packaging materials were bagged and set aside in a restricted area at the VA Medical Center and were reclaimed later in the day by the radiopharmacy.
[The VA has also] notified NRC Region III Project Manager (K. Null) ..."
HOO Note: A similar incident at the same facility was reported on 02/11/13 - see EN #48742. |
| Power Reactor | Event Number: 48779 | Facility: HARRIS
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: JASON SWAIN
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 02/25/2013
Notification Time: 08:56 [ET]
Event Date: 02/25/2013
Event Time: 08:04 [EST]
Last Update Date: 02/25/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
GERALD MCCOY (R2DO)
|
| 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 LOSS OF ASSESSMENT CAPABILITY - RADIATION MONITORS INOPERABLE FOR PRE-PLANNED MAINTENANCE
"This is a non-emergency notification. At 8:04 AM EST on February 25, 2013 radiation monitor RM-21AV-3509-1, Plant Vent Stack Wide Range Gas Monitor (WRGM) was declared inoperable for pre-planned maintenance. The maintenance is scheduled to be completed around 6:00 PM EST on 02/27/2013.
"This radiation monitor is necessary for accident assessment and is credited for Emergency Action Level (EAL) classification in the Harris Nuclear Plant Emergency Plan. Inability to classify an EAL due to an out of service monitor is considered a loss of accident assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2. The Technical Support Center-Site Emergency Coordinator has been notified about this preplanned maintenance.
"This condition does not affect the health and safety of the public or the operation of the facility.
"The NRC Resident Inspector has been notified." |
| Power Reactor | Event Number: 48781 | Facility: FORT CALHOUN
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: (1) CE
NRC Notified By: KLINT KUDLACEK
HQ OPS Officer: STEVE SANDIN | Notification Date: 02/25/2013
Notification Time: 16:01 [ET]
Event Date: 02/24/2013
Event Time: 17:00 [CST]
Last Update Date: 02/25/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
DON ALLEN (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text INVERTERS POTENTIALLY INOPERABLE DURING EMERGENCY DIESEL GENERATOR OPERATION
"During a review of the plant inverters, it has been determined that the inverters may not have been operable. The inverters were replaced during the 2008 refueling outage. It appears that either the modification did not recognize that the diesel frequency range is wider than the new inverters, or did not recognize its consequence. Consequently, when the diesel is supplying power to the buses and loads are being sequenced onto the bus, the bus frequency exceeds the inverter frequency range. This causes inverter voltage transients. Operation of the inverters has been modified to improve plant reliability. This issue was discovered during scheduled plant testing of the electrical system."
The licensee informed the NRC Resident Inspector. |
| Power Reactor | Event Number: 48782 | Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [ ] [ ] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: NEEL SHUKLA
HQ OPS Officer: VINCE KLCO | Notification Date: 02/25/2013
Notification Time: 17:49 [ET]
Event Date: 02/25/2013
Event Time: 13:13 [CST]
Last Update Date: 02/25/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
GERALD MCCOY (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | A/R | Y | 92 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC SCRAM DUE TO A TURBINE TRIP FROM A LOSS OF CONDENSER VACUUM
"At 1313 [CST] on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system.
"All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 [CST] with positive indication of isolation, and isolation valves deactivated at time 1514 [CST] to satisfy TS LCO 3.6.1.3 required actions.
"This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A).
"At 1415 [CST], Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 [CST], water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours.
"This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.'
"The NRC Resident Inspector has been notified."
All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2. | |