| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
10/06/2010 - 10/07/2010 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 46259 | Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: JIM PRIEST
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/20/2010
Notification Time: 00:10 [ET]
Event Date: 09/20/2010
Event Time: 07:00 [EDT]
Last Update Date: 10/06/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
GLENN DENTEL (R1DO)
|
| 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 ERDS AND SPDS OUT OF SERVICE FOR COMPUTER UPGRADE
"On 9/20/2010 at approximately 0700 EDT, the Hope Creek Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be taken out of service for approximately nine days to support a planned modification which will install a new (upgraded) computer system. During this timeframe, ERDS and SPDS will be unavailable. Should the need arise, plant status information will be communicated to the NRC, State and local responders using other available communication systems. SPDS and ERDS are expected to be restored on 9/29/2010. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a 'Major Loss of Assessment Capability'."
The licensee has notified Lower Alloways Creek Township and the NRC Resident Inspector of the planned outage.
* * * UPDATE AT 1540 ON 9/29/2010 FROM DAVID HALL TO ERIC SIMPSON * * *
"SPDS and ERDS systems were removed from service to support planned computer upgrade. Planned work which resulted in the removal from service of the SPDS and ERDS has been completed. The SPDS system has been returned to service. During the ERDS testing, a problem was discovered with the NRC's communication link supplied by Verizon. An update will be provided when ERDS testing is completed."
Notified R1DO (Gray).
* * * UPDATE AT 1306 ON 10/6/2010 FROM DAVID HALL TO MARK ABRAMOVITZ * * *
"The ERDS communication problem was repaired and successfully tested. SPDS and ERDS are now fully functional and all work is complete."
The licensee notified the NRC Resident Inspector.
Notified the R1DO (Joustra) |
| Fuel Cycle Facility | Event Number: 46284 | Facility: NUCLEAR FUEL SERVICES INC.
RX Type: URANIUM FUEL FABRICATION
Comments: HEU CONVERSION & SCRAP RECOVERY
NAVAL REACTOR FUEL CYCLE
LEU SCRAP RECOVERY
Region: 2
City: ERWIN State: TN
County: UNICOI
License #: SNM-124
Agreement: Y
Docket: 07000143
NRC Notified By: RANDY SHACKELFORD
HQ OPS Officer: JOHN KNOKE | Notification Date: 09/28/2010
Notification Time: 12:50 [ET]
Event Date: 09/27/2010
Event Time: 13:41 [EDT]
Last Update Date: 09/28/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
PART 70 APP A (b)(1) - UNANALYZED CONDITION
| Person (Organization):
MARK LESSER (R2DO)
JAMES RUBENSTONE (NMSS)
FUELS GROUP Email ()
| Event Text UNUSUAL BUILDUP OF MATERIAL IN ALUMINUM CENTRIFUGE AREA
"During the unloading of centrifuges in the Building 333 U-Aluminum centrifuge area, a crusty buildup of material (~1/8" thick) was observed on the inside of the centrifuge 'jacket' that contains the centrifuge bowl. The buildup was also observed on the underside of the centrifuge lid ('cake pan'). This level of material buildup was unusual and had not been previously observed. It should be noted that some dusting or spattering had been previously observed. The system is designed with drains on the bottom that are designed to prevent the accumulation of liquid within the centrifuge 'jacket'. There is also a requirement to inspect the 'jacket' when solution is observed draining from the overflows. This was considered an unanalyzed or improperly analyzed condition because the mechanism for buildup of this extent was not considered in the safety analysis (i.e. there was no indication of buildup provided by the overflows).
"The following corrective actions were taken: 1) operations in the affected area were suspended; 2) the area was posted to maintain the integrity of the as-found conditions; 3) the area was inspected by safety personnel; 4) the issue was entered into the internal Problem Identification, Resolution, and Correction System (PIRCS); 5) an Unusual Incident Evaluation was performed; 6) calculations were performed with bounding conditions; 7) photographs were taken of the equipment; 8) the system was scanned to determine U-235 mass (~46 grams U-235); 9) material samples were taken and delivered to the laboratory for analysis; and 10) an investigation has been initiated."
There were no control or control system failures. There were no actual or potential safety consequences to workers, the public, or the environment. No degradations or failures have been identified. The system is currently in a safe and stable condition. An investigation has been initiated.
The licensee has notified the NRC Resident Inspector. |
| General Information or Other | Event Number: 46300 | Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: BAYLOR RADIOSURGERY CENTER
Region: 4
City: DALLAS State: TX
County: DALLAS
License #: 05842
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: ERIC SIMPSON | Notification Date: 10/01/2010
Notification Time: 11:50 [ET]
Event Date: 09/30/2010
Event Time: [CDT]
Last Update Date: 10/01/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
THOMAS FARNHOLTZ (R4DO)
ANGELA MCINTOSH (FSME)
| Event Text TEXAS AGREEMENT STATE REPORT - MEDICAL EVENT INVOLVING GAMMA KNIFE THERAPY
The following was received from the State of Texas via e-mail:
"On October 1, 2010, the Agency [Texas Department of Health] received a phone call from the licensee's Radiation Safety Officer (RSO). The RSO stated that a medical event occurred on September 30, 2010, when, while conducting a single fraction exposure to a patient, the computer screen froze. The patient was immediately removed from the gamma knife device. Approximately five percent of the prescribed dose was delivered. The intended dose was 20 gray to one location, and 15 gray to a second location, both to be delivered simultaneously. The referring physician and patient have been notified of the event.
"The gamma knife is a Leksell Gamma Knife System, model number 24001 containing GE model 43047 sealed cobalt (Co) - 60 sources.
"Additional information will be provided as it is received in accordance with SA-300."
This is Texas Incident #: I-8790.
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Power Reactor | Event Number: 46310 | Facility: COOK
Region: 3 State: MI
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRADDOCK D. LEWIS
HQ OPS Officer: PETE SNYDER | Notification Date: 10/06/2010
Notification Time: 06:45 [ET]
Event Date: 10/06/2010
Event Time: 07:00 [EDT]
Last Update Date: 10/06/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
DAVE PASSEHL (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 | | 2 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text UNAVAILABILITY OF TSC VENTILATION SYSTEM DUE TO SCHEDULED MAINTENANCE
"At 0700 on Wednesday, October 6, 2010, the Cook Nuclear Plant (CNP) Technical Support Center (TSC) air conditioning and charcoal filtration systems have been removed from service for scheduled maintenance.
"Under certain accident conditions the TSC may become unavailable due to the inability of the air conditioning and charcoal filtration systems to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC personnel to the unaffected unit's control room if necessary.
"TSC ventilation system maintenance is scheduled to be completed by 1600 on Wednesday, October 6, 2010.
"The licensee has notified the NRC Resident Inspector."
"This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the loss of an emergency response facility."
* * * UPDATE AT 1738 ON 10/6/2010 FROM BRADDOCK LEWIS TO BILL HUFFMAN * * *
TSC ventilation was returned to service at 1645 EDT. The licensee notified the NRC Resident Inspector.
Notified the R3DO (Passehl). |
| Power Reactor | Event Number: 46311 | Facility: COOK
Region: 3 State: MI
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRADDOCK D. LEWIS
HQ OPS Officer: PETE SNYDER | Notification Date: 10/06/2010
Notification Time: 07:13 [ET]
Event Date: 10/06/2010
Event Time: 00:08 [EDT]
Last Update Date: 10/06/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
DAVE PASSEHL (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Hot Standby | 0 | Hot Standby | Event Text VALID ACTUATION OF AUXILIARY FEEDWATER SYSTEM IN RESPONSE TO VALID STEAM GENERATOR LOW-LOW LEVEL SIGNALS
"At 0008 on Wednesday, October 6, 2010, Cook Nuclear Plant (CNP) Unit 2 had a Valid Automatic Actuation of the Unit 2 Turbine Driven Auxiliary Feedwater Pump. Cook Nuclear Plant Unit 2 Reactor had been manually tripped at 0001 from 14% power per normal operating procedures in preparation for the Unit 2 refueling outage.
"Following the reactor trip, water level in steam generators 23 and 24 lowered to 21% causing an automatic start of the Unit 2 Turbine Driven Auxiliary Feedwater Pump. Manual operator actions were in progress to restore steam generator water levels at the time of the actuation. The Turbine Driven Auxiliary Feedwater Pump operated normally to provide auxiliary feedwater flow to all four Unit 2 steam generators, and steam generator levels were restored to normal post trip values. Prior to the trip of Unit 2 reactor, the Unit 2 East and West Motor Driven Auxiliary Feedwater Pumps were manually started per procedure, as part of the pre-planned reactor trip, to control steam generator water levels. The cause of the low steam generator levels is still under investigation."
The licensee has notified the NRC Resident Inspector. |
| Fuel Cycle Facility | Event Number: 46312 | Facility: PADUCAH GASEOUS DIFFUSION PLANT
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
Region: 2
City: PADUCAH State: KY
County: McCRACKEN
License #: GDP-1
Agreement: Y
Docket: 0707001
NRC Notified By: RON DOCKERY
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/06/2010
Notification Time: 21:40 [ET]
Event Date: 10/06/2010
Event Time: 11:06 [CDT]
Last Update Date: 10/06/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
RESPONSE-BULLETIN
| Person (Organization):
GERALD MCCOY (R2DO)
PETER HABIGHORST (NMSS)
| Event Text 24 HOUR NOTIFICATION UNDER BULLETIN 91-01 CONCERNING COOLDOWN VERIFICATION OF UF6 CYLINDERS
"At 1106 CDT, on 10/06/2010 the Plant Shift Superintendent was notified that the independent verification of cylinder cool down time had not been completed on the following Uranium Hexafluoride (UF6) Cylinders: PP5436, PP5453, PP5389, PP5435, PP5388, PP5424, PP5459, and PP5443 in accordance with NCSA GEN-003.
"NCSA GEN-003 requires that prior to movement of a cylinder from a liquid UF6 cylinder handling area it shall be determined, independently verified, and documented that the required cooling time has passed. The purpose of this requirement is to ensure the cylinder does not contain liquid UF6 before it is moved from a liquid handling area. Upon discovery of the violation, it was determined that the cylinders had, in fact, met the required cool down period prior to movement; however, the independent verification had not been completed.
"Since this independent verification was not completed, double contingency was not maintained. Therefore, this is being reported to the NRC as a 24-hour Event Report in accordance with NRC BL 91-01 Supplement 1.
SAFETY SIGNIFICANCE OF EVENTS
"Although an NCSA control was violated, cylinder integrity was maintained.
POTENTIAL CRITICALITY PATHWAYS (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR)
"A solid UF6 cylinder would have to have been breached and sufficient moderator entered the cylinder in order to support a criticality.
ESTIMATED AMOUNT, ENRICHMENT, FORM OF MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS)
"The assay of any material involved is less than or equal to 5.5 wt. % U235. The cylinders involved were 10 ton cylinders.
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES
"Double contingency is maintained by implementing two independent controls on one parameter (moderation).
"The first leg of double contingency relies on it being unlikely that industrial grade cranes, forklifts, and cylinder haulers would drop an ANSI N14.1 designed cylinder in such a way that it would be breached. This moderation control was maintained.
"The second leg of double contingency relies on independent verification that the required cool down time has passed, prior to moving a cylinder from a liquid cylinder handling area. This control helps ensure that the cylinder does not contain liquid UF6 prior to movement. The independent verification was not performed or documented. Therefore, this moderation control was violated. Upon discovery of the violation, it was determined that the cylinders had, in fact, met the required cool down period prior to movement.
"Double contingency relies on two independent controls on the same parameter. Since one of the two independent controls on moderation was violated, double contingency was not maintained; however, the moderation parameter was maintained.
CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED
"The cool down times for the identified cylinders have been independently verified thus bringing them back into compliance with double contingency."
The NRC Resident Inspector has been notified of this event. |
| Power Reactor | Event Number: 46313 | Facility: ROBINSON
Region: 2 State: SC
Unit: [2] [ ] [ ]
RX Type: [2] W-3-LP
NRC Notified By: GARRETT SANDERS
HQ OPS Officer: VINCE KLCO | Notification Date: 10/07/2010
Notification Time: 04:06 [ET]
Event Date: 10/07/2010
Event Time: 00:13 [EDT]
Last Update Date: 10/07/2010 | 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
| Person (Organization):
GERALD MCCOY (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR TRIP DUE TO A ONE LOOP REACTOR COOLANT LOW FLOW
"At 0013 hours EDT on October 7, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. Reactor Coolant Pump (RCP) 'C' tripped during the event.
"The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. Steam generator and pressurizer Power Operated Relief Valves (PORVs) and Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip.
"The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3.
"The Main Turbine Lube Oil Deluge System actuated without a fire during the event. A fire hose station pipe ruptured after the deluge system actuation. The fire hose station was isolated.
"In addition, the 'B' Main Feedwater (MFW) Pump tripped during the event and the 'A' Main Feedwater pump subsequently tripped due to high steam generator level in the 'C' Steam Generator at about 0023 hours.
"There is currently indication of RCP 'C' Number 2 seal leakage of approximately 2.5 gpm.
"At approximately 0405 EDT the Auxiliary Feedwater (AFW) system actuated due to a trip of MFW Pump 'A' while attempting to start the pump in accordance with procedure GP-004, Post Trip Stabilization. The AFW system actuation signal caused motor-driven AFW Pump 'B' to start. The motor-driven AFW Pump 'A' was already in operation due to the post-trip condition.
"The cause of the MFW Pump 'A' trip is under investigation. This AFW system actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A).
"The cause of the reactor trip is under investigation."
The licensee notified the NRC Resident Inspector. | |