| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
10/07/2010 - 10/08/2010 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 46258 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: KENNY HUNTER
HQ OPS Officer: DONG HWA PARK | Notification Date: 09/18/2010
Notification Time: 17:20 [ET]
Event Date: 09/18/2010
Event Time: 16:27 [EDT]
Last Update Date: 10/07/2010 | Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
| Person (Organization):
JAY HENSON (R2DO)
ALLEN HOWE (NRR)
WILLIAM GOTT (IRD)
ERIC LEEDS (NRR)
LUIS REYES (R2)
STANSFELD (DHS)
VIA (FEMA)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 65 | Power Operation | 65 | Power Operation | Event Text UNUSUAL EVENT DECLARED DUE TO HIGH-HIGH REACTOR BUILDING SUMP LEVEL ALARM
"This is a one-hour report for the discovery of a condition that met an Emergency Activation Level (EAL) for a Notification of an Unusual Event [NOUE].
"EAL HU-1 - Natural and Destructive Phenomena Affecting the Protected Area Threshold Value 6 states, 'Exceed Max Normal Operating Values specified in EOP 31 EO-EOP-014-1 SC - Secondary Containment Control Table 5 Secondary Containment Operating Water Levels.'
"Plant Hatch Unit 1 declared a NOUE on 9/18/10 at 1627 based on HU-1 Natural and Destructive Phenomena Affecting the Protected Area due to exceeding the Max Normal Operating Value for the Reactor Building NE Diagonal Floor Drain Sump. At 1615, both the Level High and Level High-High annunciators alarmed. Water has not overflowed the sump. However, due to its design, it is not possible to immediately visually confirm the sump level. The sump pumps were not running and would not run with their switches in START indicating either: (1) there is no water in the sump, (2) both pumps are OOC [Out of Commission], or (3) there is a problem in the sump level control system. Maintenance is assessing the situation."
The licensee has notified the state and local agencies and will notify the NRC Resident Inspector.
* * * UPDATE FROM STEVE BURTON TO HOWIE CROUCH @ 0427 EDT ON 9/19/10 * * *
At 0400 EDT, the NOUE was terminated. The termination criteria was that the NE Diagonal Floor Drain Sump level was in its normal operating range with no abnormal sump in-leakage detected. Compensatory measures are in place to measure level in the sump. An investigation into the abnormal level indication is still in progress.
The licensee will be notifying state and local authorities as well as the NRC Resident Inspector. Notified R2DO (Henson), IRD (Gott), NRR EO (Howe), DHS (Inser) and FEMA (Via).
* * * RETRACTION FROM STEVE BURTON TO JOE O'HARA AT 1533 EDT ON 10/7/10 * * *
"On September 18, 2010, a notification, Event Number 46258, was submitted to the NRC to report a circumstance at Plant Hatch Unit 1 which appeared to be a 'High-High' reactor building sump level alarm. Such an alarm would meet the designated Emergency Activation Level (EAL) for a Notification of an Unusual Event (NOUE). On the day of the event, it was thought that receipt of both the 'High' and 'High-High' level annunciators for the reactor building northeast diagonal floor drain sump was indicative of the sump level having reached its maximum normal operating level, and a NOUE was made.
"During a subsequent Investigation, engineering personnel determined that the 'High' and 'High-High' level switches for this sump are wired such that both alarms are actuated concurrently when the 'High' level in the sump is reached at approximately 52 Inches. The 'High-High' sump level annunciation (the level at which the NOUE must be made) Is supposed to alarm at approximately 60 inches. Engineering personnel also determined that the level logic is currently improperly configured to clear both the 'High' and 'High-High' level alarms when the sump level rises to 60 inches.
"The initial 'High' and 'High-High' level alarms in the main control room were received at approximately 1615 EDT on September 18, 2010. At that time the sump level would have reached an actual level of approximately 51 inches (albeit with a false indication of approximately 60 inches due to the configuration error noted above) based on the setpoint at which this alarm occurs. The normal leakage into this sump at the time of this event was at a rate of approximately 1 inch per hour. The sump pumps were started at approximately 1800 EDT to pump down the sump. This action resulted in a maximum actual sump level of approximately 53.5 inches, assuming the normal leakage into the sump was occurring during this period. At no time did the sump level actually reach the 'High-High' sump level criteria of 60 inches, nor did the annunciated level clear prior to pumping down the sump which also demonstrated that the 60 inch criteria was not met.
"It is now apparent that the initial NOUE was a conservative report that was properly made based on the information available to the operators at the time. The subsequent investigation provided additional information regarding actual sump level conditions including information that both 'High' and 'High-High' alarms annunciate at the 'High' sump level and would have cleared had the 'High-High' level been reached. Since the 'High-High' sump maximum normal operating level of 60 inches was never reached during this event, the determination has been made that the EAL was not appropriate given the actual sump conditions, thereby making the event non-reportable and an NOUE unnecessary.
"Based on the preceding Information, Southern Nuclear Company (SNC) hereby provides notification that Event Number 46256 is retracted."
The licensee will notify the NRC Resident Inspector.
Notified R2DO(McCoy). |
| 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: 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. |
| Power Reactor | Event Number: 46315 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: JOSHUA SISTAK
HQ OPS Officer: JOE O'HARA | Notification Date: 10/07/2010
Notification Time: 12:39 [ET]
Event Date: 10/07/2010
Event Time: 10:41 [EDT]
Last Update Date: 10/07/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
JUDY JOUSTRA (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 LOSS OF 4160 VOLT SAFETY BUS
"Oyster Creek experienced a loss of 4160 volt 'D' safety bus. Emergency diesel generator number 2 fast started and re-powered the 'D' bus, as expected. All other safety systems responded as expected. Investigation into the loss of 'D' 4160 volt safety bus is in progress."
The licensee was performing grid undervoltage testing on the 'D' 4160 at the time of the event. The licensee has staffed the Outage Control Center. There is no affect on the other 4160 volt bus. Offsite power is normal. Plant risk was 'yellow' prior to and following the event as a result of Reactor Building Closed Cooling Water Pump 1-2 out of service for planned maintenance.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 46317 | Facility: ROBINSON
Region: 2 State: SC
Unit: [2] [ ] [ ]
RX Type: [2] W-3-LP
NRC Notified By: GARRETT SANDERS
HQ OPS Officer: CHARLES TEAL | Notification Date: 10/07/2010
Notification Time: 18:08 [ET]
Event Date: 10/07/2010
Event Time: 13:15 [EDT]
Last Update Date: 10/07/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
GERALD MCCOY (R2DO)
|
| 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 FEEDWATER ISOLATION INADVERTENTLY DISABLED
"At about 1315 hours EDT on October 7, 2010, with the unit in Mode 3 and operators performing recovery actions following a reactor trip that had occurred at 0013 hours (see EN#46313), it was discovered that actions that had been performed to restore the main feedwater system had inadvertently resulted in disabling the feedwater isolation function. The feedwater isolation function, as described in Technical Specifications Section 3.3.2, Table 3.3.2 1, Function 5, requires that the feedwater isolation function be operable in Modes 1, 2, and 3, when the feedwater system is not isolated by the main feedwater isolation valves, main feedwater regulating valves, and bypass valves or by a closed manual valve. At approximately 1018 hours, during actions being taken to restore operation of the main feedwater system, the feedwater isolation key switches for the three steam generators, A, B, and C, were placed in the override/reset position. Although it was not realized at that time, this action was contrary to the Technical Specifications Section 3.3.2 operability requirements for the feedwater isolation function. This inoperability of the feedwater isolation function would have prevented the automatic feedwater isolation function described in the basis of Technical Specifications Section 3.3.2, which states that the primary function of the feedwater isolation signal is to stop excessive flow of feedwater into the steam generators. It also states that this function is necessary to mitigate the effects of overfeeding the steam generators, which could result in overcooling of the primary system. This function is actuated by a safety injection signal. There is no Technical Specifications allowed condition for both trains of the feedwater isolation function to be inoperable. Therefore, Technical Specifications Limiting Condition for Operation (LCO) 3.0.3 was applicable from the time the feedwater isolation switches were placed in the override/reset position, until the feedwater isolation function operability was restored at approximately 1329 hours. The LCO 3.0.3 completion time to be in Mode 4 within 13 hours was not exceeded. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D), any event or condition that at the time of discovery could have prevented the fulfillment of a safety function."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 46318 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: DARRELL LAPCINSKI
HQ OPS Officer: JOE O'HARA | Notification Date: 10/07/2010
Notification Time: 21:06 [ET]
Event Date: 10/07/2010
Event Time: 13:52 [CDT]
Last Update Date: 10/07/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| 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 | Y | 98 | Power Operation | 98 | Power Operation | Event Text DEGRADED FIRE BARRIER
"During a walkdown for the Fire Penetration Seal Project a degraded fire barrier was identified in the wall between the Unit 1 and Unit 2 Normal 480V Switchgear Rooms. The wall is listed as an Appendix R wall between Fire Area (FA) 37 and FA 38. The wall separates redundant safe shutdown cables.
"There is a 2 inch gap where the top of the wall meets the ceiling. The gap is filled with combustible foam - some loose and some stationary. This has been identified as a missing fire barrier such that the required degree of separation for redundant safe shutdown trains is lacking.
"A fire watch has been established as a compensatory measure and will remain in place until the fire barrier is repaired.
"The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B).
"The licensee has notified the NRC Resident Inspector of this event." | |