| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/28/2009 - 05/29/2009 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 44957 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: RANDY SAND
HQ OPS Officer: VINCE KLCO | Notification Date: 04/02/2009
Notification Time: 13:30 [ET]
Event Date: 04/02/2009
Event Time: 05:43 [CDT]
Last Update Date: 05/28/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
STEVE ORTH (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text RESIDUAL HEAT REMOVAL INOPERABLE
"At 0543 [CDT] on April 2, 2009 at the Monticello Nuclear Generating Plant (MNGP) an Operator made the following discovery during performance of his rounds. The flowrate and pressure of the #14 Residual Heat Removal Service Water (RHRSW) pump motor cooling appeared to be low. Investigation found the flow to be approximately 1 gpm. The cooling water supply flow to the pump motor cooler comes from either RHRSW or Service Water. Based on these indications, Operators declared the RHR Shutdown Cooling inoperable and entered actions for Technical Specifications 3.9.7. Actions have been completed to provide an alternate water supply to the RHRSW pump motor oil cooler. Based on the inoperability of the RHR Shutdown Cooling, this event is reportable under 10 CFR 50.72(b)(3)(v), 'An Event or Condition that could have Prevented the Fulfillment of a Safety Function-Capability to Remove Residual Heat.' The station is performing further investigation into the event and will develop corrective actions based on the results of the investigation."
The Licensee notified the NRC Resident Inspector.
* * * RETRACTION FROM RANDY SAND TO JOE O'HARA AT 1149 EDT ON 5/28/09 * * *
"Monticello is retracting the event reported based on further evaluation. An investigation of the event found test data that demonstrates the RHRSW pump would not have lost its ability to provide cooling water to the shutdown cooling system and therefore no loss of safety function occurred. The test data provides documentation that the thrust bearing oil bath temperature of the RHRSW pump motor would not have exceeded the 200 deg F limit imposed by the motor supplier (GE) at the flow rate found by the operator. The test data indicated with the cooling water at a flow rate of less than 0.9 GPM, at 65 deg F the service water temperature and flow would be sufficient to maintain the motor oil bath temperature below 200 deg F. During the event, the actual event parameters (cooling water flow rate >1 gpm and temperatures< 65 deg F) were less severe than the test parameters and therefore are bounded by the test.
"Since there was no impact on the RHRSW system's ability to provide cooling water to the RHR system, the RHR system maintained the ability to provide shutdown cooling and residual heat removal. Therefore the event can be retracted since the condition that was reported in the initial event notification report would not have resulted in the prevention of the fulfillment of a safety function (residual heat removal)."
The licensee notified the NRC Resident Inspector and will notify the State of Minnesota.
Notified R3DO(Lara) |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Hospital | Event Number: 45089 | Rep Org: ST. JOHNS MERCY MEDICAL CENTER
Licensee: ST. JOHNS MERCY MEDICAL CENTER
Region: 3
City: ST. LOUIS State: MO
County: ST. LOUIS
License #: 24-0079403
Agreement: N
Docket:
NRC Notified By: ROBERT TURCO
HQ OPS Officer: VINCE KLCO | Notification Date: 05/22/2009
Notification Time: 09:51 [ET]
Event Date: 05/21/2009
Event Time: 15:30 [CDT]
Last Update Date: 05/22/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
| Person (Organization):
RICHARD SKOKOWSKI (R3DO)
ANGELA MCINTOSH (FSME)
| Event Text MEDICAL EVENT INVOLVING A DOSE THAT IS DIFFERENT THAN PRESCRIBED
The patient was undergoing brachytherapy treatment of the prostate. After 6 seeds were implanted, the procedure was aborted because of concerns in placing additional needles into the patient. The prescribed dose was 145 Gy. However, only 6 of 88 prescribed seeds were implanted. Each seed total activity is 0.288 mCi of I-125.
Family members were notified of the aborted procedure.
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.
* * * UPDATE FROM ROBERT TURCO TO JOHN KNOKE AT 1643 EDT ON 5/22/09 * * *
This event is being retracted because the physician made a choice to terminate the brachytherapy treatment of the prostate to the patient. The physician then rewrote the procedure to the patient to limit the prescribed number of seeds to 6. Having 6 seeds implanted into the patient the prescribed dose was now met under the new directive.
Notified R3DO (Richard Skokowski) , FSME(Lydia Chang) |
| General Information or Other | Event Number: 45095 | Rep Org: WA DIVISION OF RADIATION PROTECTION
Licensee: ANVIL CORPORATION
Region: 4
City: BELLINGHAM State: WA
County:
License #: WN-IR031-1
Agreement: Y
Docket:
NRC Notified By: ARDEN SCROGGS
HQ OPS Officer: DONALD NORWOOD | Notification Date: 05/26/2009
Notification Time: 14:59 [ET]
Event Date: 06/06/2006
Event Time: [PDT]
Last Update Date: 05/26/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
MICHAEL SHANNON (R4DO)
MARK THAGGARD (FSME)
| Event Text AGREEMENT STATE REPORT - DROPPED RADIOGRAPHY DEVICE
The following was received from the State of Washington via e-mail regarding a previously unreported event:
"This is notification of an event in Washington State as reported to or investigated by the WA Department of Health, Office of Radiation Protection.
"A radiography crew was attempting to hoist a Co-60 radiography device to an elevated platform to perform work at an oil refinery. During the process of hoisting the device it slipped out of its rigging and fell an estimated 23 feet to the pavement below. No associated equipment was connected to the device and the plugs were still inserted. The metal flange on the device was dented but there was no other obvious damage. The device is an AEA model 741B, serial number B100. It contained a 5.5 Curie Co-60 Model 424-18 AEA source, serial number 2418.
"A radiation survey was performed by the radiography companies' crew and later by the RSO. The readings were normal for that device. A control cable was attached to the source inside the device and the source was rotated. It was determined to move freely and not bound in the shielded position within the device. Operations with this device were terminated and the device was sent back to its licensed storage location in Burlington, Washington. The radiography company sent the device back to QSA Global for evaluation and maintenance.
"The licensee's RSO stated that repairs were made to a bracket and the handle was replaced. The maintenance was performed. The device was resourced and relabeled and returned to the licensee.
"This incident, given [Washington state] report number WA-06-043 is closed. There was no release of activity. Personnel exposure was kept low so that no exposure limits were exceeded." |
| General Information or Other | Event Number: 45096 | Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: ALLEGHENY RODNEY
Region: 1
City: NEW BEDFORD State: MA
County:
License #: G-0112
Agreement: Y
Docket:
NRC Notified By: BRUCE PACKARD
HQ OPS Officer: JOHN KNOKE | Notification Date: 05/26/2009
Notification Time: 17:17 [ET]
Event Date: 05/26/2009
Event Time: [EDT]
Last Update Date: 05/26/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
GLENN DENTEL (R1DO)
MARK THAGGARD (FSME)
| Event Text AGREEMENT STATE REPORT - GAUGE SHUTTER STUCK OPEN
The following was received from the State of Massachusetts via fax:
"A gauge on the Z-24 press is sometimes stuck open. The equipment is out of use. A licensed contractor was contacted to fix the gauge. Gauge s/n Z3239, model number SS-3a. Radionuclide is 1000 mCi of Am-241." |
| Power Reactor | Event Number: 45098 | Facility: INDIAN POINT
Region: 1 State: NY
Unit: [ ] [3] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: DAN LYON
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/28/2009
Notification Time: 07:12 [ET]
Event Date: 05/28/2009
Event Time: 06:25 [EDT]
Last Update Date: 05/28/2009 | 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):
GLENN DENTEL (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | A/R | Y | 60 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR TRIP RESULTING FROM A HIGH STEAM GENERATOR LEVEL TURBINE TRIP
"At 0530 EDT a power reduction commenced due to elevated vibrations on 32 MBFP [Main Boiler Feedwater Pump]. When reactor power reached 60% the team was attempting to stabilize reactor power when a level excursion occurred in 32 steam generator. When level reached its high level turbine trip set point an automatic reactor trip occurred and all systems responded as expected. Auxiliary feedwater actuated as expected. The Unit is currently in mode 3 and stable with auxiliary feed water in service and reactor temperature maintained with the steam dumps to the main condenser. Investigation into the cause of the steam generator level excursion is in progress. Offsite power is available and supplying safeguard busses. Unit 2 was unaffected and remains at 100% power. The [NRC] Resident Inspector has been notified."
No safety or relief valves lifted during this event. |
| Power Reactor | Event Number: 45099 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BRIAN MATTY
HQ OPS Officer: VINCE KLCO | Notification Date: 05/28/2009
Notification Time: 08:59 [ET]
Event Date: 05/06/2009
Event Time: 13:52 [EDT]
Last Update Date: 05/28/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
GLENN DENTEL (R1DO)
|
| 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 AUTOMATIC ACTUATION OF RPS, ECCS AND EDG GENERATED BY AN INVALID ACTUATION SIGNAL
"On May 6, 2009, Beaver Valley Power Station Unit 1 was in Mode 5 during a refueling outage. At 1352, a Low Steam Line Pressure Reactor Trip and Safety Injection (SI) signal was unexpectedly received on Train A. The Train A Emergency Diesel Generator started upon the SI signal, as designed, but did not load since there was no actual emergency bus low voltage condition. The Train A Charging pump (Emergency Core Cooling Pump) was in service and operating before the generation of the SI signal, and continued to operate following the SI signal. Other plant components not isolated for plant outage conditions properly actuated in response to the generated SI signal.
"A SI Block must be inserted whenever the Solid State Protection System (SSPS) is operating in Mode 5 to prevent a SI signal from being generated on low steam line pressure since steam line pressure will be below the low steam line pressure setpoint. Since the plant was in Mode 5 (temperature less than 200F), there can be no steam in the steam lines. Thus, the receipt of a Low Steam Line Pressure Safety Injection signal while the plant was in Mode 5 was invalid.
"It was determined that the SI Block was removed due to a degraded Steam Line Pressure SI Block/Reset switch on the control room benchboard. Further troubleshooting detected a similar erratic type degradation on the Train A Pressurizer SI Block switch. Both switches were replaced.
"This event is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A) since it involved an actuation of the Reactor Protection System (RPS) per 10 CFR 50.73(a)(2)(iv)(B)(1), an Emergency Core Cooling System (ECCS) pump per 10 CFR 50.73(a)(2)(iv)(B)(3), and an Emergency Diesel Generator (EDG) per 10 CFR 50.73(a)(2)(iv)(B)(8). However, pursuant to 10 CFR 50.73(a)(1), this event is being reported via this telephone notification, instead of
submitting a written Licensee Event Report, since the automatic actuation of the RPS, ECCS Pump and EDG was not generated by a valid actuation."
The licensee notified the NRC Resident Inspector. |
| General Information or Other | Event Number: 45100 | Rep Org: WESTINGHOUSE ELECTRIC
Licensee: WESTINGHOUSE ELECTRIC
Region: 1
City: PITTSBURGH State: PA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: HANK SEPP
HQ OPS Officer: VINCE KLCO | Notification Date: 05/28/2009
Notification Time: 10:15 [ET]
Event Date: 05/28/2009
Event Time: [EDT]
Last Update Date: 05/28/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
GLENN DENTEL (R1DO)
GEORGE HOPPER (R2DO)
JULIO LARA (R3DO)
OMID TABATABAI (NRO)
JOHN THORP (NRR)
| Event Text PART 21 REPORT - AIR CIRCUIT BREAKER DEFICIENCY
The following information was received from Westinghouse via facsimile:
Westinghouse provided information pursuant to the requirements to 10 CFR 21 to report an issue concerning the failure of a Westinghouse DB-100 circuit breaker to close after a new manufacturing cotter pin was installed during refurbishment. Westinghouse has supplied the air circuit breaker at nuclear generating stations.
When a DB-100 air circuit breaker failed to close on demand, the air circuit breaker was returned to Westinghouse to investigate the cause of this failure. The air circuit breaker was successfully cycled several times at Westinghouse before the failure to close was repeated.
Upon discovery of the DB-100 circuit breaker failing to close, Westinghouse identified plants that have purchased safety related DB-75 and DB-100 circuit breakers and reviewed the configuration of available breakers and drawings.
Westinghouse has notified the affected plants which are: R. E. Ginna, Indian Point Unit 2, Point Beach and H. B. Robinson.
Westinghouse will be issuing a Nuclear Safety Advisory Letter (NSAL) documenting this issue. |
| Power Reactor | Event Number: 45103 | Facility: COOPER
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: STEVE WHEELER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/29/2009
Notification Time: 03:37 [ET]
Event Date: 05/28/2009
Event Time: 22:06 [CDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
MICHAEL SHANNON (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 CONTROL ROOM EMERGENCY FILTRATION SYSTEM EXCESSIVE LEAKAGE
"On 28 May 2009 the Control Room Emergency Filtration System (CREFS) was declared inoperable due to a degraded Control Room Envelope (CRE). Two CRE boundary doors were found with excessive leakage. Based on the identified leakage, reasonable assurance that CREFS would fulfill its safety function could not be established.
"The CRE boundary doors support CREFS at CNS. CREFS is a single train system and per 10CFR50.72(b)(3)(v)(D) an 8 hour report is required due to the fact that at the time of discovery this condition could have prevented the fulfillment of the safety function of an SSC that is required to mitigate the consequences of an accident.
"Actions to implement mitigating actions have been initiated in accordance with plant Technical Specifications.
"The NRC Senior Resident Inspector has been notified of the condition." |
| Power Reactor | Event Number: 45104 | Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES SMIT
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/29/2009
Notification Time: 04:49 [ET]
Event Date: 05/28/2009
Event Time: 21:40 [CDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
JULIO LARA (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 LOSS OF CONTROL POWER TO ECCS VALVES
"At 2140 on 5/28/09 Braidwood Station identified a loss of control power for a Safety Related MCC (Motor Control Center) which provided power to SVAG (Single Valve Actuation Group) valves associated with both trains of the ECCS system. The MCC is normally de-energized to maintain the valve power removed in accordance with Tech Specs for ECCS. Loss of the control power for the associated MCC would prevent operation of these valves, which would prevent realignment of components required for transfer to cold leg recirculation and hot leg recirculation for long term core cooling.
"Entry was made into LCO 3.5.2, ECCS Operating, and LCO 3.0.3 due to inoperability of both trains of ECCS based on the inability to realign portions of both trains of the ECCS system from injection to cold leg recirculation and subsequent hot leg recirculation. At 2230 on 5/28/09 preparations had been completed for ramp off line per LCO 3.0.3.
"Troubleshooting was performed and a blown control power fuse was identified and replaced at 2319 on 5/28/09. No Unit ramp was initiated.
"The NRC Resident Inspector was notified." | |