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Event Notification Report for January 6, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
01/05/2006 - 01/06/2006

** EVENT NUMBERS **


42224 42237 42239 42243

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 42224
Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: MICHAEL HIMEBAUCH
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 12/23/2005
Notification Time: 02:22 [ET]
Event Date: 12/22/2005
Event Time: 22:25 [EST]
Last Update Date: 01/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
HIRONORI PETERSON (R3)

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

EECW TEMPERATURE CONTROL VALVE NOT FULLY OPEN

"While performing Div. 1 & 2 Emergency Equipment Cooling Water (EECW)/Emergency Equipment Service Water (EESW) Valve Lineup Verification surveillance on 12/22/05, the temperature control valve (TCV) on both divisions of EECW were found to be approximately 95% open rather than their required full open position. The system design requires that the TCV, or the associated TCV bypass valve, be in the full open position during system startup to avoid a potentially damaging pressure transient from occurring. Both divisions of EECW and all supported systems (including HPCl, both divisions Core Spray, and both divisions of RHR) were declared INOPERABLE at 2225 EST. Multiple LCO Required Actions were entered, including entry into LCO 3.0.3. At 2250 EST, Div. 1 EECW was restored to OPERABLE status by fully opening the TCV bypass valve and isolating the TCV, and LCO 3.0.3 was exited. At 2252 EST, Div. 2 EECW was restored to OPERABLE status by fully opening the TCV bypass valve and isolating the TCV, and all associated LCO Required Actions were exited. Reactor power remained at 100% throughout the event. The NRC resident inspector has been notified. This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition and 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident."

The licensee is investigating the cause of the valve not being fully open. The licensee notified the NRC Resident Inspector.

* * * UPDATE PROVIDED BY YEAGER TO ROTTON AT 1524 ON 01/05/06 * * *

"This is a retraction of NRC Event #42224. Based on further engineering review, it is concluded that no potential damage from a pressure transient would occur as a result of the TCV being approximately 95% open. System startup pressure transient concerns reflected in the operating procedures originated from a previously-experienced pressure transient resulting from void collapse against a closed TCV. System startup with the as-found TCV position still provides a sufficiently-open flow path to preclude void collapse against a closed boundary. Additionally, Engineering has determined that system operation with a 90% open TCV would have no significant impact on total system flow and the cooling function. Therefore, both divisions of EECW and all other supported systems (including HPCI, both divisions of Core Spray, and both divisions of RHR) were operable with the TCV in the approximately 95% open position."

The licensee notified the NRC Resident Inspector. Notified R3DO (Ring).

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Power Reactor Event Number: 42237
Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: RICHARD HARRIS
HQ OPS Officer: STEVE SANDIN
Notification Date: 12/31/2005
Notification Time: 19:16 [ET]
Event Date: 12/31/2005
Event Time: 14:12 [CST]
Last Update Date: 01/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
JEFFREY CLARK (R4)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 95 Power Operation 94 Power Operation

Event Text

AUTOMATIC ACTUATION OF THE EMERGENCY FEEDWATER (EFW) DUE TO A TRANSIENT ON THE "B" MAIN FEEDWATER (MFW) PUMP

"The EFW actuation occurred when the 'B' MFW pump RPM rose then dropped to ~4000 rpm. The MFW pump then recovered immediately. This transient caused EFIC [Emergency Feedwater Initiation Control] to actuate and both EFW pumps received start signals on invalid low SG level from the EFIC low range level instruments. These instruments measure level based upon a dp [differential pressure] across an orifice and are not considered reliable at 95% power and full MFW flow. All other SG level instruments indicate SG level was above EFIC setpoint during the MFW pump transient. This can be concluded by reviewing the OTSG [Once Through Steam Generator] level prior to and during the transient. Prior to the transient the EFIC low range level instruments indicated a level of ~24 inches. During the transient the EFIC low range level instruments indicated as low as 4 inches. However, the Startup Range level only lowered from ~122 inches prior to the transient to ~120 inches.

"Both EFW pumps were immediately overridden and stopped once it was verified this was not an actual under feed condition to the OTSGs. No EFW injection into the OTSGs occurred due to the EFW actuation."

There was no ongoing maintenance at the time which would have explained the "B" MFW pump transient.

The licensee informed the NRC Resident Inspector.

* * * UPDATE EVENT FROM FRED VAN BUSKIRK TO JOE O'HARA ON 1/5/06 AT 0942 * * *

"On 12/31/05, an 8-hour notification (EN# 42237) was made by Arkansas Nuclear One reporting an automatic actuation of Emergency Feedwater (EFW). The report was submitted pursuant to the requirements of 10 CFR 50.72 (b)(3)(iv)(A) Valid System Actuation. The actuation of EFW occurred as a result of a "B" Main Feedwater (MFW) pump transient which caused an invalid low Steam Generator (SG) level signal from the Emergency Feedwater Initiation and Control (EFIC) instrumentation. As discussed in the original event report, the low SG level EFIC instruments do not provide valid indication at 95% power and full MFW flow. As a result of the elevated flow rate during this perturbation, an invalid indication below the low SG level setpoint was produced resulting in the system actuation. All other SG level instrumentation indicated that actual SG levels remained within the normal operating band, confirming that no low level condition existed and that this event represented an invalid actuation. Accordingly, this update revises Event Notification 42237 to be submitted pursuant to 10 CFR 50.73 (a)(2)(iv)(A) and the 60-day Optional 10 CFR 50.73 (a)(1) requirement - Invalid Actuation of EFW. EFID and EFW systems functioned as designed in response to the invalid low SG level signal.

"The original event report stated that there was no EFW injection into the steam generators as a result of the actuation; however, subsequent reviews of historical Safety Parameter Display System (SPDS) data indicated that the electric EFW pump (P-7B) fed the steam generators for approximately 5 seconds during the event."

The licensee notified the NRC Resident Inspector. R4DO (Shannon) notified.

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General Information or Other Event Number: 42239
Rep Org: CALIFORNIA RADIATION CONTROL PRGM
Licensee: TWINING LABORATORIES
Region: 4
City: FRESNO State: CA
County:
License #: 1014-10
Agreement: Y
Docket:
NRC Notified By: KEN FUREY
HQ OPS Officer: BILL HUFFMAN
Notification Date: 01/03/2006
Notification Time: 16:22 [ET]
Event Date: 12/31/2005
Event Time: [PST]
Last Update Date: 01/03/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
MICHAEL SHANNON (R4)
JIM WHITNEY (E-MAIL) (TAS)
MEXICO VIA FAX ()
PATRICIA HOLAHAN (NMSS)

Event Text

CALIFORNIA AGREEMENT STATE REPORT OF A STOLEN TROXER GAUGE

The State provided the following information via facsimile:

The RSO for Twining Laboratories called to inform the State of California Radiologic Health Branch that a Troxler 3430, Serial # 32684, was stolen the evening of New Years Eve (12/31/05). The gauge was located in an employees garage in San Bernadino, CA. The gauge was chained in the bed of a pickup. The tab that locks the lid of the transport box was cut and the gauge, block, and charger were taken. The employee had worked half a day on Saturday and did not want to drive his vehicle all the way back to the company's Bakersfield office.

The police have been notified of the theft.

Although not stated in the report, Troxler gauges typically contain an 8 millicurie Cesium-137 source and a 40 millicurie Americium-241/Beryllium source.

State Report: 010306

THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL

Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.

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Power Reactor Event Number: 42243
Facility: RIVER BEND
Region: 4 State: LA
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: GLENN KRAUSE
HQ OPS Officer: JEFF ROTTON
Notification Date: 01/05/2006
Notification Time: 15:44 [ET]
Event Date: 01/05/2006
Event Time: 09:30 [CST]
Last Update Date: 01/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
MICHAEL SHANNON (R4)

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

UNANALYZED CONDITION INVOLVING RCIC OPERATION DURING MCR EVACUATION

"During an engineering assessment, a condition was found which does not meet the assumptions of the River Bend post-fire safe shutdown analysis. 10CFR50 Appendix R states that for alternate shutdown capability (i.e. shutdown from outside the main control room), reactor coolant system process variables shall be maintained within those predicted for a loss of normal AC power, and the fission product boundary integrity shall not be affected. Generic Letter (GL) 86-10, 'Implementation of Fire Protection Requirements,' states that the following assumptions are required for evaluation of a control room fire: 1) fire-induced spurious operation of safe shutdown components has occurred; 2) offsite power is lost; and, 3) the emergency diesel generators (DGs) do not automatically start.

"Based on the conservative assumptions imposed by GL 86-10, the following control room fire scenario must be addressed. A fire is assumed to cause motor-operated valve E51-MOVF063, the inboard steam supply to Reactor Core Isolation Cooling (RCIC) turbine to close. The same fire requires the main control room (MCR) to be evacuated, and during relocation to the Division 1 Remote Shutdown panel, offsite power is lost.

"The post-fire safe shutdown analysis has evaluated RCIC to be available from the Remote Shutdown Panel in order to maintain reactor water level, and that the Division 1 and 3 DGs are started locally. The Division 2 DG is not analyzed to remain free of damage caused by the MCR fire. Since valve E51-MOVF063 is powered from Division 2, and there is no Division 2 power available to re-open the valve, steam would not be available to power the RCIC turbine. E51-MOVF063 is located in the drywell, making manual operation of the valve impractical. Therefore, RCIC is postulated to not be available to maintain reactor level. Establishing reactor level control is a time-critical function that is required to occur within ten minutes of MCR evacuation in order to meet one of the Appendix R safe shutdown performance goals.

"This condition involves compliance with 10CFR50, Appendix R. Plant equipment remains operable. The scope of this analysis deficiency is limited to the MCR fire scenario, with three concurrent failures. The MCR Is continuously manned. The affected cables in the MCR under-floor area are protected by fire detection and automatic suppression systems, which would rapidly detect and smother a fire. Introduction of ignition sources, such as work involving welding or grinding, is strictly controlled by station procedures.

"While the assumptions of the post-fire safe shutdown analysis are not met for this scenario, it has been verified that the components required to properly align the Division 1 Residual Heat Removal system in the low pressure coolant injection mode would be available at the Division 1 Remote Shutdown Panel. Control of three safety-relief valves is also available at the Division 1 Remote Shutdown Panel to depressurize the reactor vessel for low pressure injection. An analysis is under way to determine the response of reactor water level, given these conditions."

The licensee notified the NRC Resident Inspector.

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