| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/10/2010 - 11/12/2010 ** EVENT NUMBERS ** |
 | General Information or Other | Event Number: 46340 | Rep Org: ASCO VALVE
Licensee: AREVA
Region: 1
City: AIKEN State: SC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: ROBERT ARNONE
HQ OPS Officer: JOE O'HARA | Notification Date: 10/18/2010
Notification Time: 13:09 [ET]
Event Date: 09/18/2010
Event Time: [EDT]
Last Update Date: 11/11/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
VIVIAN CAMPBELL (R4DO)
PART 21 GRP ()
| Event Text POTENTIAL EXTERNAL LEAKAGE IN SOLENOID VALVE
The following notification was received via fax:
"On 9/18/10 a single solenoid valve was returned to ASCO with a reported problem of external leakage at the bonnet area below the coil housing. The valve was returned from Cooper Nuclear Station through AREVA, who was the distributor.
"The returned valve was retested at ASCO. No external leakage was observed when the valve was tested in the de-energized state. However, when the valve was tested in the energized state, the reported leakage was confirmed. The root cause of the leakage was determined to be a misaligned O-ring between the solenoid base sub-assembly and the valve body.
"The customers that were shipped affected valves are being notified of the potential non-conformance. ASCO will recommend the affected valves be returned to be retested in accordance with updated test procedures."
* * * UPDATE FROM ROBERT ARNONE TO JOE O'HARA VIA FAX AT 1308 ON 11/11/10 * * *
"In our continuing investigation of the external leakage in NP8320 solenoid valves (Ref A), ASCO has identified an additional group of valves which could be potentially susceptible to such leakage. The initial review focused on our standard NP8320 products and did not include our special construction valves. However, we expanded our review to include all special as well as standard NP8320 valves. As a result, the quantity of potentially affected valves has increased from 174 to 438. In response to this new information, revised notices were sent to the original customers and the newly identified customers: ASCO Canada, SPX Industries, and Flowserve.
"As of this date, 115 valves have been returned. None of these returned valves has exhibited external leakage when retested per the updated procedure."
Through their expanded review process, ASCO did not identify any additional commercial nuclear power plant customers which had purchased valves susceptible to the leakage issue.
Notified R4DO (Gaddy) and Part 21 Group via e-mail. |
| General Information or Other | Event Number: 46404 | Rep Org: NE DIV OF RADIOACTIVE MATERIALS
Licensee: OMAHA PUBLIC POWER DISTRICT
Region: 4
City: OMAHA State: NE
County:
License #: 01-39-04
Agreement: Y
Docket:
NRC Notified By: WAYNE GOLSDORF
HQ OPS Officer: JOHN KNOKE | Notification Date: 11/08/2010
Notification Time: 17:59 [ET]
Event Date: 11/08/2010
Event Time: 10:15 [CST]
Last Update Date: 11/08/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
VINCENT GADDY (R4DO)
LYDIA CHANG (FSME)
| Event Text AGREEMENT STATE REPORT - PROCESS GAUGE SHUTTER COULD NOT BE LOCKED CLOSED
The following information was reported via fax by the State of Nebraska
"The licensee uses fixed industrial gauges for measuring densities in flyash hoppers at the Stations precipitator building. The industrial fixed gauges contain a Cesium 137 sealed source. The Cesium 137 sources were originally installed in April, 1984 and at the time contained 50 milliCuries per source. The Sources were manufactured and installed by Kay Ray Inc. The gauge is Model No. 7062BP, and housing serial number 17666. The source capsule manufacture was Amersham model No. CDC 800 and K-R Ref. Number serial number 15607-V.
"The maintenance group at North Omaha wanted to replace bags in the Unit 5 Transfer House Tertiary Collector. The Shift Supervisor and Operations was notified of the "Tag-out request" as well as the Laboratory. The first step of the "tag-out" is to secure the Transfer House level detectors, the sealed radiation sources. There are two sources in the Transfer House. Both source shutters closed properly, as indicated by surveys. The keys for the interlock Mechanism for the Primary Collector Tank could not be turned to lock the handle in the Closed Position. (Reference: Attachment 2 of Registry for the Kay-Ray Fixed Gauge Model 7062BP). The shutters for both sources remain operable, and can be opened and closed. This 'event' happened at 10:15 am. No personnel were exposed to radiation during this event." |
| General Information or Other | Event Number: 46406 | Rep Org: OHIO BUREAU OF RADIATION PROTECTION
Licensee: LIMA REFINING COMPANY
Region: 3
City: LIMA State: OH
County:
License #: 31201020001
Agreement: Y
Docket:
NRC Notified By: STEPHEN JAMES
HQ OPS Officer: JOHN KNOKE | Notification Date: 11/09/2010
Notification Time: 12:12 [ET]
Event Date: 10/19/2010
Event Time: [EST]
Last Update Date: 11/09/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
ANN MARIE STONE (R3DO)
GLENDA VILLAMAR (FSME)
| Event Text AGREEMENT STATE REPORT - FIXED GAUGE SHUTTER NOT OPERATING PROPERLY
The following information was reported via email by the State of Ohio:
"A licensed service provider (AHP) visited the licensee's site on 10/19/10 and completed an inspection of six fixed level gauges with the following results:
a) Four device shutters functioned as designed and were secured closed for maintenance activities.
b) One shutter device was inoperable due to the fact that the shutter positioning pin was rusted and could not be removed. Through lubricating and gentle mechanical persuasion the pin was removed and the shutter cycled properly.
c) The last device had a shutter operating handle that turned freely but did not rotate the shield block into the beam path. This device obviously did not function as designed and meets the reporting criteria. Through the use of more aggressive measures that did not include disassembly the shutter was eventually closed.
"These activities were closely monitored by the service provider's representative and at no time were licensee personnel exposed to the direct beam. The shutter on this device is now seized in the closed position. Upon identification of the shutter hanging up the service provider contacted the Bureau by telephone and e-mail.
"All six devices remain bolted in their fixed positions. The licensee intends to have the service provider unbolt and transfer all six devices to a secure storage location on licensee property where they will be properly packaged for transfer for disposal."
The 2 sealed sources are both 50 milliCuries of Cs-137. The manufacturer of the fixed gauge is General Nucleonics, Model # 13250, serial # 114 & 116.
Ohio Item Number: OH100027 |
| Power Reactor | Event Number: 46409 | Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: BENJAMIN GEISS
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 11/10/2010
Notification Time: 13:00 [ET]
Event Date: 11/10/2010
Event Time: 10:56 [EST]
Last Update Date: 11/10/2010 | 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):
RONALD BELLAMY (R1DO)
THEODORE QUAY (NRR)
JANE MARSHALL (IRD)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown | Event Text UNPLANNED AUTOMATIC SCRAM AND HIGH PRESSURE COOLANT INJECTION SYSTEM INITIATION DURING TESTING
"At 10:56 [EST] on Wednesday, November 10, 2010, Nine Mile Point Unit [1] One automatically scrammed from rated power. The cause of the scram was Main Steam Isolation Valve (MSIV) closure. The MSIV closure occurred during Instrument and Control Lo-Lo Level Surveillance Testing. The Lo-Lo Surveillance Test has been secured and all Reactor Protection System (RPS) Level Signals returned to normal. Two of four MSIVs went closed; troubleshooting to follow to determine the cause of the equipment malfunction [unexpected MSIV closure].
"Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated. At Nine Mile Point Unit One, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. At 10:58, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was initially established on the Emergency Condensers (ECS). The MSIVs have been re-opened and the normal heat removal capability has been re-established [to the Main Condensers].
"All off-site power sources remain available [with a normal electrical alignment].
"10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours when a valid actuation of the Reactor Protection
System occurs.
"10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the High Pressure Coolant
Injection System occurs.
"There are no other adverse impacts to the station based on this event."
All control rods inserted and the unit is stable in Mode 3 with reactor pressure and temperature approximately 600 psig and 480 degrees. All other safety systems operated as expected.
The licensee notified the NRC Resident Inspector and the New York State Public Service Commission. |
| Power Reactor | Event Number: 46410 | Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: BOB MURRELL
HQ OPS Officer: JOHN KNOKE | Notification Date: 11/10/2010
Notification Time: 13:49 [ET]
Event Date: 11/10/2010
Event Time: 05:18 [CST]
Last Update Date: 11/10/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
ANN MARIE STONE (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text RHR PUMP TRIPPED WHILE OPERATING IN SHUTDOWN COOLING MODE
"On November 10, 2010, at approximately 0518 hours, with the plant in Mode 5 during Refueling Outage (RFO) 22, the 'A' Residual Heat Removal (RHR) pump tripped while operating in the shutdown cooling mode resulting in an interruption of the primary means of decay heat removal for approximately 30 minutes. During this period the maximum increase in reactor temperature was approximately 2 degrees Fahrenheit with a calculated time to boil of approximately 33.9 hours. There was no loss of decay heat removal due to the fact that both trains of Fuel Pool Cooling system and the Reactor Water Cleanup system remained in service. At the time of this event, the plant was in the process of restoring motive power to MO-1909, Outboard Shutdown Cooling Isolation valve. Motive power had previously been isolated to the valve as part of a preplanned evolution of transferring the power supply to 'B' Reactor Protection System (RPS). Due to a failure to isolate the control power to MO-1909 when RPS power had been transferred, MO-1909 automatically closed when motive power had been restored due the existence of a Primary Containment Isolation System (PCIS) signal that was initiated when 'B' RPS power had been transferred. The closure of MO-1909 resulted in the isolation of the common shutdown cooling pathway, and therefore prevented both the 'A' and the 'B' RHR systems from removing decay heat. Preliminary investigations into this event indicate that the failure to isolate control power to MO-1909 occurred due to an existing procedure deficiency for transferring RPS power supplies.
"As a result of the closure of MO-1909, Operations entered Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal, and Technical Specification (TS) Limiting Condition for Operations (LCO) 3.9.7 Condition A; Required RHR Shutdown Cooling Subsystem Inoperable and performed the required actions of the AOP and TS. At approximately 0547, shutdown cooling was restored when the 'C' RHR pump was placed in shutdown cooling. TS 3.9.7 and AOP 149 were subsequently exited at 0551. During the duration of this event adequate decay heat removal existed as part of the site's Shutdown Risk Management in that two loops of Fuel Pool Cooling were in-service and Feed and Bleed utilizing the Control Rod Drive pumps was available. Additionally, Reactor Water Cleanup was in service and remained in service for the duration of this event. Note that RHR shutdown cooling was considered available during this event due to the fact that there were no component failures associated with MO-1909 preventing it from being immediately re-opened.
"This event is being reported as an event or condition that at the time of discovery could have prevented fulfillment of a safety function of structures or systems that are needed to remove residual heat under 10 CFR 50.72 (b)(3)(v)(B).
"The [NRC] Resident Inspectors have been notified." |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 46414 | Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MIKE DEBOARD
HQ OPS Officer: DONALD NORWOOD | Notification Date: 11/12/2010
Notification Time: 15:44 [ET]
Event Date: 11/12/2010
Event Time: 08:18 [CST]
Last Update Date: 01/20/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
ANN MARIE STONE (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 | 100 | Power Operation | 100 | Power Operation | Event Text CONTROL ROOM OUTSIDE AIR INTAKE NOBLE GAS CHANNEL SETPOINTS NON-CONSERVATIVE
"At 0818 CST, 11/12/10, Radiation Protection determined that the setpoints for the Control Room Outside Air Intake Noble Gas channels are non-conservative. This affects Tech Spec 3.3.7 required monitors 0PR31B, 0PR32B, 0PR33B and 0PR34B (Noble gas channels). The current setpoints for 0PR31B, 0PR32B, 0PR33B and 0PR34B are High Alarm 9.55E-05 microCi/cc and Alert Alarm 9.55E-06 microCi/ml. The calculated required setpoints are High Alarm 6.61E-05 microCi/cc and Alert Alarm 6.61E-06 microCi/cc. This is approximately a 30% decrease in setpoints from the current setpoints. LCO 3.3.7 conditions A and B were entered at 0818, 11/12/10. LCO 3.3.7 conditions A and B required actions completion time is 1 hour. All required actions were complete at 0900 11/12/10, less than 1 hour.
"This report is being made per 10 CFR 50.72(b)(3)(v), Event or condition that could have prevented fulfillment of a safety function, 8 hour non-emergency notification.
"Both trains of radiation monitor setpoints were non-conservative and actuate control room ventilation in emergency mode. The margin available in the control room dose analysis will be reviewed to confirm impact on safety function. The setpoints have been non-conservative since 1999."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION FROM BART KELLER TO JOHN SHOEMAKER AT 1414 EST ON 01/20/11 * * *
"At 0818 CST, 11/12/2010, Radiation Protection determined that the setpoints for the Control Room Outside Air Intake Noble Gas channels are non-conservative. This affects Tech Spec 3.3.7 required monitors 0PR31B, 0PR32B, 0PR33B, 0PR34B (Noble Gas channels). ENS notification was made under ENS 46414 under 10 CFR 50.72(b)(3)(v)(D).
"The design basis accidents do not credit automatic actuation of the Control Room Outside Air Intake system to the Emergency mode from a high radiation signal. Therefore, the high radiation signal is not needed to mitigate the consequences of an accident and this event did not result in a safety system functional failure.
"Therefore, ENS notification 46414 is being retracted."
The licensee notified the NRC resident inspector. Notified R3DO (Bloomer) |
| Power Reactor | Event Number: 46415 | Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MIKE DEBOARD
HQ OPS Officer: DONALD NORWOOD | Notification Date: 11/12/2010
Notification Time: 15:44 [ET]
Event Date: 11/12/2010
Event Time: 13:00 [CST]
Last Update Date: 11/12/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
ANN MARIE STONE (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 | 100 | Power Operation | 100 | Power Operation | Event Text INACCURATE INFORMATION PROVIDED IN LICENSE AMENDMENT REQUEST
"At 1300, on November 12, 2010, Exelon Generation Company LLC concluded that inaccurate information contained in the PRA technical bases for a 1987 License Amendment Request (LAR) for Byron and Braidwood Stations would have potentially impacted the acceptability of the LAR by the NRC. The LAR was to extend Allowed Outage Times (AOT) from 72 hours to 7 days for several systems, to include the Component Cooling (CC) and Residual Heat Removal (RH) Systems.
"The original design intent of the CC system was that each unit has two independent CC pumps and a fifth pump (U0) CC pump could be used as an operable spare for any of the unit specific pumps. This is how CC was modeled in the PRA technical justification for the 1987 LAR. However, a piping configuration design flaw that was recently evaluated in that the U0 CC pump could not be considered an operable spare for either unit's B pumps was not correctly modeled in the PRA.
"During the evaluation to assess the potential significance of this CC design flaw on the PRA justification for the 1987 LAR, another potentially significant discrepancy was discovered in that it appears the operational practice to always split CC trains after a design basis LOCA was not modeled correctly in the RH analysis.
"Administrative controls have been put in place to restrict the AOT for the CC pumps and RH trains to the pre-LAR timeframe of 72 hours pending the permanent corrective actions. In addition, administrative controls have been put in place to prohibit the U0 CC pump from being an operable spare for either unit's B trains.
"This event is being reported as an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii).
"The NRC Resident Inspectors have been notified" |
| Power Reactor | Event Number: 46416 | Facility: BYRON
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: LEO WEHNER
HQ OPS Officer: DONALD NORWOOD | Notification Date: 11/12/2010
Notification Time: 16:33 [ET]
Event Date: 11/12/2010
Event Time: 13:00 [CST]
Last Update Date: 11/12/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
ANN MARIE STONE (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 | 100 | Power Operation | 100 | Power Operation | Event Text INACCURATE INFORMATION PROVIDED IN LICENSE AMENDMENT REQUEST
"At 1300, on November 12, 2010, Exelon Generation Company LLC concluded that inaccurate information contained in the PRA technical bases for a 1987 License Amendment Request (LAR) for Byron and Braidwood Stations would have potentially impacted the acceptability of the LAR by the NRC. The LAR was to extend Allowed Outage Times (AOT) from 72 hours to 7 days for several systems, to include the Component Cooling (CC) and Residual Heat Removal (RH) Systems.
"The original design intent of the CC system was that each unit has two independent CC pumps and a fifth pump (U0) CC pump could be used as an operable spare for any of the unit specific pumps. This is how CC was modeled in the PRA technical justification for the 1987 LAR. However, a piping configuration design flaw was recently evaluated in that the U0 CC pump could not be considered an operable spare for either unit's B pumps was not modeled in the PRA.
"During the evaluation to assess the potential significance of this CC design flaw on the PRA justification for the 1987 LAR, another potentially significant discrepancy was discovered in that it appears the operational practice to always split CC trains after a design basis LOCA was not modeled correctly in the RH analysis.
"Administrative controls have been put in place to restrict the AOT for the CC pumps and RH trains to the pre-LAR timeframe of 72 hours pending the permanent corrective actions. In addition, administrative controls have been put in place to prohibit the U0 CC pump from being an operable spare for either unit's B trains.
"This event is being reported as an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii).
"The NRC Resident Inspectors have been notified" |
| Power Reactor | Event Number: 46417 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: MARTIN RAJKOWSKI
HQ OPS Officer: DONALD NORWOOD | Notification Date: 11/12/2010
Notification Time: 20:42 [ET]
Event Date: 11/12/2010
Event Time: 12:10 [CST]
Last Update Date: 11/12/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
ANN MARIE STONE (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 47 | Power Operation | 47 | Power Operation | Event Text UNANALYZED 10CFR50 APPENDIX R SCENARIO
"At approximately 1210 on November 12, 2010, a fire protection assessment identified a potentially unanalyzed condition in the plant's 10CFR50 Appendix R analysis. In the unlikely event of a fire in the main control room or cable spreading room, coincident with a fire induced loss of off site power, in which the control room must be evacuated, Operations personnel would proceed to the Alternate Shutdown System (ASDS) panel to perform required safe shutdown activities. During certain scenarios, High Pressure Coolant Injection (HPCI) may start on a reactor low-low water level signal or on high drywell pressure signal. Since HPCI is not controlled manually from the ASDS, a postulated fire induced short could prevent the HPCI system high reactor water level trip and possibly result in reactor vessel overfill.
"Operation of the Safety Relief Valves in this condition has not been analyzed. For this scenario, the following unlikely sequence of events is required:
1. A fire would have to occur in the main control room or cable spreading room
2. The fire would have to be significant enough to require main control room evacuation
3. Offsite power would have to be lost
4. HPCI would have to initiate
5. A fire induced short preventing HPCI from tripping automatically would have to occur
"Applicable safety systems remain operable, and Operations personnel are trained in procedures to handle complex fire scenarios, including fires in the main control room and cable spreading room. Additionally, the cable spreading room is protected by an automatic Halon fire suppression system.
"As a precaution, a fire watch has been established as a compensatory measure. This event is being reported under 10CFR50.72(b)(3)(ii)(B)."
The licensee notified the NRC Resident Inspector. | |