| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
02/06/2013 - 02/07/2013 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 48701 | Facility: INDIAN POINT
Region: 1 State: NY
Unit: [ ] [3] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: JOHN DIGNAM
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 01/25/2013
Notification Time: 12:38 [ET]
Event Date: 01/24/2013
Event Time: 23:38 [EST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
| Person (Organization):
MARC FERDAS (R1DO)
FSME RESOURCE ()
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text RADIAC CALIBRATION SOURCE WINDOW INTERLOCK FAILURE
"At 23:38 hours on January 24, 2013, while performing daily source checks on portable radiation meters, the Shepherd Model 78-2 Calibrator (Serial # 9047) failed a pre-operational source interlock check. The 130 mCi /130 Ci Cs-137 sources were able to be raised while the shield door was not fully shut.
"This is reportable under 10 CFR 70.50(b)(2) since the interlock is required by license to prevent the accidental raising of the source with the door open. There is no equipment considered to be redundant to the interlock. The cause of the failure has not been determined at this time.
"No personnel were exposed to the sources. Sources were returned to the shielded position, the calibrator was locked and removed from service. A survey of the calibrator after securing the sources showed dose rates within expected ranges. The affected calibrator is not part of any installed plant equipment and has no impact on plant operation."
The licensee notified the NRC Resident Inspector.
* * * UPDATE FROM STEVE PRUSSMAN TO CHARLES TEAL ON 2/6/13 AT 1105 EST * * *
The licensee has changed the reporting criteria to 10 CFR 30.50(b)(2). The NRC Resident Inspector has been informed.
Notified R1DO (Powell). |
| Agreement State | Event Number: 48704 | Rep Org: NC DIV OF RADIATION PROTECTION
Licensee: FLOWSERVE CORPORATION
Region: 1
City: RALEIGH State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAMES ALBRIGHT
HQ OPS Officer: BILL HUFFMAN | Notification Date: 01/29/2013
Notification Time: 08:16 [ET]
Event Date: 12/28/2012
Event Time: 15:15 [EST]
Last Update Date: 01/29/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
RONALD BELLAMY (R1DO)
FSME_EVENTS_RESOURCE (E-MA)
| Event Text AGREEMENT STATE - STUCK SOURCE ON A RADIOGRAPHY CAMERA
The following is a summary of a report received via email by the North Carolina Division of Radiation Protection Radioactive Material Branch concerning a source that became stuck while using a radiography camera at a Flowserve facility in Raleigh, NC.
On December 28, 2012 at 3:15 pm during radiographic operations with a Co-60 projector model number 741AE s/n A218 containing a Co-60 source, a stuck source equipment incident occurred when the source failed to retract into the shielded position at the permanent radiographic facility located at Flowserve Flow Control Division in Raleigh, NC.
The licensee implemented its source retrieval procedures and successfully retrieved the source back into the projector without exposure to the staff involved.
Inspection of the guide tube setup identified that the connector that connected the flexible guide tubes was not fully engaged.
All radiographers and assistants were briefed on the stuck source equipment incident and all conditions found. The importance of daily checks was stressed along with the need to perform an inspection of your guide tube setup when making any type of change. All radiographers and assistants were involved in the inspection of the two flexible guide tubes and comparison to our other guide tubes and nothing was identified as a potential problem.
The incident was monitored by the RSO at each stage from beginning to end to ensure that all possible safety and security concerns were adequately addressed.
It is the determination of the RSO that at no time was the safety of the employees, radiographers or the public at risk as a result of this incident. |
| Agreement State | Event Number: 48707 | Rep Org: PA BUREAU OF RADIATION PROTECTION
Licensee: UNIVERSAL WELL SERVICES
Region: 1
City: State: PA
County: LYCOMING
License #: PA-1446
Agreement: Y
Docket:
NRC Notified By: JOSEPH M. MELNIC
HQ OPS Officer: DONG HWA PARK | Notification Date: 01/30/2013
Notification Time: 08:17 [ET]
Event Date: 01/14/2013
Event Time: [EST]
Last Update Date: 01/30/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
RONALD BELLAMY (R1DO)
FSME EVENTS RESOURCE (EMAI)
| Event Text AGREEMENT STATE REPORT - SHUTTER FAILURE
The following was received from the Commonwealth of Pennsylvania via facsimile:
"On January 15, 2013, the licensee sent notification via email to the Central Office [PA Bureau of Radiation Protection] about an event that took place on January 14, 2013. This email was received by Central Office [PA Bureau of Radiation Protection] on January 16, 2013. It is reportable within 24 hours under 10CFR30.50(b)(2).
"During use of the gauge, the density readings were not as anticipated. The electronic technicians replaced the detector with no change in output. Based on the output readings, the suspicion is that the shutter is closed and cannot be opened. The gauge has been taken out of service and stored securely in their Meadville Warehouse.
"The device is identified as:
Manufacturer: Berthold Technologies USA, LLC
Model: LB 8010
Serial #: 10185
Source Serial #: 1160/10
Isotope: Cs-137
Activity: 20 mCi
"The manufacturer is being contacted to investigate and perform repairs if needed. A reactive inspection will be performed by the Department [PA Bureau of Radiation Protection]."
Event Report No: PA 130002 |
| Power Reactor | Event Number: 48720 | Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: AMY CORDENER
HQ OPS Officer: CHARLES TEAL | Notification Date: 02/06/2013
Notification Time: 11:21 [ET]
Event Date: 02/06/2013
Event Time: 09:00 [EST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
RAY POWELL (R1DO)
|
| 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 PLANNED MAINTENANCE ON PLANT PROCESS COMPUTER
"This report is being made in accordance with 10CFR50.72(b)(3)(xiii). Calvert Cliffs Nuclear Power Plant will perform planned maintenance on the Unit 2 Plant Process Computer (PPC) and associated network infrastructure on February 06, 2013. The maintenance will remove the Unit 2 PPC for a planned duration of 12 hours and will render the Unit 2 SPDS out of service for this timeframe. Once the maintenance starts, the effected equipment and functions, including SPDS, can be returned to service within five minutes, if required.
"Should an emergency be declared during this period, the control room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment at all times. Control room emergency response personnel will use emergency response procedures to disseminate plant parameter data points to the affected Emergency Response Facilities until the U-2 PPC is restored. MIDAS (Meteorological Data) and Unit 2 ERDS [Emergency Response Data System] data transmission will remain functional during the maintenance window. All work associated with this plant data network software installation will be performed in an expeditious manner consistent with the goal of minimizing unavailability of the systems listed above. Back-out criteria has been identified as part of the work package. A test of all systems will be performed at the completion of the upgrade."
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 1713 EST ON 2/6/2013 FROM AMY CORDNER TO MARK ABRAMOVITZ * * *
The plant process computer was returned to service at 1700 EST.
The licensee will notify the NRC Resident Inspector.
Notified the R1DO (Powell). |
| Power Reactor | Event Number: 48721 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: JOHN D SELLERS
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 02/06/2013
Notification Time: 11:39 [ET]
Event Date: 12/08/2012
Event Time: 01:16 [EST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| 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 60-DAY OPTIONAL TELEPHONE NOTIFICATION OF INVALID SYSTEM ACTUATION
"On December 8, 2012, at 0116 EST, with Unit 1 operating near 100 percent rated thermal power (RTP), annunciator 601-135 'Torus Level High' was received in the main control room and immediately cleared. Torus water level was reading between 148.1 inches and 148.8 inches on torus water level sensors but torus water level indications from the High Pressure Coolant Injection (HPCI) sensors were observed to be fluctuating between 149 inches and 150 inches immediately after the event. Technical Specification LCO 3.6.2.2 requires torus water level to remain greater than or equal to 146 inches and less than or equal to 150 inches. The Technical Specification allowable limit is 154 inches for the HPCI torus water level sensors and the actual set point for the HPCI suction swap is 152 inches. When the HPCI torus water level instruments' reference legs are not completely filled, fluctuations in indicated level occur. This was determined to be the cause for this invalid actuation. Enough fluctuation occurred such that the HPCI suction swap instruments completed the actuation logic causing the HPCI suction swap design function to occur. This resulted in HPCI automatically realigning to the torus. It had been manually aligned to the condensate storage tank (CST), which is its primary source. HPCI DID NOT realign to mitigate an event or condition nor did it realign for actuation and injection to send water to the reactor vessel. Should the automatic suction swap fail to occur, the Technical Specifications require that the suction be manually realigned to the torus in accordance with plant procedures and training. The operability of HPCI was maintained throughout the event and no HPCI system actuation signal (low reactor water level or high drywell pressure) was received during the event.
"At approximately 0337 EST, Instrumentation and Controls technicians began backfilling the reference legs for sensors 1E41N062B, 1T48N010A, and 1T48N021A. This task was completed at approximately 0503 EST at which time the normal torus water level and HPCI torus water level instrument indications were in agreement. The HPCI suction source was then realigned to the CST in accordance with plant procedures at approximately 0507 EST. This report is being made as an actuation of a system named in 50.73(a)(2)(iv)(B)(4) and is being reported in accordance with 50.73(a)(2)(iv)(A) and under the provision of 50.73(a)(1) based on an NRC interpretation that this actuation within the HPCI system was considered an ECCS actuation. Since the actual torus water level never actually reached 152 inches, the automatic realignment was determined to be an invalid actuation. SNC [Southern Nuclear Company] takes exception to the NRC interpretation on reportability of this condition, but are making this 60 day phone LER notification on a voluntary basis while this issue is being pursued with the NRC. The initiating signal was invalid and this report is being made under the auspices of 10 CFR 50.73(a)(1) for invalid specified system actuations."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 48722 | Facility: POINT BEACH
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: DENNY SMITH
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 02/06/2013
Notification Time: 13:27 [ET]
Event Date: 02/06/2013
Event Time: 11:43 [CST]
Last Update Date: 02/06/2013 | Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
MICHAEL KUNOWSKI (R3DO)
ERIC LEEDS (NRR)
CINDY PEDERSON (R3)
JANE MARSHALL (IRD)
MICHELE EVANS (NRR)
|
| 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 UNUSUAL EVENT - LOSS OF OFFSITE POWER TO ESSENTIAL BUSES FOR GREATER THAN 15 MINUTES
"At 1132 CST on 2/6/13, Point Beach Unit 1 experienced a loss of all offsite power due to the loss of 1X-03 High Voltage Station Aux Transformer. The high side circuit switcher opened. However, the 1X-03 transformer did not lock out resulting in G-01 and G-03 Emergency Diesels energizing the 1A-05 and 1A-06 4160 VAC busses (safety related)."
The cause of the loss of offsite power is being investigated. In order to exit the event, the plant requires either 345 kV power to be restored or cross-tie plant power between the units. If cross-tied, Technical Specifications require the gas turbine to be placed on the cross-tie within 24 hours. The plant remained at 100% power and there was no effect on Unit 2.
The licensee notified the NRC Resident Inspector.
Notified the DHS SWO, FEMA, DHS NICC, and the Nuclear SSA (via e-mail).
* * * UPDATE AT 1500 ON 2/6/2013 FROM JAMIE WEIGANDT TO MARK ABRAMOVITZ * * *
"There were specified system actuations with the original event. Emergency diesel generators [EDG] G-01, 2, 3, and 4 started on the undervoltage conditions with G-01 and G-03 loading onto their respective buses. 1P-53, motor driven aux feed pump started as designed on the EDG breaker closures. Offsite power has been restored to safeguards buses and the EDGs have been removed from the buses. Troubleshooting continues on the initial fault. All other systems functioned as designed. The unusual event has been terminated as of 1340 CST on 2/6/13."
The safeguards buses are being powered from Unit 2 through the electrical cross-tie.
The licensee notified the NRC Resident Inspector.
Notified the R3DO (Kunowsky), NRR EO (Evans), IRD (Marshall), DHS SWO, FEMA, DHS NICC, and the Nuclear SSA (via e-mail). |
| Power Reactor | Event Number: 48723 | Facility: WATTS BAR
Region: 2 State: TN
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRIAN McINLNAY
HQ OPS Officer: BILL HUFFMAN | Notification Date: 02/06/2013
Notification Time: 16:00 [ET]
Event Date: 07/28/2009
Event Time: [EST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| 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 THAT COULD HAVE RESULTED IN AN INCREASED MAXIMUM FLOOD LEVEL
"On July 28, 2009, TVA identified, in the Corrective Action Program, the potential to overtop and fail earthen embankments at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams. This condition could have resulted in an increase in the probable maximum flood (PMF) level at Watts Bar Unit 1. TVA initiated immediate actions to address the condition by conducting additional analyses and the development of contingent actions. Additional actions were developed including the installation of modular flood barriers [which were] completed in December 2009. The barriers increase the effective height of the affected embankments preventing their overtopping and failure. The increase in PMF could have affected plant equipment including the emergency diesel generator system, the essential raw cooling water system, the thermal barrier booster pumps and the control room chillers.
"Additional details regarding the modular flood barriers and the results of TVA's subsequent hydrologic analyses for Watts Bar were discussed in public meetings between TVA and the NRC staff on July 7, 2010 and May 31, 2012, and provided in TVA letters to the NRC dated July 19, 2012, October 30, 2012 and January 18, 2013.
"This report addresses a condition as described in 10 CFR 50.72(b)(3)(ii)(B). Affected safety related equipment is currently operable.
"The NRC Resident Inspector has been notified of this condition."
See related event notifications from Browns Ferry (EN #48724) and Sequoyah (EN #48725). |
| Power Reactor | Event Number: 48724 | Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [2] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: RODNEY NACOSTE
HQ OPS Officer: BILL HUFFMAN | Notification Date: 02/06/2013
Notification Time: 16:31 [ET]
Event Date: 07/28/2009
Event Time: [CST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| 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 | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text UNANALYZED CONDITION THAT COULD HAVE RESULTED IN AN INCREASED MAXIMUM FLOOD LEVEL
"On July 28, 2009, TVA identified, in the Corrective Action Program, the potential to overtop and fail earthen embankments at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams. Prior Browns Ferry Nuclear (BFN) analysis did not consider the potential to overtop and fail the earthen embankments at Cherokee Dam. This condition could have resulted in an increase in the probable maximum flood (PMF) level at Browns Ferry Nuclear Plant Units 1, 2 and 3. TVA initiated immediate actions to address the condition by conducting additional analyses and developing contingent actions. Additional actions were developed including the installation of modular flood barriers [which were] completed in December 2009. The barriers increase the effective height of the affected embankments preventing their overtopping and failure.
"Additional details regarding the modular flood barriers and the results of TVA's subsequent hydrologic analyses were discussed in a public meeting between TVA and the NRC staff on July 7, 2010.
"This report addresses a condition as described in 10 CFR 50.72(b)(3)(ii)(B). Subsequent analyses, completed in November 2012, determined that there are no past operability concerns with this condition at BFN.
"The NRC Resident inspector has been notified of this condition."
See related event notifications from Watts Bar (EN #48723) and Sequoyah (EN #48725). |
| Power Reactor | Event Number: 48725 | Facility: SEQUOYAH
Region: 2 State: TN
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: HAROLD HOWLE
HQ OPS Officer: BILL HUFFMAN | Notification Date: 02/06/2013
Notification Time: 17:10 [ET]
Event Date: 07/28/2009
Event Time: [EST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| 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 UNANALYZED CONDITION THAT COULD HAVE RESULTED IN AN INCREASED MAXIMUM FLOOD LEVEL
"On July 28, 2009, TVA identified, in the Corrective Action Program, the potential to overtop and fail earthen embankments at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams. This condition could have resulted in an increase in the probable maximum flood (PMF) level at Sequoyah Nuclear (SQN) Units 1 & 2. TVA initiated immediate actions to address the condition by conducting additional analyses and developing contingent actions. Additional actions were developed including the installation of modular flood barriers [which were] completed in December 2009. The barriers increase the effective height of the affected embankments preventing their overtopping and failure. The increase in PMF could have affected plant equipment including the emergency diesel generator system and the essential raw cooling water system.
"Additional details regarding the modular flood barriers and the results of TVA's subsequent hydrologic analyses for SQN were discussed in public meetings between TVA and the NRC staff on July 7, 2010 and May 31, 2012, and provided in TVA letters to the NRC dated August 10, 2012, October 30, 2012, and January 18, 2013.
"This report addresses a condition as described in 10 CFR 50.72 (b)(3)(ii)(B). Affected safety-related equipment is currently operable.
"The NRC Resident Inspector has been notified of this condition."
See related event notifications from Watts Bar (EN #48723) and Browns Ferry (EN #48724). |
| Power Reactor | Event Number: 48726 | Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DAN STERMER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 02/06/2013
Notification Time: 21:23 [ET]
Event Date: 02/06/2013
Event Time: 15:24 [PST]
Last Update Date: 02/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
DON ALLEN (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 | | 2 | N | N | 0 | Refueling | 0 | Refueling | Event Text UNANALYZED CONDITION FOR CONTROL ROOM VENTILATION
"On February 06, 2012, at 1524 PST, engineers reviewing dose analyses for non-LOCA, non-fuel handling accident analyses identified deficiencies in the analyses. The analyses of concern include a locked reactor coolant pump rotor, a control rod ejection accident, main steam line break, and steam generator tube rupture. These deficiencies brought into question whether the 30 day control room operator dose received following one of these accidents would meet the station licensing basis limits of up to 5 rem whole body or its equivalent.
"In response to this concern, plant operators placed the control room ventilation system in its safeguards alignment, thereby ensuring the events would continue to be bounded by the analysis of the large break loss of coolant accident.
"The NRC Resident Inspector has been notified." | |