| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
09/30/2010 - 10/01/2010 ** EVENT NUMBERS ** |
 | Fuel Cycle Facility | Event Number: 45655 | Facility: WESTINGHOUSE ELECTRIC CORPORATION
RX Type: URANIUM FUEL FABRICATION
Comments: LEU CONVERSION (UF6 to UO2)
COMMERCIAL LWR FUEL
Region: 2
City: COLUMBIA State: SC
County: RICHLAND
License #: SNM-1107
Agreement: Y
Docket: 07001151
NRC Notified By: GERALD COUTURE
HQ OPS Officer: JOE O'HARA | Notification Date: 01/25/2010
Notification Time: 18:33 [ET]
Event Date: 01/25/2010
Event Time: 11:30 [EST]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
PART 70 APP A (b)(1) - UNANALYZED CONDITION
| Person (Organization):
MARVIN SYKES (R2DO)
BRIAN SMITH (NMSS)
| Event Text UNANALYZED CONDITION - OVERFLOW OF URANIUM BEARING AMMONIATED WASTEWATER
"It was reported to the EH&S Management that on January 24, 2010 a spill of approximately 200 gallons of uranium bearing ammoniated (5-7%) wastewater overflowed from the 'Q' tanks into the diked area below the tanks. These tanks are the final filtration prior to transfer of this liquid effluent to the outside treatment facility. Operators received a high level alarm and responded by shutting down the process in accordance with the operational procedure, with the overflow occurring for approximately six minutes. This event was the result of a pump failure in the tank discharge line. Notification was made to the on duty Health Physics (HP) staff and the on duty Incident Commander. Health Physics staff responded within minutes and used a Drager counter to determine the ammonia concentrations present. Readings in the immediate area of the dike were as high as 256 ppm ammonia. Readings in adjacent areas of the facility were approximately 150 ppm ammonia. Non-essential personnel were evacuated and essential personnel were instructed to don PPE-respirators with ammonia cartridges.
"Operations cleanup of the area was completed and with normal plant ventilation running the ppm ammonia concentrations were returned to < 25 ppm within approximately two hours. The failed pump has been repaired and returned to service. Based on the quick response of the HP staff, evacuations and appropriate use of PPE, no workers were exposed to significant concentrations and no medical attention was necessary.
"Notification is made based on 10CFR70 Appendix A (b)(1) 'Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 10CFR70.61.' The potential for a loss of containment was recognized and evaluated in one of the Process Hazards Analysis (PHA) which constitutes the Integrated Safety Analysis for this system. The PHA identified several initiating events which could lead to a high level and loss of containment event. The appropriate safeguards were identified, including the procedural responses, the evacuation during such an emergency of the workers in the enclosed chemical area, and the use of appropriate PPE. The consequences of the event were identified as a potential for personnel inhalation and exposure hazard from the uranium bearing ammoniated wastewater. However, the PHA did not specifically indentify that the potential existed for the consequences to exceed the Intermediate Consequence criteria for credible events. In accordance with SNM-1107 License Requirements for the Columbia Plant Intermediate Consequences are those that have the potential for a worker to receive greater than or equal to ERPG-2 chemical exposures. (ERPG-2 value for ammonia is 150 ppm.) Since the Q-Tank contains comingled uranium and chemicals, the Intermediate Consequences of 10CFR70.61 apply. Failure to identify that an Intermediate Consequence event was credible led to that event not being included in the Conversion ISA Summary ISA-03 and not designating Items Relied on For Safety (IROFS) for that accident sequence.
"Corrective Actions: As stated previously, the pump which failed has been repaired and returned to service. Actions taken by the staff to mitigate the event were appropriate and in accordance with approved procedures. Normal ventilation system operation reduced the concentrations to acceptable levels. [These actions are] complete.
"The safeguards identified in the PHA will be evaluated in the ISA and appropriate selections of IROFS will be made based on that evaluation and included in the ISA summary. [These actions are] in progress."
* * * UPDATE AT 1053 ON 9/30/2010 FROM GERARD COUTURE TO ERIC SIMPSON * * *
"Time and Date of Event: September 30, 2010, 0800.
"Previously Westinghouse had reported to the Nuclear Regulatory Commission (NRC) in Licensee Event Report #45655, that the conversion Integrated Safety Analysis (ISA) did not specifically indentify that the potential existed for the consequences of overflows or spills in the conversion wastewater system to exceed the Intermediate Consequence criteria for credible events based on chemical exposure. In accordance with SNM-1107 License Requirements for the Columbia Plant intermediate consequences are those that have the potential for a worker to receive greater than or equal to ERPG-2 chemical exposures. In the required written follow up to that event (LTR-RAC-10-16, February 23, 2010). Westinghouse committed to perform an extent of condition review of other chemical release scenarios and designate necessary Items Relied On For Safety (IROFS). Westinghouse further provided NRC the casual analysis related to that event, (LTR-RAC-10-28, April 6, 2010) which identified the need to quantitatively evaluate chemical release scenarios. Westinghouse has completed the necessary evaluations and determined that additional scenarios exist which have the potential to exceed the performance requirements of 10 CFR 70.61. EH&S Management has now determined that these analyses, the necessary ISA and ISA Summary revisions, and the identification of appropriate IROFS are complete. For these new ISAs the necessary controls, procedures and equipment are in place and the performance requirements are met in a fully compliant manner taking into account the newly identified IROFS.
"The extent of condition review identified additional areas within the facility where chemical releases have the potential to exceed consequence thresholds for facility workers. There is one area where potential consequences could challenge off-site receptor chemical consequence criteria. The areas covered by the extent of condition review impacted by this updated chemical analysis are Conversion, the Scrap Uranium Processing, the Solvent Extraction System, and the facility Wastewater Tanks.
"Notification is made based on 10CFR70 Appendix A (b)(1) 'Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 10CFR70.61.' The potential for a loss of containment was recognized and evaluated in the applicable Process Hazards Analysis (PHA) which constitutes the Integrated Safety Analysis. The PHAs identified initiating events which could lead to loss of containment. The appropriate safeguards were identified, including the procedural responses, the evacuation during such an emergency of the workers in the enclosed chemical area, and the use of appropriate PPE. Failure to identify in the areas mentioned above those chemical release consequences could exceed 10CFR70.61 criteria led to these events not being included in the applicable ISA Summary and not designating Items Relied on For Safety (IROFS) for these accident sequences.
"Corrective Actions: As stated previously, the associated Integrated Safety Analysis, Integrated Safety Analysis Summaries, and designation of IROFS to ensure the performance requirements are satisfied is complete for the extent of condition reviews previously committed to by Westinghouse.
"Full implementation of the revised ISA Summaries, to include all required procedure revisions and appropriate training will be completed by October 30, 2010."
Notified R2DO (Lesser) and NMSS EO (Rubenstone). |
| Power Reactor | Event Number: 46232 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: KENT SLOAN
HQ OPS Officer: CHARLES TEAL | Notification Date: 09/07/2010
Notification Time: 10:43 [ET]
Event Date: 09/07/2010
Event Time: 10:43 [EDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
WILLIAM COOK (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 PLANT IN-PROCESS COMPUTER OOS FOR MAINTENANCE
"The Beaver Valley Power Station (BVPS) Unit 1 In-Plant Computer (IPC) will be taken out of service for approximately 4 weeks (9/7/10 - 9/30/10) to implement a planned modification. The current IPC is being replaced and a computer outage is required to allow for installation of a new IPC. During this time period the Emergency Response Data System (ERDS) data link to the NRC will not be available at BVPS Unit 1. Other computer based systems not directly associated with the IPC (e.g. Safety Parameter Display System (SPDS), meteorological data) will remain in operation. ERDS parameters will be available to be monitored by control board indications and temporary computer system set up prior to the IPC outage. Compensatory actions have been developed to direct one of the Technical Support Center (TSC) Operations communicators to respond to the control room during a BVPS Unit 1 emergency, should it occur, to facilitate data transfer to the NRC while the ERDS is out of service. Work on replacing the IPC and returning ERDS to service will be ongoing continuously until complete.
"This is an 8-hour reportable event per 10 CFR 50.72(b)(3)(xiii) Major Loss of Assessment Capability. The operation of BVPS Unit 1 and Unit 2 plant systems will not be adversely affected due to this planned action. BVPS Unit 2 ERDS will not be affected by these modifications.
"The NRC Resident Inspector has been notified."
* * * UPDATE FROM THOMAS MIGDAL TO HOWIE CROUCH @ 2215 EDT ON 9/30/10 * * *
"The Beaver Valley Power Station Unit 1 In-Plant computer has been returned to service. The Emergency Response Data System data link to the NRC is now available.
"The NRC Resident Inspector has been notified."
Notified R1DO (Gray). |
| General Information or Other | Event Number: 46277 | Rep Org: OK DEQ RAD MANAGEMENT
Licensee: SAINT JOHN MEDICAL CENTER
Region: 4
City: TULSA State: OK
County:
License #: OK-00376-02
Agreement: Y
Docket:
NRC Notified By: MORGEN BUCKNER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 09/24/2010
Notification Time: 12:25 [ET]
Event Date: 01/10/2008
Event Time: [CDT]
Last Update Date: 09/24/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
JACK WHITTEN (R4DO)
ANGELA MCINTOSH (FSME)
| Event Text AGREEMENT STATE REPORT - MEDICAL MISADMINISTRATION
The following report was submitted via e-mail:
"On January 16, 2008, Saint John Medical Center (SJMC) Lic. # OK-00376-02, of Tulsa, OK notified the Oklahoma Department of Environmental Quality (ODEQ) that on January 10, 2008 a patient had been administered a dose of I-131 that differed from the intended dose by greater than 20%. The intended dose was 100 mCi. The administered dose was 25 mCi. The misadministration occurred because the 100 mCi dose provided by Nuclear RX, PC (NRX) Lic. # OK-31035-01MD of Tulsa, OK was divided among three capsules. Two capsules contained 25 mCi, while the third contained 50 mCi I-131. The bottle received by SJMC was opaque and stated that it contained one capsule of 309 mCi because of a software error at NRX. When the dose was administered to the patient, one capsule was dispensed from the bottle, while the other two stuck in the bottom of the bottle. The presumed empty bottle was then repackaged and shipped back to NRX where it was discovered that two capsules remained in the bottle. It was determined that the capsules contained a total of 75 mCi I-131. The capsules were not discovered before shipping because the tech at SJMC errantly surveyed the package before placing the bottle inside. Corrective actions by SJMC involve surveying the bottle prior to returning it to the transport shielding, refresher training on transport requirements in 49 CFR, and determining the transport index (TI) by survey outside of the hot lab. Corrective actions by NRX included contacting the software developer for an update that would ensure that all shipments are accurately labeled. NRX also agreed to write the number of capsules contained in each bottle on the lid of the bottle.
"Due to an oversight, this event was not reported at the time it was received by ODEQ."
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. |
| General Information or Other | Event Number: 46286 | Rep Org: OHIO BUREAU OF RADIATION PROTECTION
Licensee: CLEVELAND CLINIC FOUNDATION
Region: 3
City: CLEVELAND State: OH
County:
License #: OH02110180013
Agreement: Y
Docket:
NRC Notified By: MICHAEL SNEE
HQ OPS Officer: ERIC SIMPSON | Notification Date: 09/28/2010
Notification Time: 15:01 [ET]
Event Date: 09/27/2010
Event Time: [EDT]
Last Update Date: 09/28/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
STEVE ORTH (R3DO)
KEVIN HSUEH (FSME)
| Event Text OHIO AGREEMENT STATE REPORT - 95% MEDICAL UNDERDOSAGE DUE TO EQUIPMENT FAILURE
The following report was received from the State of Ohio via email:
"On September 27, 2010, the Gamma Knife gave a Fatal Error and terminated treatment to a patient. The error appears to be a failed computer disc drive. The safety system of the Gamma Knife functioned as designed, moving the patient out of the treatment machine and closing the Gamma Knife doors. The patient was safely removed from the treatment room. A service representative was immediately contacted and repair of the Gamma Knife is in progress.
"It is intended to give the remaining dose from the plan to the patient once the Gamma Knife is repaired."
The device in question is a Leksell, Model Perfexion Gamma Knife unit [S/N MV010], which contains a 13,824 Ci Co-60 source. The intended dose was 1400 rad. The delivered dose was 71.5 rad. The target organ was the brain. There is no effect on the patient.
Ohio report #: OH100021.
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: 46292 | Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [1] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: KENT MILLS
HQ OPS Officer: STEVE SANDIN | Notification Date: 09/30/2010
Notification Time: 09:39 [ET]
Event Date: 09/30/2010
Event Time: 06:50 [EDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
MEL GRAY (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 | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFSITE NOTIFICATION - INADVERTENT ALARM ACTUATION
"At 0650 [EDT] on September 30, 2010, site personnel were notified by multiple Calvert County residents of a single siren activation (Siren C21) at 0645 [EDT] for approximately 2-3 minutes.
"Following a detailed review of the site's Alert and Notification System, the site identified multiple activations of the siren in addition to the first inadvertent activation of the siren that occurred at 0645 [EDT]. The multiple siren activations were not related to any condition or event at Calvert Cliffs. The siren has experienced multiple losses of power as a result of the effects from heavy rain and wind associated with the passing of Tropical Storm Nicole. Each time the power had restored to the siren, it performed a full cycle test as opposed to the standard silent test.
"County Officials were notified and are taking actions to communicate to Calvert County residents concerning the inadvertent siren actuation.
"Site maintenance locally deactivated the siren at 0827 [EDT] and will further investigate the extent of issue with the siren once the Tropical Storm Nicole has passed. In response to deactivating the siren, Calvert County has implemented the back-up method of route alerting in the event that actual siren activation is required.
"The NRC Resident Inspector has been informed of the activations." |
| Other Nuclear Material | Event Number: 46293 | Rep Org: DEL VALLE GROUP
Licensee: CRMI
Region: 1
City: SAN JUAN State: PR
County:
License #: 52-25430-01
Agreement: N
Docket:
NRC Notified By: CARMELO CALDERON
HQ OPS Officer: ERIC SIMPSON | Notification Date: 09/30/2010
Notification Time: 14:15 [ET]
Event Date: 09/29/2010
Event Time: 09:00 [EDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
| Person (Organization):
MEL GRAY (R1DO)
KEVIN HSUEH (FSME)
ILTAB VIA EMAIL ()
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text LOST/STOLEN MOISTURE DENSITY GUAGE
An employee returning from a job site in Bayamon, Puerto Rico, stopped overnight at his home in San Lorenzo with a licensed moisture density gauge in the back of his truck. On the next day, while traveling to the next job site in Miramar, San Juan, the employee noticed that the moisture density gauge was missing from the truck bed. The driver retraced his steps, but was unable to locate the gauge and was not able to determine whether it had fallen off of the back of the truck or was stolen overnight.
The San Lorenzo Police Department has been notified. The police report number is 2010606705020.
The Puerto Rico Department of Health has been notified. A press release is planned to assist in recovering the moisture density gauge.
The gauge is a Seaman Model C200, S/N L508, which contains a 8 mCi Cs-137 source and a 40 mCi Am/Be source.
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. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
| Power Reactor | Event Number: 46294 | Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PAIGE KEMP
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/30/2010
Notification Time: 15:36 [ET]
Event Date: 09/29/2010
Event Time: 14:15 [EDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
INFORMATION ONLY
| Person (Organization):
MARK LESSER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text DISCOVERY OF MICROTHERM INSULATION IN CONTAINMENT WHICH COULD IMPACT SUMP STRAINER PERFORMANCE
"On September 12, North Anna Power Station Unit 1 entered a scheduled refueling outage. During the outage while performing reviews for a design change, it was determined that some uncontrolled plant drawings indicated that Microtherm insulation may have been installed on some piping locations inside the containment. The Microtherm insulation impacts the new containment sump strainers installed to address NRC GSI-191 requirements. It was subsequently confirmed via walk downs that the Unit 1 containment contained some Microtherm insulation.
"Based on conflicting information on the possible installation of Microtherm inside Unit 2 containment and the identification of Microtherm insulation on Unit 1, a conservative decision was made to shutdown Unit 2 on September 29, and the Unit was cooled down to Mode 5 to allow a detailed inspection to be performed. The areas where the Microtherm insulation was potentially installed could not be inspected with the Unit at power.
"Subsequent inspections have confirmed that the Microtherm insulation was installed in some areas of the Unit 2 containment. These locations include the reactor vessel nozzles and reactor coolant pump casings. Plans are being developed to remove this insulation on both units.
"The overall impact of the Microtherm insulation on the containment sump operation is being evaluated by engineering."
The licensee has notified the NRC Resident Inspector and the Louisa County Administrator. |
| General Information or Other | Event Number: 46296 | Rep Org: ABB, INC.
Licensee: ABB, INC.
Region: 1
City: CORAL SPRINGS State: FL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: CHAD BUCKWALTER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/30/2010
Notification Time: 17:35 [ET]
Event Date: 09/09/2010
Event Time: [EDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
THOMAS FARNHOLTZ (R4DO)
MARK LESSER (R2DO)
PART 21 GROUP ()
| Event Text PART 21 REPORT ON SOLID STATE RELAYS
ABB, Inc. notified the NRC of solid state relays that failed to comply with ABB manufacturing specifications. Subject relays are 27H catalog numbers 411R0175-DP-1E and 411R0175-1E. The failure to comply is in regard to improper installation of a jumper on the main printed circuit board (PCB) of the relay. Manufacturing specifications require the jumper leads to extend far enough beyond the surface plane of the PCB to ensure proper connection. The jumper enables the relay to provide high-speed, instantaneous trip capability. If the jumper were not installed properly, the trip function may be erratic. The subject relays were manufactured between October 1, 2009 and September 9, 2010.
Two safety-related versions of this relay have been purchased by South Texas Project. ABB, Inc. is investigating whether any nuclear plants have purchased the affected relays for use in non safety-related applications.
ABB will be notifying South Texas Project of this defect. |
| Power Reactor | Event Number: 46297 | Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: SCOTT LINDQUIST
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/30/2010
Notification Time: 21:39 [ET]
Event Date: 09/30/2010
Event Time: 16:16 [CDT]
Last Update Date: 09/30/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
THOMAS FARNHOLTZ (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 UNANALYZED CONDITION DUE TO POTENTIAL FLOODING VIA CONDENSATE DRAIN LINES
"USAR Section 2.7.1.2, River Stage and Flow, states flooding protection against the 1,014 foot flood in the auxiliary building is provided by removable flood barriers and sandbagging. When required, these flood barriers are installed in openings leading to safety related equipment on the 1,007 foot and 1,011 foot floor elevations.
"It has been identified that the condensation drains from the switchgear room's air handling units VA-87 and VA-88 (located in the auxiliary building), and the upper electrical penetration room's air handling units VA-85 and VA-86 (located in the auxiliary building), have no isolation valves or check valves to prevent backflow from the drain line's discharge in the turbine building basement. This means that flooding of the turbine building above approximately the 1011 foot elevation (floor level of the switchgear rooms) would result in water back-flowing via the drain lines into the switchgear rooms.
"River level is currently at the 999' 6" elevation and stable. Procedure changes are currently being developed to block the affected drain lines."
River level has never reached the 1011 foot elevation at the facility.
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 46298 | Facility: SOUTH TEXAS
Region: 4 State: TX
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRAD HARDT
HQ OPS Officer: STEVE SANDIN | Notification Date: 10/01/2010
Notification Time: 01:35 [ET]
Event Date: 09/30/2010
Event Time: 19:04 [CDT]
Last Update Date: 10/01/2010 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
THOMAS FARNHOLTZ (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 STANDBY DIESEL GENERATOR AUTOSTART DUE TO LOSS OF A SWITCHYARD BUS DURING MAINTENANCE
"At 1904 [CDT] on 9/30/2010, the South Texas Project (STP) North switchyard bus was lost due to a Transmission & Distribution Service Provider (TDSP) human performance error that occurred while performing maintenance on breaker Y-0530. This resulted in a loss of power to Standby transformer 1 which was supplying power to the Unit 1 B Train Engineered Safety Features (ESF) 4160v bus. The B Train Standby Diesel Generator automatically started due to the Loss of Offsite Power (LOOP) on its associated bus. The Mode II (LOOP) ESF loads sequenced onto the bus. All safety related equipment responded as expected. Action (e) of Technical Specification (TS) 3.8.1.1, 'AC Electrical Power Sources', was momentarily entered due to the loss of two independent offsite circuits while the North bus was de-energized. The North bus was de-energized for approximately 5 minutes. Action (a) of Technical specification (TS) 3.8.1.1, 'AC Electrical Power Sources', was entered due to the loss of one independent offsite circuit. All Technical Specification Limiting Condition of operation have been exited at this time.
"An 8 hour notification is required for this event due to the valid actuation of safety related equipment as described in 10CFR50.72 (b) (3) (iv) (A). [A notification is required for] any event or condition that results in valid actuation of any of the systems listed in paragraph (b) (3) (iv) (B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
"The NRC Resident Inspector has been notified."
Unit 2 briefly entered a technical specification limited condition of operation, while all electrical buses remained energized. | |