Event Notification Report for May 11, 2010

AGREEMENT STATE REPORT - REPORT OF LEAKING SEALED SOURCE

The following information was received by e-mail:

"On Tuesday May 11, 2010, the licensee discovered during a routine leak test that a Ni-63 source in a Shimadzu GC-14A gas chromatograph was leaking, presenting 0.0142 microCi. The source was identified as an NRD model N-100, serial number 616355, with an original activity of 9.98 mCi on 3/23/03 and a current activity of 9.50 mCi.

"A second leak test confirmed that this was not a false positive. Surveys of the instrument and area presented no evidence that contamination had spread.

"The device containing the Ni-63 source was removed from service and taken to the licensee's radioactive waste storage facility."

Ohio Report: OH100006

* * * UPDATE FROM STEPHEN JAMES TO CHARLES TEAL ON 7/8/10 AT 1128 * * *

"The source has been removed and packaged for disposal with the licensee's next LLRW shipment."

LOSS OF TECHNICAL SUPPORT CENTER

"On 05-11-10 at 09:30 hours the Technical Support Center (TSC) ventilation, filtration and climate control system was identified as nonfunctional. The TSC was declared nonfunctional. A plan has been implemented for performance of TSC functions in alternate locations in accordance with site emergency implementation procedures.

"At 13:00 hours on 05-11-10, the system was placed back in service but is still considered non-functional while additional trouble shooting is being performed and final repairs are made."

The licensee notified the NRC Resident Inspector.

OFFSITE NOTIFICATION

"[On] May 10, 2010 at 1710 hours a USEC (United States Enrichment Corporation) Protective Force Officer entered the X-104 Police Headquarters Weapon Cleaning Area to perform cleaning maintenance on his assigned weapon. The officer began to disassemble the weapon [while] pointing the weapon in a safe direction. He pulled the trigger to release the slide and the weapon discharged (this action is required to remove the slide from this type of weapon). The officer in question was the only person in the room at the time of the incident. The officer was not struck by the discharge, however he did sustain minor powder burns to his hand. The area and weapon was immediately secured by Protective Force Management personnel. USEC Fire/EMS were summoned to the incident scene where the officers injury was evaluated at which time the officer refused treatment. Personal statements were collected from all personnel in the immediate area. A critique was conducted. The Protective Force Manager has generated a Long Term Order to provide compensatory actions to prevent recurrence of an accidental discharge of a weapon due to the same or similar circumstances. An internal investigation as well as an independent investigation are currently in progress.

"This event was reportable to the Department of Energy per XP2-RA-RE1001 6.2.1, which requires a formal notification. This formal notification constitutes the need to report this event to the Nuclear Regulatory Commission within four (4) Hours per criteria listed in procedure UE2-RA-RE1030, Appendix 'D', P."

MANUAL REACTOR SCRAM DUE TO LOSS OF CONTROL ROD DRIVE CHARGING WATER HEADER PUMPS

"On May 11, 2010, at approximately 2318 hours, a manual Reactor Protection System (RPS) actuation was initiated as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.1.5 'Control Rod Scram Accumulators.' Control Rod Drive (CRD) charging water header pressure was less than 1520 psig (i.e., no CRD pumps operating) and there were multiple accumulator faults on withdrawn control rods. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the core and the plant is currently stable, in Mode 3 (Hot Shutdown) with reactor pressure at approximately 930 psig. No Emergency Core Cooling Systems were required or utilized to respond to the event and there were no other 10 CFR 50.72 reportable actuations. Reactor coolant level is being maintained in its normal band by the feedwater system and decay heat is being removed by the condenser. The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available if needed.

"The cause of the event initiator, an invalid Division 2 Loss of Coolant Accident (LOCA), i.e., High Drywell Pressure/Low Reactor Vessel Water Level, signal, is currently under investigation. Prior to the manual RPS Actuation, the invalid LOCA signal resulted in invalid actuations of Division 2 equipment and systems including, the Division 2 Emergency Diesel Generator (EDG), (which started but did not load onto the bus), Low Pressure
Coolant Injection B and C subsystems (which started the pumps but did not inject into the vessel), discharge of the Suppression Pool Makeup subsystem B into the suppression pool, startup of the Control Room Emergency Recirculation subsystem B, and isolation of Group 2B Containment isolation valves which included the Nuclear Closed Cooling System Containment Return Isolation valve that was not already closed. The affected equipment is being restored in accordance with plant procedure. The NRC Resident Inspector has been notified."

The licensee experienced an instrumentation rack loss of power which appears to have resulted in the inadvertent Division 2 initiation. The initiator of this event also and led to a loss of power to both control rod drive charging water header pumps resulting in charging water header pressure less than required and related accumulator faults which placed the licensee in a technical specification required shutdown condition. The action statement allows only 20 minutes to restore the condition which was insufficient time for the licensee to correct the condition so a manual scram was initiated from 100% power. The scram was characterized as an uncomplicated scram and all system responses (not related to the initial instrument fault) functioned as required.

UNANALYZED CONDITIONS DISCOVERED DURING NFPA 805 TRANSITION REVIEW

During the licensee's NFPA 805 transition review process, several unanalyzed conditions were discovered but determined to be not reportable at that time. During subsequent review, the licensee determined these conditions did meet reporting requirements.

The following unanalyzed conditions affect all three Browns Ferry units:

"On 5/11/2010, it was determined that in the event of an Appendix-R fire, multiple hot shorts affecting reactor pressure instrument loops, Safety Relief Valves (SRV) overpressure logic or ADS [Automatic Depressurization System] logic could cause 2 to 13 SRVs to spuriously open, for certain fire areas. The current Appendix R safe shutdown analysis only assumes 2 SRVs spuriously open. The issue has significant safety impact due to the potential for one fire scenario to result in spurious opening of multiple SRVs, loss of low pressure inventory makeup, and loss of the condensate system for inventory makeup, which would challenge adequate core cooling during performance of Safe Shutdown Instructions.

"On 8/18/2010, it was determined that in the event of an Appendix-R fire, fire induced circuit damage can potentially result in the inability to manually close the following Motor Operated Valves: Residual Heat Removal Heat Exchanger outlet valves and Emergency Equipment Cooling Water pump cross-tie valves. The failure to be able to manually close these valves could result in the loss of decay heat removal function and loss of credited diesel generators to power required safe shutdown equipment. These issues have significant safety impact since the capability to manually close these valves is necessary to ensure adequate core cooling during performance of BFN Safe Shutdown Instructions.

"On 9/30/2010, it was determined that in the event of an Appendix-R fire, fire induced multiple hot shorts could cause both Inboard and Outboard RHR test return valves, and Drywell Spray and Suppression Pool Spray valves to spuriously open due to damage to the valve control circuit cables. This could result in draining of the Pressure Suppression Chamber Head Tank and the affected low pressure Emergency Core Cooling System loop piping (RHR or CS). Consequently, the discharge pipe in the credited Residual Heat Removal (RHR) loop may not be filled and vented when the Safe Shutdown Instructions (SSIs) call for the RHR pump to be started. The resulting water hammer could result in piping system damage resulting in loss of core cooling and decay heat removal functions and loss of suppression pool inventory. Additionally, single spurious actuation of Core Spray (CS) test return valves due to fire damage to their control circuits could have the same results. These issues have significant safety impact since they would challenge the ability to provide adequate core cooling during performance of Safe Shutdown Instructions.

On 8/22/2011, two unanalyzed conditions were discovered:

First, "it was determined that, in the event of an Appendix-R fire in certain areas, fault propagation due to loss of the breaker control circuit in conjunction with power cable damage could result in de-energization of the associated 4kV Shutdown Board. This potential exists since some 4kV Shutdown Board load breakers are not equipped with separate fuses for trip circuits extending beyond the board. This condition could result in a loss power to credited safe shutdown equipment that would challenge the ability to provide adequate core cooling during performance of BFN Safe Shut down Instructions.

Second, "it was determined that in the event of an Appendix-R fire in certain areas, Multiple Spurious Operations (MSO) could result in the Main Steam Isolation Valves failing to close, or to re-open. This potentially results in a challenge to control inventory loss during performance of Safe Shut down Instructions.

The following unanalyzed condition only affects Unit 2:

"On 8/18/2010, it was determined that in the event of an Appendix-R fire, fire induced circuit damage can potentially result in the inability to manually close certain Main Steam Drain Line Motor Operated Valves. The current Appendix R safe shutdown analysis credits manual closure of these valves. Failure to close these valves results in loss of suppression pool inventory which could challenge adequate core cooling during performance of BFN Safe Shutdown Instructions."

Compensatory actions in the form of fire watches to mitigate all these conditions are in place in accordance with the BFNP Fire Protection Report. The licensee will make the required 60-day written reports in accordance with 10CFR50.73(a)(2)(ii)(B). These events were entered into the licensee's Corrective Action Program.

The licensee has notified the NRC Resident Inspector.