Event Notification Report for December 22, 2005

SAFETY EQUIPMENT FAILURE

"While performing semi-annual criticality accident alarm system (CAAS) testing, one unit in a pair of detectors failed to initiate the site wide alarm. Spare unit was immediately installed and all systems successfully tested. Production facility covered by this CAAS [was] in a shutdown status at the time of testing. Subsequent troubleshooting indicated a faulty electronic relay contact in the failed unit.

"Testing of CAAS is conducted on a semi-annual basis in accordance with procedure, NFS-HS-A-80, Sections 5.5 & 5.6. [The] Detector pair on 2nd floor of Oxide Conversion Facility did not activate site-wide alarm as expected. Alarm indication did occur as expected at the local read-out panel, and at the central alarm panel located in an adjacent building occupied by security guards. Investigation revealed failed relay contact in Eberline Instruments Model RMS-3 read-out meter. [The] Unit [was] replaced, and [a] subsequent test [was] satisfactory.

"No actual safety consequences occurred as a result of this event; however, there was a risk of potential health and safety consequence to the occupational workforce, involving significant radiation exposure from accidental criticality event with no warning to initiate prompt site-wide evacuation."

The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM MICHAEL TESTER TO HUFFMAN AT 1626 EST ON 2/06/06 * * *

Following evaluation of this event by the licensee's Part 21 review committee, Nuclear Fuel Services has reached the conclusion that this event was the result of a design defect in the relay used in the RMS-3 read-out meter. This event is being updated to reflect the Part 21 reportability conclusion. The Eberline RMS-3 read-out meter is manufactured by Thermo Electron Corporation. Nuclear Fuel Services has been in contact with Eberline during its investigation and Eberline is aware of the conclusions. The defective relay is manufactured by Potter and Brumsfield.

Immediate corrective actions included replacement of the defective equipment, and re-testing to ensure operability; long term corrective action includes design and installation of PLC based surveillance equipment to continuously monitor the function of system components by NFS Engineering Department in approximately 3 - 6 months.

The licensee notified the NRC Resident Inspector. The R2DO (Bernhard), NMSS EO ( Janosko) and Part 21 coordinator (Markley) have been notified.

* * *UPDATE BY HUFFMAN ON 2/7/06 * * *

This event has been decontrolled to make it publicly available and permit information about this problem to be shared with all affected parties. NRC management has determined that the report does not contain information about sensitive operations at the NFS site.

R2DO(Bernhard) and NMSS (Morell) notified.

FITNESS FOR DUTY

A licensed supervisor had a confirmed positive for alcohol during a "for cause" fitness-for-duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details.

The licensee notified the NRC Resident Inspector.

AGREEMENT STATE REPORT OF TRAFFIC ACCIDENT INVOLVING RADIOACTIVE MATERIAL

A truck carrying 600 millicuries of Technetium-99m (150 ml in two containers) was involved in a traffic accident in Lawrence, Kansas. The containers were undamaged by the accident and a survey of the containers and truck determined that there was no leakage or radioactive contamination. The Technetium has been secured and taken into possession by representatives of Cardinal Health.

Kansas Report # 7831227

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).

AFW START DURING MAIN FEEDWATER PUMP TRIP

"While performing a plant shutdown due to a packing leak on #24 Feedwater Regulating Valve on 12/22/05 at 0208 at about 67% power, #22 MBFP tripped while swapping lube oil coolers. #21 and #23 motor driven auxiliary feedwater pumps auto started. All systems responded properly. Entered appropriate abnormal operating procedure (2-AOP-FW-1, Loss of Main Feedwater). Reduced power to within the capacity of one MBFP and established conditions to shutdown and realign #21 and #23 ABFPs to Auto. On 12/22/05 at 0218, #21 and #23 ABFPs were shutdown and aligned for automatic operation."

The licensee notified the NRC Resident Inspector.

HPCI DECLARED INOPERABLE

"The HPCI flow control current loop was found degraded in such a manner that HPCI would not perform its intended function. HPCI was declared inoperable per TS 3.5.E.2 (14 day LCO). Repair efforts have been initiated."

The licensee notified the NRC Resident Inspector.

* * * RETRACTION FROM M. RAMSEY TO M. RIPLEY AT 1601 ET ON 02/16/06 * * *

"On 12/22/05, the HPCI System flow indicator 'spiked' a number of times with the system in its normal standby line-up. Control Room Operators declared the HPCI system inoperable and entered a 14 day LCO per TS 3.5.E.2, to troubleshoot and repair the system as appropriate.

"The safety function of HPCI system is to provide and maintain an adequate coolant inventory inside the reactor vessel to prevent fuel clad conditions from exceeding 10CFR50.46 criteria as a result of postulated small breaks in the Nuclear System process barrier. To ensure that this safety function is met, the HPCI System must be able to deliver a minimum actual flow rate of 3570 gallons per minute (gpm) to the reactor vessel. The actual flow rate was reduced below the Technical Specification value of 4250 gpm to account for potential flow instrument string inaccuracies. The delivery of 3570 gpm of actual flow demonstrates that sufficient margin to safety function is maintained during accident conditions. In automatic operation, the HPCI flow controller uses the flow signal to maintain a flow rate of 4250 gpm and is designed to automatically adjust developed flow rate in response to changes in flow signals.

"System operation is tested quarterly by monitoring flow to ensure that the HPCI system can perform its safety functions. The Technical Specification surveillance requirement states that the HPCI system shall deliver at least 4250 gpm at normal reactor operating pressure when recirculating to the Condensate Storage Tank.

"The event investigation revealed that the cause of the flow spikes was a malfunctioning electrical component in the power supply module within the flow controller. Power to the flow transmitter is supplied by this module. The power supply would malfunction for a few seconds and then would return to normal operation, thereby resulting in step changes in the HPCI flow signal. It is conservative to assume the actual HPCI flow rate would have been reduced by the full amount of the flow rate, for the entire duration of the flow signal. Additional conservatism also exists because the HPCI system flow is unlikely to have been able to respond to the full value of a short duration flow spike.

"When calculating the average HPCI flow rate, the impact of the flow spikes decreases as a longer period is evaluated because the duration of all of the flow spikes added together is very short and comparably longer periods of stable performance exist between the individual spikes. Therefore, it is conservative to evaluate the shortest period of time that HPCI is required to perform its safety functions that is stated in the Design Bases Document as 1000 seconds. With HPCI operable, the worst case period occurred just prior to it being declared 'Inoperable'. HPCI was calculated to have developed an actual flow rate of greater than 4191 gpm during the worst case 1,000 second period.

"The calculation for 'HPCI Flow Control & Indication Loop Accuracy' provides total loop accuracy of plus or minus 216.2 gpm at 4250 gpm for Design Bases Accident conditions. A calculated flow rate of 4191 gpm, minus the flow instrument loop uncertainty of 216 gpm yields a worst case actual flow rate of 3975 gpm. A flow rate of 3975 gpm is greater than the 3570 gpm assumed in event calculations.

"The subject flow controller power supply module was removed and bench tested to demonstrate that sufficient operating margin existed to provide reasonable assurance that the power supply would not have failed if called upon to mitigate a design bases accident with sufficient margin. The testing simulated HPCI operation at 4250 gpm for more than 24 hours. No flow spikes were recorded during this period. Based upon these test results and the analysis provided above, if the subject power supply component had remained in service and HPCI operation was required, the system would have performed its required safety function for a duration exceeding any analyzed event.

"ENS Event Number 42219, completed on 12/22/05, is being retracted."

The licensee notified the NRC Resident Inspector. Notified R1 DO (J. Trapp)

VALID ACTUATION SIGNAL DUE TO HIGH STEAM GENERATOR WATER LEVEL

"While performing a plant shutdown due to a packing leak on #24 Feedwater Regulating Valve [see event #42217] on December 22, 2005 at 0550 hours, Indian Point Unit 2 received a 22 Steam Generator Water High-High Level signal at 73% narrow range level. This resulted in tripping the standby 21 Main Boiler Feed Pump which then resulted in a start signal being sent to both motor driven Auxiliary Feedwater Pumps. Both of the motor driven Auxiliary Feedwater Pumps were already operating and feeding the steam generators when this start signal was received. Operators restored 22 Steam Generator Level to normal. All systems responded properly."

The licensee will notify the NRC Resident Inspector.