Event Notification Report for February 07, 2005

DISCOVERY OF AFTER-THE-FACT EMERGENCY CONDITION (UNUSUAL EVENT)

At 1734 on 2/7/2005, received an Area Radiation Monitor alarm in the reactor building basement airlock area at approximately 75Mr/hr. The valid alarm caused entry into EOP flowcharts for High Radiation in the Secondary Containment. Check of the relay room monitor showed 100 Mr/hr. Investigation showed that a failed open main steam line drain valve combined with placing hydrogen water chemistry in service caused the increasing radiation levels. A downstream steam line drain isolation valve was closed to isolate the steam flow path past the monitor. Hydrogen Water Chemistry injection rate was lowered. The Rad levels returned to about 4 Mr/hr (normal levels) following valve closure (at 1745). The EOPs were exited at 1755. After review of the event it was determined that an unusual event should have been entered at the time of the EOP entry (EAL AU2, Unexpected Increase of Plant Radiation Levels), 1734 and exited at 1745 when area radiation levels returned to normal values. This is an after the fact notification of a missed emergency classification.

The NRC Resident Inspector was notified.

* * * UPDATE ON 2/10/05 AT 1907 HRS. EST BY CROUCH * * *

This event notification reporting criteria has been changed to comply with the guidance provided in NUREG-1022, Rev. 2, "Event Reporting Guidelines 10CFR 50.72 and 50.73". There is specific discussion that states (in part) that if a licensee discovers that a condition existed which met the emergency plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of discovery, an actual declaration of the emergency class is not necessary. An ENS notification of the missed classification is an acceptable alternative.

The Headquarters Operations Officer notified R3DO (Duncan).

PLANT HAD A FREON LEAK IN THE DRYWELL CHILLER ROOM

The plant reported that there was a freon leak in the Drywell Chiller Room located in the Unit 2 Reactor Building. Maintenance was being performed on the refrigeration equipment relief valve when the gas began to escape into the room. There were six men working in the room and two were affected by the freon gas. They were treated on site and released. Oxygen content in the room was 20.5% and LEL for hydrocarbons was alarming at 19%. The drywell chiller room is secured and being cleared of the 1,200 pounds of freon gas. At this time an investigation is being made to determine the cause which may have been due to the relief valve opening or the workers removing the valve. The plant entered the NOUE due to a "toxic gas release". The NOUE will be terminated when the room is habitable.

The NRC Resident Inspector was notified along with State and Local agencies.

* * * UPDATE PROVIDED BY FRANK GORLEY TO JEFF ROTTON AT 0303 EST ON 02/08/05 * * *

Licensee reported that the NOUE was terminated at 0245 EST after the freon had been successfully cleared from the Drywell Chiller Room. Licensee will notify the NRC Resident Inspector.

Notified FEMA (Liggett), DHS (Knox), IRD (Wilson), R2DO (Payne), and NRR EO (Beckner).

RESIDUAL HEAT REMOVAL SYSTEM INOPERABLE DUE TO EMERGENCY DIESEL GENERATOR TRIP DURING TESTING

"This report is being made pursuant to 10CFR50.72(b)(3)(v)(B) 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat;'

"This report is being made due to a trip of the Emergency Diesel Generator during testing that resulted in the RHR loops potentially becoming depressurized. This has the potential to render all RHR Shutdown Cooling unavailable and prevent the removal of decay heat.

"Sequence of events (all times CST):
At 12:00 [02/07/05], Shutdown Cooling was removed from service to prepare for Sequential Load testing of DG #1. This was a planned evolution. At this time decay heat was being removed by the fuel pool cooling system with 2 fuel pool cooling pumps and 2 fuel pool cooling heat exchangers. Time to boil was calculated to be 26 hours.

"At 15:58, the Sequential Load Test commenced on the inoperable DG. The DG came up to speed and sequenced on the initial loads (RHR pumps, a CS pump and a SW pump). Shortly into the sequencing of the DG, the DG tripped due to a blown fuse in the DG control circuit. Sequential loading was not completed. The trip occurred between 13 seconds and 20 seconds of the sequential load. This resulted in the initial loads losing power. Procedurally, the minimum flow valves for the RHR and CS pumps were being remotely opened from the Control Room at the time the DG tripped. This resulted in low-pressure alarms on both RHR systems and one CS system. One fuel pool cooling pump was deenergized, per design, during the sequential load test. Both fuel pool cooling heat exchangers remained in service. With these conditions, the fuel pool cooling lineup does not qualify as an alternate decay heat removal method.

"At 16:04, both RHR loops were declared inoperable due to depressurizing the RHR loops. At 16:02, the tripped fuel pool cooling pump was restored to operation and previous decay heat removal was restored. No unexpected rise in temperature occurred during the time that only 1 fuel pool cooling pump was in operation. This reestablished the fuel pool cooling system as an alternate decay heat removal method.

"At 19:11, the B loop of RHR was returned to a standby lineup and declared operable.

"At this time investigation into why the DG fuse blew is ongoing. All indications are that other equipment performed as designed."

The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM C. BLAIR TO M. RIPLEY 1548 EST 03/08/05 * * *

The following is a correction to the original report received via facsimile (licensee text in quotes):

"Instead of the minimum flow valves for RHR and CS being opened, the suppression pool inboard cooling valve for RHR and the test line recirculation valve for CS were being opened."

The licensee will notify the NRC Resident Inspector. Notified R4 DO (T. Pruett)

* * * RETRACTION FROM COY BLAIR TO MARK ABRAMOVITZ 3/31/2005 AT 14:40 * * *

The following information was provided by the licensee (licensee text in quotes):

"On 2/7/2005 at 1558 CST, Cooper Nuclear Station made an 8 hour 50.72 non-emergency notification to the NRC. The report was made pursuant to 10 CFR 50.72(b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat.' A control power failure during Emergency Diesel Generator #1 (DG) surveillance testing resulted in the loss of the Residual Heat Removal (RHR) pressure maintenance pump. This resulted in the potential de-pressurization and unavailability of all RHR Shutdown Cooling (SDC) and the ability to remove decay heat using RHR. NUREG 1022 Revision 2 defines the safety functions to be considered for Reportability under this section of the rule as being those that are listed in the regulation itself. Thus, the lost safety function being reported was 'remove decay heat'.

"Plant conditions prior to the testing were: Mode 5 (Refueling) with the Reactor Vessel and Drywell heads removed and reactor water level flooded up and Spent Fuel Pool transfer gates removed. Division II RHR was in service providing SDC for decay heat removal. In preparation for the DG testing and in accordance with Technical Specifications, all RHR SDC was removed from service. With RHR SDC out of service, reactor coolant circulation was verified to be by natural circulation with operators monitoring reactor coolant temperatures once per hour. Alternate decay heat removal was provided by the credited lineup of two Fuel Pool Cooling (FPC) pumps and two FPC heat exchangers. FPC receives cooling water from the Reactor Equipment Cooling System (REC), which in turn is cooled by the Service Water System (SW). During the preparation period (approximately 4 hours) for the DG #1 testing, reactor coolant temperature was allowed to slowly go from 85 degrees Fahrenheit to 90 degrees Fahrenheit.

"During load sequencing testing of DG #1, the DG tripped due to a control system failure and de-energized the Division I 4160 Volt (V) critical bus. (Note: The bus was previously de-energized for a short period of time as part of the test.) This caused the pump providing pressure maintenance for the RHR to trip potentially depressurizing the RHR loop (Division II) that had been lined up to provide SDC. A conservative decision was made to declare Division II SDC inoperable during the DG trip recovery.

"If the test had proceeded as planned one RHR pump would have been running in Division I in the test mode (pumping water to the suppression pool). No RHR pumps would have been running in Division II (lined up to allow the Division I test to be conducted). DG #2 remained in normal standby lineup. Division II 4160 V bus was energized supplying power to connected loads. Due to the DG #1 trip the Division I 4160 V bus was deenergized. Shutdown Cooling using RHR could not be placed in service as a result of the test lineup established for DG #1 testing. Reactor coolant circulation was by natural circulation and reactor decay heat removal was by one FPC pump and two FPC heat exchangers. The trip of one FPC pump is expected and verified during this surveillance test. REC was operating with cooling supplied by Division II SW.

"During the period of time after the DG trip and prior to the restoration of electrical power to the Division I 4160 V bus, coolant circulation continued by natural circulation with one FPC pump and two FPC heat exchangers providing decay heat removal. At approximately the time of the DG trip coolant temperature was 90 degrees Fahrenheit. Just after power was restored coolant temperature was 89 degrees Fahrenheit. Operators had adjusted REC temperatures and flows to provide additional cooling to Fuel Pool Cooling. An additional FPC pump was started to provide a two FPC pump and two FPC heat exchanger lineup for reactor decay heat removal. The small variation in coolant temperature demonstrates that the FPC lineup was adequate to provide decay heat removal.

"Engineering performed an evaluation to investigate bulk water temperature response to the event with one FPC pump and two heat exchangers supplying cooling with the fuel pool gates removed. The results show extended periods of time for pool heat-up and are considered bounding. It takes 21 hours for the pool temperature to reach 150 degrees Fahrenheit and 94 hours for the bulk temperature to reach a maximum value of 182 degrees Fahrenheit. Based on this evaluation CNS concludes the maximum bulk temperature would not exceed 182 degrees Fahrenheit.

"As discussed above, RHR SDC was removed from service to support Emergency Diesel Generator surveillance testing. While RHR SDC was out of service, reactor coolant circulation was provided by natural circulation. At the same time, the safety function of decay heat removal was provided by Fuel Pool Cooling. Since the decay heat removal safety function was never lost this is not a reportable event."

The licensee notified the NRC Resident Inspector.

Notified the R4DO (Graves).


* * * UPDATE ON 04/07/05 @ 0725 BY COY BLAIR TO CHAUNCEY GOULD * * *

The following is a change to paragraphs 3 and 4 of the above retraction statement


"During sequential load testing of DGI, the normal expected response after loads are sequenced on, is to have an RHR pump in each division recirculating back to the suppression pool via the suppression pool cooling line. This path is established when the respective RHR pump automatically starts. During load sequencing testing of DG # 1, the DG tripped due to a control system failure and de-energized the Division I 4160 Volt (V) critical bus. (Note: The bus was previously de-energized for a short period of time as part of the test.). Due to the timing of the DG failure, both RHR pumps started and both suppression pool cooling valves were opened. Subsequently the DG tripped and the RHR pumps stopped due to no power available. The suppression pool cooling valves were unable to be closed prior to depressurizing both RHR loops. A conservative decision was made to declare Division II SDC inoperable during the DG trip recovery.

"DG #2 remained in normal standby lineup. Division II 4160 V bus was energized supplying power to connected loads. Reactor coolant circulation was by natural circulation and reactor decay heat removal was by one FPC pump and two FPC heat exchangers. The trip of one FPC pump is expected and verified during this surveillance y test. REC was operating with cooling supplied by Division II SW."


The NRC Resident Inspector will be informed.

Reg 4 RDO(Linda Howell) was notified.

MANUAL REACTOR TRIP DURING STARTUP DUE TO INDICATION OF MISALIGNED ROD

The following information was provided by the licensee via facsimile:

"While withdrawing Control Bank 'A' during the Reactor Startup, Rod B-10 indicated a rapid drop from approximately 42 steps to 17 steps on the CERPI (Computer Enhanced Rod Position Indication) panel. The reactor operator stopped withdrawal of 'A' control bank and the CERPI indication for rod B-10 remained at 17 steps. The remaining CERPIs in 'A' control bank varied from 40 to 45 steps.

"The startup was terminated and the reactor was manually tripped in accordance with AP-1 and 1-E-0 [was] initiated. All systems functioned as required on the trip. Initial investigation by I & C and Engineering found no problems with the CERPI indication. Rod Drop time data from the CERPI program shows all rods in Control Bank 'A' had a drop time of 0.32 to 0.38 seconds with the exception of B-10, which had a drop time of 0.18 seconds.

" An investigation is ongoing as to the cause of rod B-10 misalignment.

"This notification is being made pursuant to 10 CFR 50.72(b)(3)(iv)(A). The NRC resident was notified of this event."

The reactor was subcritical in the Source Range at the initiation of the event. All rods inserted fully during the manual reactor trip. S/G level is being maintained by main feedwater and decay heat removal is via the S/G PORV.

INVALID SPECIFIED SYSTEM ACTUATION

The following information was obtained from the licensee via facsimile (licensee text in quotes):

"This telephone notification is provided in accordance with 10 CFR 50.73(a)(1), to report an invalid actuation reportable under 10 CFR 50.73(a)(2)(iv)(A).

"On February 7, 2005, at 0756 [CST], with Unit 2 in Mode 3 'Hot Shutdown' in preparation for the 2005 refueling outage, the 2A Reactor Protection System (RPS) bus unexpectedly de-energized during performance of LOS-RP-W1, 'Manual Scram Instrumentation.' As a result, an RPS Bus A half scram and Division 1 Primary Containment Isolation System (PCIS) isolations were received, including the isolation of shutdown cooling. All affected containment isolation valves closed as designed.

"An inspection found that the 2A RPS MG set output breaker had opened. RPS Bus A was transferred to its alternate feed, and the associated containment isolations were reset. Shutdown cooling was restored at 0845 hours on February 7, 2005.

"Troubleshooting found that the 2A RPS MG set voltage regulator had failed. The voltage regulator was replaced and successfully tested, and the 2A RPS MG set was returned to service. Laboratory testing of the failed voltage regulator circuit board found that a solder joint had failed on load-dropping resistor R8, which caused the excitation output voltage to decrease sharply. This resulted in a trip of the 2A RPS MG set output breaker on low voltage.

"The apparent cause of the failed solder joint was a manufacturing defect. Corrective actions include visual inspections of RPS MG set voltage regulatory circuit boards in the field and in the storeroom, and the development of enhanced receipt inspection requirements.

"This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an invalid actuation of containment isolation valves in more than one system."

The licensee has notified the NRC Resident Inspector.

AGREEMENT STATE STOLEN SOURCE

The following information was provided by the State of Louisiana Radiation Protection Division (text in quotes):

"On February 7, 2005, Dr. Khalil's Cardiology practice was broke(n) into. A 118 microcuries sealed source of Cs-137 was stolen from the lead pig at the facility. The serial number for the source is 3560380A-16. The model number and manufacturer was not reported at this time. The individual that stole the source was caught at the facility, but apparently he made [a] few trips in and out of the facility. He does not remember what he did with the source. The individual was passed out [intoxicated] at the facility when he was caught. Kilgores, Inc provides Health Physics support for the facility. He [Kilgores, Inc] did a survey of the facility and could not locate the source. This event occurred on February 7, 2005 and was reported to the Louisiana Department of Environmental Quality on March 2, 2005."

LA Report Number: LA050002