Event Notification Report for May 06, 2009

AGREEMENT STATE REPORT- MISSING SEALED SOURCES

"UCLA was unable to locate 2 sealed sources which were listed in their inventory database. The first source was 50 micro Ci Co 60 as of 5/1/2000. Current activity is 15.28 micro Ci which exceeds the 1 microCi 24 hour reporting requirement. The second source was listed in their database as 1.83 micro Ci of Cm 244 as of 10/12/1988.

"UCLA believes the original activity for the Cm 244 source to be 1.83 nano Ci, and was listed incorrectly in their database. They contacted the Manufacturer, Isotope Products Laboratory, who informed them that they did not manufacture micro Curie quantities, but only nano Curie quantities of Cm 244. The current activity of the Cm 244 source (decayed to date of notification) is either 832 nano Curies or 0.832 nano Curies. The former exceeds the 1 nano Ci 24 hour reportable quantity requirements.

"UCLA believes the Cm 244 Source was installed in an instrument, but they were unable to locate the instrument at the time of notification. UCLA is continuing to search for the missing sources."

The UCLA RSO (Radiation Safety Officer) believes the sources are still on the campus.

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.

MANUAL REACTOR TRIP DUE TO FEED REGULATING VALVE FAILURE

"On 5/6/09 at 2256 Sequoyah Unit-1 was manually tripped from 100% reactor power. The manual trip was in response to the failure of Loop 1 Feed Water Regulating Valve (FRV). Manual control was attempted to control level in Loop 1 Steam Generator however Loop 1 FRV failed to respond. A manual reactor trip was initiated as a result of this failure. In addition, Auxiliary Feedwater (AFW) initiated as required due to a Feedwater Isolation signal. The Loop 1 FRV did not isolate from the Feedwater Isolation signal, however the Loop 1 Feedwater Isolation Valve closed as designed.

"The Plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 [degrees] F and 2235 psig, with Auxiliary Feedwater supplying the steam generators and Steam Dumps to Main Condenser removing decay heat.

"Maintenance activities have been initiated to repair the Loop 1 FRV."

All rods inserted on the trip. No safety or relief valves lifted as a result of the transient. The plant is in its normal shutdown electrical lineup. No grid instabilities exist and there was no effect on Unit 2.

The licensee has notified the NRC Resident Inspector.

AUTOMATIC ACTUATION OF RPS, ECCS AND EDG GENERATED BY AN INVALID ACTUATION SIGNAL

"On May 6, 2009, Beaver Valley Power Station Unit 1 was in Mode 5 during a refueling outage. At 1352, a Low Steam Line Pressure Reactor Trip and Safety Injection (SI) signal was unexpectedly received on Train A. The Train A Emergency Diesel Generator started upon the SI signal, as designed, but did not load since there was no actual emergency bus low voltage condition. The Train A Charging pump (Emergency Core Cooling Pump) was in service and operating before the generation of the SI signal, and continued to operate following the SI signal. Other plant components not isolated for plant outage conditions properly actuated in response to the generated SI signal.

"A SI Block must be inserted whenever the Solid State Protection System (SSPS) is operating in Mode 5 to prevent a SI signal from being generated on low steam line pressure since steam line pressure will be below the low steam line pressure setpoint. Since the plant was in Mode 5 (temperature less than 200F), there can be no steam in the steam lines. Thus, the receipt of a Low Steam Line Pressure Safety Injection signal while the plant was in Mode 5 was invalid.

"It was determined that the SI Block was removed due to a degraded Steam Line Pressure SI Block/Reset switch on the control room benchboard. Further troubleshooting detected a similar erratic type degradation on the Train A Pressurizer SI Block switch. Both switches were replaced.

"This event is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A) since it involved an actuation of the Reactor Protection System (RPS) per 10 CFR 50.73(a)(2)(iv)(B)(1), an Emergency Core Cooling System (ECCS) pump per 10 CFR 50.73(a)(2)(iv)(B)(3), and an Emergency Diesel Generator (EDG) per 10 CFR 50.73(a)(2)(iv)(B)(8). However, pursuant to 10 CFR 50.73(a)(1), this event is being reported via this telephone notification, instead of
submitting a written Licensee Event Report, since the automatic actuation of the RPS, ECCS Pump and EDG was not generated by a valid actuation."

The licensee notified the NRC Resident Inspector.