Event Notification Report for May 22, 2000

LICENSEE REPORTED A STOLEN PORTABLE GAUGE.

The licensee reported on May 23 that at approximately 6:00 PM the previous evening a Seaman Nuclear C-200 gauge, serial number A-872, with 4.5 milligrams of Ra-226 was stolen. The licensee's technician had stopped by a friend's house in South Seattle on his way home. While there, his pickup truck was stolen from in front of the house. The gauge was locked, in a locked transport box, and the box was chained and locked in the bed of the truck. The licensee believes the thief wanted only the truck and will probably abandon the gauge. The police were notified and the licensee was advised to notify the local papers and post a reward.

Overexposures? Unknown.

Disposition/recovery: Unknown at this time.

* * * UPDATE ON 05/30/00 @ 1400 BY FRAZEE TO GOULD * * *

On May 27, 2000, the gauge was found intact and abandoned along a side street in South Seattle, Washington. (The abandoned pickup truck was eventually found elsewhere.) Police and Fire Department personnel responded to the scene and called 206-NUCLEAR (the Washington State radiation emergency line). The Division's Emergency Response Duty Officer determined that the gauge was not leaking, surveyed the exterior, and impounded it. The retrieval was covered by all the major Seattle television stations. The gauge will be returned to the licensee.

The R4DO (Sanborn) and NMSS EO ( Hickey) were notified by the NRC Operations Officer.

MISSING TRITIUM EXIT SIGN

During the semiannual inspection, one Permx model P160 series tritium exit sign was discovered missing. It is believed to have been destroyed during demolition of a wall as part of an expansion of the plant. A search of the plant was conducted and the contractors involved in the demolition were questioned with no real findings. A written report will be submitted to the NRC Region 2 office.

- ESF ACTUATION OF REACTOR WATER CLEANUP SYSTEM FOR UNKNOWN REASONS -

At 0253 on 05/22/00, the plant experienced an isolation of the Reactor Water Cleanup System on a Division 2 isolation signal when the Residual Heat Removal heat exchanger vent valve closed unexpectedly. An apparent electrical power supply spike to the Division 2 isolation instrumentation occurred. No testing or maintenance activities or electrical storms were occurring at the time of the isolation. The licensee is investigating the cause of the isolation.

The licensee notified the NRC Resident Inspector.


* * * UPDATE ON 5/23/00 @ 1317 BY SANFORD TO GOULD * * * RETRACTION

Upon further review, "the power supply fluctuation was determined to be caused by a failed capacitor in a Division 2 regulating transformer and the Division 2 electrical distribution subsystem was declared inoperable. The momentary power system perturbation caused a partial ESF isolation to be generated with the voltage on the Division 2 electrical distribution system remaining at the specified value. Since there was not an actual loss of power to the divisional subsystem that would have required an ESF actuation on loss of power, this is not a valid signal. Therefore, in accordance with 10 CFR 50.72(b)(2)(ii)(B), Reactor Water Clean-Up isolations from invalid signals are not reportable.

"Additionally, the Residual Heat Removal (RHR) Heat Exchanger Vent valve is a single component of a complex train (Both RHR and Isolation System) and does not, in itself, mitigate the consequences of a significant event. Therefore, in accordance with the guidance provided in NUREG 1022, Section 3.3.2, the vent valve closure is not reportable.

"The electrical distribution power supply was restored to Operable at 1701, May 22. 2000."

The NRC Resident Inspector was notified. The Reg 3 RDO (Hiland) was informed.

REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM MAY NOT BE PROTECTED BY STEAM LEAK DETECTION CIRCUITRY.

A LICENSEE ENGINEERING REVIEW IDENTIFIED A SECTION OF RCIC SYSTEM PIPING WHICH MAY NOT BE PROTECTED BY THE STEAM LEAK DETECTION CIRCUITRY. SPECIFICALLY, THE STEAM SUPPLY ISOLATION SIGNAL MAY NOT ISOLATE A POSTULATED CRACKED STEAM SUPPLY LINE PRIOR TO CRACK PROPAGATION AS IN THE DESIGN INTENT OF THE STEAM LEAK DETECTION SYSTEM. THE LICENSEE IS CONTINUING WITH ITS INVESTIGATION. THE RCIC SYSTEM WAS NOT DECLARED INOPERABLE AND THE LICENSEE HAS NOT ENTERED ANY LCO ACTION STATEMENTS.

THE NRC RESIDENT INSPECTOR WAS NOTIFIED.

* * * RETRACTED ON 6/19/00 AT 1254 EST BY MARK ABRAMSKI TO FANGIE JONES * * *

"A review of plant design and licensing basis documentation has confirmed that the design and licensing basis requirements for the SLD system as defined in the UFSAR have not been properly maintained. It is therefore not clear that the ability to detect and isolate a leak prior to the leak propagating to a break in this section of pipe is in fact a design basis requirement.

"LER 99-013 identified the condition where the SLD system was outside the design basis of the plant because the system could not reliably detect a 7 GPM leak (but could detect a 25 GPM leak). The review of plant design and licensing basis documentation discussed above has identified the need to revise corrective actions identified in LER 99-013 to include a more comprehensive redefinition of the design and licensing basis requirements for the SLD system. The event reported in 10 CFR 50.72 notification log number 37016 is therefore an extent of the condition reported in LER 99-013 and will therefore be discussed in the revision to LER 99-013.

"On this basis 10 CFR 5072 notification log number 37016 is redundant to the condition reported in LER 99-013 and 10 CFR 50.72 notification log number 37016 is therefore retracted."

The licensee notified the NRC Resident Inspector. The R1DO (Robert Summers) was notified.

POSITIVE TEST FOR ALCOHOL

AN ENTERGY EMPLOYEE TESTED POSITIVE FOR ALCOHOL FOLLOWING A FOR-CAUSE TEST. UPON NOTIFICATION, THE EMPLOYEE'S PLANT ACCESS AUTHORIZATION WAS SUSPENDED AT RIVER BEND STATION PENDING FURTHER INVESTIGATION.

THE NRC RESIDENT INSPECTOR WILL BE NOTIFIED.

- AUTO Rx SCRAM FROM 100% AFTER REPLACEMENT OF MAIN TURBINE CONTROL VALVE SOLENOID -

On 05/21/00, the Unit 2 Main Turbine #1 Control Valve Fast Acting Solenoid failed its required Tech Spec Surveillance Test.

At 2159 CDT on 05/22/00, during return to service activities following replacement of this solenoid, Unit 2 automatically scrammed from 100% power due to receipt of an APRM High-High RPS Actuation signal. All control rods inserted completely. No safety/relief valves lifted. Steam is being dumped to the main condenser. During the transient, reactor vessel water level dropped to +8 inches (normal level is +30 inches) and all PCIS Group II actuations occurred, as expected. These actuations included Reactor Building Ventilation Valves closed, Train 'A' Standby Gas Treatment System auto started (Train 'B' SBGT System was inoperable), Control Room Ventilation System isolated, and Containment Isolation Valves closed. Quad Cities has one Emergency Diesel Generator (EDG) for each unit and a shared EDG. The Unit 2 EDG received a spurious auto start signal during the 4 KV auxiliary bus power transfer to the reserve transformer. The licensee secured the Unit 2 EDG.

Unit 2 is now stable in Condition 3 (Hot Shutdown) with reactor vessel water level within its normal band.

This event had no effect on Unit 1 which is at 100% power.

The licensee is investigating the cause of this reactor scram.

The licensee notified the NRC Resident Inspector.

* * * UPDATE AT 0708 ON 05/23/00 BY JOHN LECHMAIER TO JOLLIFFE * * *

During the above transient, the reactor vessel water level reached +8 inches and all PCIS Group II actuations occurred, as expected. The reactor vessel water level then increased to +48 inches, the Reactor Feedwater Pump trip level. At 2215 CDT, the Reactor Water Cleanup (RWCU) System was placed in service to assist in reactor vessel water level control. The reactor vessel water level decreased again and subsequently, at 2229 CDT, reached +8 inches. The RWCU System isolated and all PCIS Group II actuations occurred.

Unit 2 remains stable in Condition 3 (Hot Shutdown) with reactor vessel water level within its normal band.

The licensee notified the NRC Resident Inspector.

The NRC Operations Officer notified the R3DO Pat Hiland.

- HIGH PRESSURE COOLANT INJECTION SYSTEM AUTO ISOLATED WHILE BRINGING IT ON LINE -

At 2338 on 05/22/00, the High Pressure Coolant Injection (HPCI) System isolated due to a spurious high steam flow isolation signal. A valid high steam line flow condition would normally be indicative of a piping break; however, a piping break did not occur. Although the signal was not valid, this is nonetheless considered an Engineered Safety Features (ESF) Actuation since a containment isolation had occurred.

The Hope Creek reactor was in Operational Condition 2 (Startup) with Reactor Coolant System (RCS) pressure at 200 psig and power at about 4%. The HPCI System steam line warmup, which is required to place the HPCI System in a standby alignment, had just commenced. When shutdown, the HPCI steam line is isolated from the reactor by three containment isolation valves. These valves are an outboard valve, an inboard valve, and a bypass valve around the inboard valve for steam line warm-up purposes. As part of the steam line warmup procedure, the outboard valve is fully opened and the inboard bypass valve is throttled to slowly heat up and pressurize the steam line. Once the steam line is pressurized, the inboard valve is opened.

During the initial phases of the steam line warmup process, a high steam line flow signal was generated when the bypass valve was throttled open. This occurred due to steam pressure and flow perturbations within the steam line. As a result, an isolation signal was generated and the bypass valve automatically closed as expected for this isolation signal. Prior to resetting the isolation signal, the HPCI steam line was verified to be operable. The HPCI System warmup was then recommenced and the HPCI System is currently in its standby alignment.

The licensee notified the NRC Resident Inspector.