Event Notification Report for March 02, 1999

- MANUAL Rx TRIP FROM 100% DUE TO CLOSURE OF LOOP 3 MAIN FEEDWATER ISOLATION VALVE -

WHILE IN THE PROCESS OF HANGING AN OUTAGE CLEARANCE TAG ON UNIT 1 EQUIPMENT (SEE EVENT #35421), PLANT TECHNICIANS ERRONEOUSLY PULLED THE FUSES TO THE UNIT 2 LOOP 3 MAIN FEEDWATER ISOLATION VALVE (#2HV-5229) TO #3 STEAM GENERATOR CAUSING THE VALVE TO CLOSE. CONTROL ROOM OPERATORS OBSERVED THE LOW WATER LEVEL IN #3 STEAM GENERATOR AND THE STEAM FLOW/FEED FLOW MISMATCH ANNUNCIATORS.

AT 0206 ON 03/02/99, CONTROL ROOM OPERATORS MANUALLY TRIPPED THE REACTOR FROM 100% POWER. ALL CONTROL RODS INSERTED COMPLETELY. THE AUXILIARY FEEDWATER SYSTEM ACTUATED, AS EXPECTED. STEAM IS BEING DUMPED TO THE MAIN CONDENSER.

DURING THE TRIP, AN ELECTRO-HYDRAULIC CONTROL (EHC) SYSTEM POWER SUPPLY FAILED CAUSING ERRONEOUS INDICATIONS OF MAIN CONDENSER VACUUM AND EHC PRESSURE.

UNIT 2 IS STABLE IN MODE 3 (HOT STANDBY).

THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR.

ICE CONDENSER DESIGN BASIS BYPASS FLOW COULD BE EXCEEDED FOLLOWING A STEAM LINE RUPTURE AND STRUCTURAL FAILURE IN THE STEAM GENERATOR COMPARTMENT.

"This is Don Kosloff at American Electric Power, DC Cook Plant, calling with a notification under 10CFR50.72 (b)(2)(i), a condition which was found while the reactor was shutdown, which if it had been found while the reactor was operating, would have resulted in the nuclear power plant being in an unanalyzed condition, outside the design basis.

"The Cook units are pressurized water reactors with ice condenser containments. The ice condenser containment requires that the steam and air flowing from the lower containment compartment in the event of an accident be routed to the upper containment compartment through the ice condenser ice bed. To accomplish this, a structural barrier, called the divider barrier, separates the lower and upper containment compartments. The divider barrier includes the walls of the ice compartment, the upper deck, the gates separating the reactor cavity from the refueling canal, portions of the walls of the refueling canal, and the compartments enclosing the upper portion of the steam generators and pressurizer.

"The design basis limits bypass flow around the ice bed to no more than 5 square feet. For Unit 1, the installed bypass through the refueling cavity drains is equivalent to 2.2 square feet plus approximately 0.36 square feet bypass via the backdraft damper of the air return fans, totaling 2.56 square feet of bypass area. The construction of Unit 2 is similar. On August 12, 1998, it was determined that the cumulative bypass area for unit 1, had been equivalent to 5.71 square feet. This condition was reported on September 11,1998 in LER 50-315/98-037. During a recent review of existing structural design calculations, AEP personnel determined that there was a potential for a steam line break to overstress the associated steam generator compartment. Such a steam line break inside a steam generator compartment could cause an additional bypass opening through the reinforced concrete walls of the compartment. Such an opening would increase the cumulative effective bypass opening. The size of such an opening has not been determined. Therefore, the safety significance of this potential condition cannot yet be determined. Additional calculations are planned."

The resident inspector has been informed.

Plant reported a loss of emergency response capability.

The Emergency Operations Facility requires three shifts of personnel. Each shift requires 7 communicators. All 21 communicators participated in EP drills. However, only 5 of the 21 have completed qualification documentation. Should an event occur that requires use of the EOF, the EP supervisor will mentor the unqualified communicators. An additional corrective action will be reinstatement of individuals previously qualified. Both Units are currently in Mode 5.


The Resident Inspector was informed.


*** UPDATE ON 3/2/99 @ 2240 BY DEPUYDT TO GOULD ***

THIS UPDATE IS TO CLARIFY THE INITIAL INFORMATION PROVIDED.

EACH EOF "TEAM" HAS 12 CRITICAL POSITIONS. OF THESE 12 POSITIONS, 7 ARE COMMUNICATOR POSITIONS. FOR THE 3 EOF TEAMS THAT COOK HAS, A TOTAL OF 21 QUALIFIED COMMUNICATORS ARE REQUIRED.

CURRENTLY, THEY HAVE ONLY 5 FULLY QUALIFIED COMMUNICATORS OUT OF THE REQUIRED 21, AND OF THOSE 5, FOUR ARE ALL QUALIFIED FOR THE SAME POSITION.

THEREFORE, DUE TO THE SHORTAGE OF QUALIFIED COMMUNICATORS TO MAN THE REQUIRED CRITICAL POSITIONS, THE EOF CANNOT BE MANNED WITH A FULL TEAM OF QUALIFIED INDIVIDUALS. AS AN INTERIM ACTION, SHOULD AN EVENT OCCUR THAT REQUIRES THE ACTIVATION OF THE EOF, THE INDIVIDUALS CURRENTLY ON THE CALL OUT LIST FOR THOSE COMMUNICATOR POSITIONS WOULD BE CALLED OUT AND THE EMERGENCY PLANNING SUPERVISOR WOULD MENTOR THOSE UNQUALIFIED INDIVIDUALS IN THEIR DUTIES.

WITHIN THE NEXT 24 HOURS, CORRECTIVE ACTION WILL BE IMPLEMENTED BY REQUALIFICATION AND REINSTATEMENT OF THE INDIVIDUALS WHO WERE PREVIOUSLY QUALIFIED FOR THE COMMUNICATOR POSITIONS.

THE RESIDENT INSPECTOR WILL BE NOTIFIED.

THE REG 3 RDO(CREED) WAS NOTIFIED.

A SOIL MOISTURE DENSITY GAUGE INVOLVED IN A MAJOR VEHICLE ACCIDENT.

A TROXLER GAUGE MODEL # 3440 (SERIAL #21764 CONTAINING 8 mCi OF Cs-137 AND 40 mCi OF
Am-241:Be) WAS TRANSPORTED IN A VEHICLE THAT WAS INVOLVED IN A MAJOR TRAFFIC ACCIDENT INVOLVING 3 FATALITIES. THERE WAS NO APPARENT DAMAGE TO THE GAUGE AND ITS CASE. THE GAUGE WILL BE LEAK TESTED AND RESULTS WILL BE FORWARDED TO THE STATE RADIATION CONTROL OFFICE. THERE MAY BE MEDIA INTEREST.

TROXLER GAUGE FOUND IN SHOPPING MALL IN WEST MEMPHIS ARKANSAS

A PRIVATE CITIZEN FOUND A TROXLER GAUGE MODEL 3411B AT A SHOPPING MALL IN WEST MEMPHIS, ARKANSAS. THE CITIZEN REPORTED IT TO THE POLICE AND THE WEST MEMPHIS FIRE DEPT HAZMAT TEAM RESPONDED TO THE MALL. THE GAUGE, WHICH BELONGED TO A TENNESSEE LICENSEE, WAS TURNED INTO THE ARKANSAS DIVISION OF RADIATION CONTROL. THE STATE OF TENNESSEE DEPARTMENT OF RADIOLOGICAL HEALTH HAS BEEN NOTIFIED BY THE DIVISION OF RADIATION CONTROL OF THE ARKANSAS DEPARTMENT OF HEALTH. TENNESSEE WILL HAVE THEIR LICENSEE RETRIEVE THE GAUGE FROM THE ARKANSAS DIVISION OF RADIATION CONTROL. THERE APPEARED TO BE NO DAMAGE TO THE GAUGE OR ITS CASE AND NO ABNORMAL RADIATION LEVELS WERE MEASURED. THERE WILL BE MEDIA INTEREST.

A POTENTIAL ERROR WITH THE DOSE TO CONTROL ROOM PERSONNEL WAS IDENTIFIED

CR-99-0127 IDENTIFIES A POTENTIAL ERROR WITH THE DOSE TO CONTROL ROOM PERSONNEL DURING A POSTULATED RESIN FIRE ACCIDENT AS DESCRIBED IN THE LICENSEE'S UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR). THE MAXIMUM ORGAN DOSE REPORTED IN THE UFSAR (SECTION 6.4.4.1) IS 3.83 REM FOR BONE. USING THE SAME CONSERVATIVE SOURCE-TERM ASSUMPTIONS GIVEN IN THE UFSAR AND THE DOSE CONVERSION FACTORS FROM FEDERAL GUIDANCE REPORT 11, THE PRELIMINARY RECALCULATED DOSE IS 54 REM CDE (BONE) AND 3.2 REM CEDE. THE 54 REM BONE DOSE EXCEEDS THE 50 REM LIMIT PRESCRIBED IN 10CFR50 APPENDIX A, GENERAL DESIGN CRITERION (GDC) 19, "CONTROL ROOM."

THIS ERROR IS SPECIFIC TO THE CALCULATED CONTROL ROOM DOSE FROM THE RESIN FIRE ACCIDENT SCENARIO. A SIMILAR ANALYSIS WAS PERFORMED FOR THE RESIN FIRE ACCIDENT SCENARIO TO CALCULATE THE OFFSITE DOSE TO MEMBERS OF THE PUBLIC. THIS CALCULATION USED A DIFFERENT METHODOLOGY THAN THE CONTROL ROOM DOSE ANALYSIS AND HAS BEEN VERIFIED.

IF THE PLANT OPERATED AS PRESENTLY ALLOWED, THE POTENTIAL EXPOSURE TO CONTROL ROOM OPERATORS COULD EXCEED THE DESIGN BASIS OF GDC 19 FOR THE LIMITING RADIOLOGICAL
EVENT, A SPENT RESIN FIRE. IN ACTUAL PRACTICE, THE REQUIRED RADIOACTIVE CONTENT OF RESIN DOES NOT EXIST AT ANY ONE PLACE ON SITE, THEREFORE, AN ACTUAL EVENT EXCEEDING THE GDC 19 CRITERIA IS NOT CURRENTLY POSSIBLE.

THE RESIDENT INSPECTOR WILL BE INFORMED.

* * * UPDATE AT 1813 ON 3/18/99, BY BOWER, TAKEN BY WEAVER * * *
The licensee is retracting this event based on a new calculation using the actual curie content of the contamination in the resin. Best estimates put the bone dose due to a resin fire at slightly over 20 rem, well below the 50 rem limit. Hoo notified R1DO (McFadden).

THE LICENSEE IS MISSING A CONSISTENCY GAUGE FOR A PAPER MACHINE

THE GAUGE, NDC SYSTEMS MODEL # 104, CONTAINS 80 MILLICURIES OF AM-241 AND IS USUALLY MOUNTED TO THE PAPER MACHINE. IT WAS LAST INVENTORIED 6 MONTHS AGO WHEN IT WAS LOCATED IN A BOX IN ONE OF THEIR OFFICES. THEY MAY BE ABLE TO LOCATE IT WHEN THEY CAN CONTACT AN EMPLOYEE WHO MAY KNOW WHERE IT IS.

* * * UPDATE 0947EST ON 3/6/99 FROM LARRY LAIRD TO S.SANDIN * * *

THE LICENSEE LOCATED THE MISSING SOURCE. IT WAS FOUND IN THE ORIGINAL SHIPPING CONTAINER STORED IN THEIR TEAM ROOM. NOTIFIED R1DO(MODES) AND EO(HODGES).

AUXILIARY FEEDWATER LEAKING BACK INTO MAIN FEEDWATER SYSTEM

THE LICENSEE DISCOVERED THAT THE AUXILIARY FEEDWATER SYSTEM WAS LEAKING BACK INTO THE MAIN FEED SYSTEM ("B" FEED LINE) PAST TWO IN-SERIES CHECK VALVES. THESE CHECK VALVES ARE CONSIDERED CONTAINMENT ISOLATION VALVES. LEAKAGE IS IN EXCESS OF THEIR ALLOWABLE LIMITS. AUXILIARY FEEDWATER IS CURRENTLY FEEDING THE STEAM GENERATORS. THE LEAKAGE HAS BEEN ISOLATED USING MANUAL VALVES. THE UNIT WAS IN STARTUP FOLLOWING A REFUELING OUTAGE WHEN THE CONDITION WAS FOUND.

THE RESIDENT INSPECTOR WILL BE NOTIFIED.


***RETRACTION ON 03/24/99 AT 1132ET BY RANDY HASTINGS TAKEN BY MACKINNON***

Subsequent evaluation has clarified that only one of the series check valves performs a containment isolation function, 2CS-476AA (Reference PBNP FSAR Figure 5.2-3). The main steam and feed water system are considered a closed system inside containment and the piping is part of the containment boundary. That boundary remained intact during this event. Accordingly, the integrity of the containment boundary was never in question and this condition did not result in a principal safety barrier being seriously degraded.

An engineering evaluation has also been completed to assess whether the leakage observed past these check valves would have resulted in less than the required auxiliary feed water flow to be available for decay heat removal. This evaluation concluded that even with check valve leakage in excess of that observed during this event, the auxiliary feed water system was capable of providing sufficient flow to the steam generators to meet the design basis for decay heat removal. R3DO (Pat Hiland) notified.

The NRC Resident Inspector was notified of this event by the licensee.