U.S. Nuclear Regulatory Commission Operations Center Event Reports For 11/06/1998 - 11/09/1998 ** EVENT NUMBERS ** 34896 34907 34979 35003 35004 35005 35006 35007 35008 35009 35010 35011 !!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!! +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 34896 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: BRUNSWICK REGION: 2 |NOTIFICATION DATE: 10/09/1998| | UNIT: [1] [] [] STATE: NC |NOTIFICATION TIME: 18:00[EDT]| | RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 10/09/1998| +------------------------------------------------+EVENT TIME: 15:00[EDT]| | NRC NOTIFIED BY: DAVE JESTER |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |CHRIS CHRISTENSEN RDO | |10 CFR SECTION: | | |AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION | | |NLCO TECH SPEC LCO A/S | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |1 N Y 100 Power Operation |100 Power Operation | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | - STANDBY LIQUID CONTROL SYSTEM INOPERABLE DUE TO LEAKING DRAIN LINE END CAP | | - | | | | AN END CAP ON THE DRAIN LINE FROM THE INOPERABLE UNIT 1 STANDBY LIQUID | | CONTROL (SLC) SYSTEM DRAIN VALVE #1-C41-F015 HAD PREVIOUSLY BEEN INSTALLED | | TO PREVENT LEAKAGE. MAINTENANCE PERSONNEL WERE CUTTING INTO THE DRAIN LINE | | IN ORDER TO REPLACE THE LEAKING END CAP. UPON CUTTING THE PIPE, WATER | | FLOWED OUT FROM THE CUT AND SHORTED OUT THE CUTTING DEVICE. PLANT | | PERSONNEL CLOSED THE SLC STORAGE TANK SUPPLY VALVE #1-C41-F001 IN ORDER TO | | ISOLATE THE LEAK TO COMPLETE THE END CAP REPLACEMENT. THIS ACTION DISABLED | | BOTH TRAINS OF THE SLC SYSTEM DUE TO A COMMON CAUSE FAILURE. | | | | AT 1648 ON 10/09/98, THE LICENSEE DECLARED THE SLC SYSTEM INOPERABLE. TECH | | SPEC LCO A/S 3.1.7 REQUIRES THE LICENSEE TO RESTORE THE SLC SYSTEM TO | | OPERABLE STATUS WITHIN 8 HOURS. | | | | AN EVALUATION OF THE SAFETY SIGNIFICANCE OF THIS EVENT IS THAT THE SLC | | SYSTEM WOULD NOT HAVE BEEN CAPABLE OF PERFORMING ITS INTENDED SAFETY | | FUNCTION TO MITIGATE THE CONSEQUENCES OF AN ACCIDENT, i.e., TO SHUT THE | | REACTOR DOWN DURING AN ANTICIPATED TRANSIENT WITHOUT SCRAM. | | | | THE LICENSEE HAS REMOVED AND REPLACED THE OLD END CAP AND PRESENTLY IS IN | | THE PROCESS OF FILLING, VENTING AND PRESSURE TESTING THE SLC SYSTEM. | | | | THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. | | | | *** UPDATE ON 11/06/98 @1045 BY ELBERFELD TO GOULD *** RETRACTION | | | | FURTHER EVALUATION OF THIS EVENT HAS DETERMINED THAT, THROUGHOUT THE | | MAINTENANCE EVOLUTION, LEAKAGE THROUGH THE SUCTION HEADER DRAIN ISOLATION | | VALVE WOULD NOT HAVE ADVERSELY AFFECTED THE SLC SYSTEM'S ABILITY TO SHUT | | DOWN THE REACTOR. FURTHERMORE, THE SLC SYSTEM WAS PROPERLY REMOVED FROM | | SERVICE AND THE APPROPRIATE TS LCO 3.1.7 ENTERED WHEN IT WAS DETERMINED THAT | | THE REPAIRS COULD BE MADE IN A MORE CONTROLLED MANNER WITH THE SLC STORAGE | | TANK ISOLATED. DURING THE PERIOD OF THE LCO FOR SLC SYSTEM OPERABILITY, AN | | AUXILIARY OPERATOR WAS AVAILABLE TO REOPEN THE STORAGE TANK OUTLET ISOLATION | | VALVE IF REQUIRED. THEREFORE, IT WAS DETERMINED THE SAFETY FUNCTION OF THE | | SLC SYSTEM WAS NOT LOST AND THE EVENT IS BEING RETRACTED. | | | | THE RESIDENT INSPECTOR WAS NOTIFIED. | | | | THE REG 2 RDO (FREDRICKSON) WAS INFORMED. | +------------------------------------------------------------------------------+ !!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!! +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 34907 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: ROBINSON REGION: 2 |NOTIFICATION DATE: 10/13/1998| | UNIT: [2] [] [] STATE: SC |NOTIFICATION TIME: 09:00[EDT]| | RXTYPE: [2] W-3-LP |EVENT DATE: 10/13/1998| +------------------------------------------------+EVENT TIME: 08:00[EDT]| | NRC NOTIFIED BY: HAROLD CHERNOFF |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: FANGIE JONES +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |CHRIS CHRISTENSEN RDO | |10 CFR SECTION: | | |DDDD 73.71 UNSPECIFIED PARAGRAPH | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |2 N Y 100 Power Operation |100 Power Operation | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | PLANT SECURITY EVENT | | | | SAFEGUARD SYSTEM DEGRADATION RELATED TO DETECTION/POWER SUPPLY FUNCTIONS. | | IMMEDIATE COMPENSATORY MEASURES TAKEN UPON DISCOVERY. THE LICENSEE | | INFORMED THE NRC RESIDENT INSPECTOR. | | | | REFER TO THE HOO LOG FOR ADDITIONAL DETAILS. | | | | ***UPDATE ON 10/13/98 AT 1624 EDT FROM CHERNOFF TAKEN BY MACKINNON*** | | | | THREE AREAS FOUND THAT DID NOT HAVE PROPER PROTECTED AREA/VITAL AREA | | DEFENSES. IMMEDIATE COMPENSATORY MEASURES TAKEN UPON DISCOVERY. THE | | LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. R2DO (MARK LESSER) NOTIFIED. | | | | REFER TO THE HOO LOG FOR ADDITIONAL DETAILS. | | | | * * * RETRACTION 1420 11/6/98 FROM COOK/CHERNOFF TAKEN BY STRANSKY * * * | | | | "This is a retraction of an event notification made on October 13, 1998, at | | 0912 hours (NRC Event No. 34907) and a follow up to Event No. 34907 made on | | October 13, 1998, at 1624 hours. These notifications concerned a condition | | that was based on the belief that vital equipment was discovered not | | properly protected. Upon discovery of this event an investigation was | | initiated to determine if the areas in question contained vital equipment. | | | | "Based on detailed review of the current Physical Security and Safeguards | | Contingency Plan and related correspondence, it has been determined that the | | areas in question do not contain vital equipment. | | | | "Based on the fact that equipment in the subject areas is not vital in | | accordance 10 CFR 73, the one hour report originally made in accordance with | | 10 CFR 73.71, is hereby retracted." | | | | THE NRC RESIDENT INSPECTOR HAS BEEN INFORMED OF THIS RETRACTION. NOTIFIED | | R2DO (PAUL FREDRICKSON). | +------------------------------------------------------------------------------+ !!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!! +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 34979 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: HATCH REGION: 2 |NOTIFICATION DATE: 11/01/1998| | UNIT: [] [2] [] STATE: GA |NOTIFICATION TIME: 17:00[EST]| | RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 11/01/1998| +------------------------------------------------+EVENT TIME: 14:00[EST]| | NRC NOTIFIED BY: VARNADORE |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |THOMAS PEEBLES R2 | |10 CFR SECTION: | | |AESF 50.72(b)(2)(ii) ESF ACTUATION | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ | | | |2 N N 0 Cold Shutdown |0 Cold Shutdown | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | INVALID REACTOR WATER CLEANUP (RWCU) ISOLATION SIGNAL. | | | | WHILE ATTEMPTING TO SEAL VALVE SEAT LEAKAGE ON A DRAIN VALVE FOR THE RWCU | | HEAT EXCHANGER, AN OPERATOR CRACKED OPEN THE VALVE WHICH RESULTED IN A HIGH | | DIFFERENTIAL FLOW SIGNAL CAUSING RWCU ISOLATION. DURING THIS ISOLATION, | | ONE OF THE ISOLATION VALVES DID NOT CLOSE AS REQUIRED AFTER RECEIVING THE | | ISOLATION SIGNAL. AN INVESTIGATION IS BEING MADE INTO WHY THE VALVE DID NOT | | CLOSE. THE ISOLATION WAS RESET WITHIN 48 MINUTES. THE RESIDENT INSPECTOR | | WAS NOTIFIED. | | | | *** UPDATE ON 11/06/98 @ 0926 BY GORLEY TO GOULD *** RETRACTION | | | | PLANT ENGINEERS HAVE DETERMINED THAT THE SYSTEM FUNCTIONED NORMALLY. THE | | VALVES ARE CONTROLLED BY SEPARATE CHANNELS OF LOGIC CALIBRATED TO SLIGHTLY | | DIFFERENT TRIP SETPOINTS. THE MAGNITUDE OF THE DIFFERENTIAL FLOW FELL | | BETWEEN THE TWO SETPOINTS, SO ONE VALVE CLOSED AS IT SHOULD HAVE. THE TRIP | | SIGNAL FOR THE OTHER VALVE DID NOT STAY ABOVE THE SETPOINT LONG ENOUGH FOR | | ITS TIME DELAY TO EXPIRE, SO THE VALVE REMAINED OPEN. A SUBSEQUENT FUNCTION | | TEST SUCCESSFULLY VERIFIED THE OPERABILITY OF THE VALVE WHICH AT FIRST | | APPEARED TO HAVE FAILED. IN ADDITION, THE REACTOR WATER CLEANUP | | DIFFERENTIAL FLOW SIGNAL IS NOT AN ENGINEERED SAFETY FEATURE PER THE HATCH | | UNIT 2 FSAR SECTION 7.3.2.2.2. FOR THESE REASONS THE EVENT NOTIFICATION IS | | BEING RETRACTED. | | | | THE RESIDENT INSPECTOR WAS NOTIFIED. REG 2 RDO (FREDRICKSON) WAS NOTIFIED. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35003 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: FARLEY REGION: 2 |NOTIFICATION DATE: 11/06/1998| | UNIT: [1] [] [] STATE: AL |NOTIFICATION TIME: 10:00[CST]| | RXTYPE: [1] W-3-LP,[2] W-3-LP |EVENT DATE: 11/06/1998| +------------------------------------------------+EVENT TIME: 06:00[CST]| | NRC NOTIFIED BY: LERO |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |PAUL FREDRICKSON R2 | |10 CFR SECTION: | | |NINF INFORMATION ONLY | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |1 N N 0 Refueling |0 Refueling | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | THE LICENSEE DETERMINED >1% OF THE TUBES IN THE 1A STEAM GENERATOR (SG) ARE | | DEFECTIVE. | | | | WHILE PERFORMING A STEAM GENERATOR INSPECTION, AS PART OF THEIR REFUELING | | OUTAGE, >1% OF THE TUBES IN THE 1A STEAM GENERATOR WERE FOUND TO BE | | DEFECTIVE. THIS PUTS THE 1A SG INSPECTION RESULTS IN CATEGORY C-3 AND | | REQUIRES THE LICENSEE TO NOTIFY THE NRC PER TS. | | | | THE LICENSEE HAS CURRENTLY COMPLETED INSPECTION ON 40% OF 1A SG, 10% OF 1B | | SG AND 25% OF 1C SG. AT THIS TIME IT IS THOUGHT THE DEFECTS ON THE 1A SG | | TUBES MAY BE A COMBINATION OF OD AND ID STRESS CORROSION CRACKING, | | PRIMARILY IN THE TUBE EXPANSION AREA OF THE TUBE SHEET AND SLUDGE PILE | | AREA. | | | | THE LICENSEE PLANS TO INSPECT 100% OF ALL TUBES IN EACH STEAM GENERATOR AND | | PLUG OR REPAIR DEFECTIVE TUBES AS REQUIRED. | | | | THE RESIDENT INSPECTOR WAS NOTIFIED. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35004 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: MILLSTONE REGION: 1 |NOTIFICATION DATE: 11/06/1998| | UNIT: [] [] [3] STATE: CT |NOTIFICATION TIME: 13:00[EST]| | RXTYPE: [1] GE-3,[2] CE,[3] W-4-LP |EVENT DATE: 11/06/1998| +------------------------------------------------+EVENT TIME: 12:00[EST]| | NRC NOTIFIED BY: GENE OLSON |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: BOB STRANSKY +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |CLIFFORD ANDERSON R1 | |10 CFR SECTION: | | |APRE 50.72(b)(2)(vi) OFFSITE NOTIFICATION | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ | | | | | | |3 N Y 5 Startup |5 Startup | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | OFFSITE NOTIFICATION REGARDING SITE NPDES PERMIT | | | | THE LICENSEE'S ENVIRONMENTAL ENGINEERING GROUP CONTACTED THE STATE OF | | CONNECTICUT REGARDING PROPOSED REVISIONS TO CALCULATIONS USED IN REPORTING | | VARIOUS NATIONAL POLLUTION DISCHARGE ELIMINATION SYSTEM (NPDES) PERMIT | | PARAMETERS. THE NRC RESIDENT INSPECTOR WILL BE INFORMED OF THIS REPORT BY | | THE LICENSEE. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35005 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: CRYSTAL RIVER REGION: 2 |NOTIFICATION DATE: 11/06/1998| | UNIT: [3] [] [] STATE: FL |NOTIFICATION TIME: 15:00[EST]| | RXTYPE: [3] B&W-L-LP |EVENT DATE: 11/06/1998| +------------------------------------------------+EVENT TIME: 12:00[EST]| | NRC NOTIFIED BY: WILLIAM KISNER |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: BOB STRANSKY +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |KERRY LANDIS R2 | |10 CFR SECTION: | | |AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |3 N Y 100 Power Operation |100 Power Operation | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | POTENTIAL FOR DECAY HEAT REMOVAL FLOW CONTROL VALVE MISPOSITIONING FOLLOWING | | TESTING | | | | "In order to start during an Engineered Safeguards actuation, the Decay Heat | | Removal Pumps require a 500 psi / LPI actuation signal coincident with a | | START permissive from Block 4 of the 1500 psi / HPI actuation sequence. If a | | large break LOCA occurred coincident with a loss offsite power, the 500 psi | | actuation would occur almost immediately. However, the DH Pumps would not | | receive a START permissive for twenty-five seconds. The twenty-five second | | delay is based on a ten second start time for the Emergency Diesel | | Generators and the High Pressure Injection block loading sequence. The | | DH/LPI flow control valves, DHV-110 and DHV-111, are not released to control | | until thirty seconds after the DH Pumps start. Therefore, the valves will | | not begin to control LPI flow until fifty-five seconds into the event. | | | | "Large Break LOCA analyses assume that full LPI flow exists thirty-five | | seconds into the event. If a flow control valve is not in a position which | | provides the required flow at the moment the accident occurs, its associated | | DH/LPI train will not provide the required flow at the time assumed in | | accident analyses. If, during standby, the valve is not opened far enough, | | the train will provide a reduced flow until after the valve is released to | | control. If the control valve is too far open during standby, the train will | | provide excessive flow until the valve begins to control. | | | | "Decay Heat Removal System surveillance tests are governed by Surveillance | | Procedures SP-340B and SP-340E. Each procedure includes instructions for | | verifying the full stroke of DHV-110 and DHV-111. At the end of valve stroke | | verification, the procedures direct Operators to close the valves, and then | | throttle them open for ten to fifteen seconds. Based on stroke time data for | | DHV-110 and DHV-111, this would leave the valves between eleven and | | seventeen percent open. This valve position is not sufficient to provide the | | required LPI flow. However, instructions for the DH Pump tests require that | | pump flow remain between 2940 gpm and 3060 gpm. During testing, the DH/LPI | | control valves maintain this flow. At the conclusion of DH Pump tests, the | | procedures direct Operators to secure the pumps, but do not direct Operators | | to reposition DHV-110 and DHV-111. Therefore, the valves would be left in a | | position which would provide adequate flow. | | | | "The DH Pumps may be used to recirculate BWST prior to sampling the tank in | | accordance with SP-320. Operators recirculate the tank in accordance with | | SP-340B/E, or OP-404. During BWST recirculation, DHV-110 and DHV-111 | | automatically control flow. At the conclusion of BWST recirculation, the | | procedures direct Operators to secure the pumps, but do not direct Operators | | to reposition the control valves. Once again, DHV-110 and DHV-111 would be | | left in a position which would provide adequate flow. | | | | "The most likely cause for DHV-110 or DHV-111 to be incorrectly positioned | | is post maintenance valve stroking. If the valves are not properly | | positioned following maintenance, they will not provide adequate flow. | | | | "Decay Heat Removal System surveillance were last performed in September of | | 1998. The 'A' train test, SP-340B, was performed on 9/24/98. The 'B' train | | test, SP-340E, was performed on 9/4/98. Based on review of plant logs and | | maintenance records, neither DHV-110 nor DHV-111 has been repositioned since | | the completion of surveillance testing. Therefore, there is reasonable | | assurance that both valves are properly positioned, that both trains will | | provide the required flow during an accident, and both DH/LPI trains are | | operable." | | | | THE NRC RESIDENT INSPECTOR HAS BEEN INFORMED BY THE LICENSEE. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35006 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: SUMMER REGION: 2 |NOTIFICATION DATE: 11/06/1998| | UNIT: [1] [] [] STATE: SC |NOTIFICATION TIME: 16:00[EST]| | RXTYPE: [1] W-3-LP |EVENT DATE: 11/06/1998| +------------------------------------------------+EVENT TIME: 15:00[EST]| | NRC NOTIFIED BY: MICAEL ZACCONE |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: BOB STRANSKY +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |KERRY LANDIS R2 | |10 CFR SECTION: | | |AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |1 N Y 100 Power Operation |100 Power Operation | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | POTENTIAL TO EXCEED 10 CFR PART 100 DOSE LIMITS DUE TO SEISMIC EVENT DURING | | LOCA | | | | "Design Engineering performed a Virgil C. Summer Nuclear Station (VCSNS) | | specific analysis which shows that l0CFR100 limits could be violated due to | | a seismic design error in the ECCS system. This analysis shows that in a | | seismic event, the 3/4" line to Relief Valve XVR-08116B would exceed its | | design limit, resulting in leakage sufficient to exceed 10CFR 100 offsite | | dose limits during a Loss of Coolant Accident (LOCA) with fuel damage. This | | design condition has existed since original plant design. | | | | "This issue is REPORTABLE under 10 CFR 50.72 (b)(1)(ii)(B) as 'Any Event or | | condition during operation that results in the nuclear power plant being in | | a condition that is outside the design basis of the plant.' The basis for | | this conclusion is that the postulated pipe failure would result in primary | | coolant leakage into the Auxiliary Building of a magnitude that if fuel | | damage (assumed in plant accident analysis) has occurred, 10CFR100 offsite | | dose limits would be exceeded. | | | | "Design Engineering performed a VCSNS specific seismic and support analysis | | and concluded that the removal of the incorrectly designed supports will | | reduce the stress to within code allowables and that the pipe will remain | | properly supported. A Non-conformance Notice (NCN) has been written under | | CER 98-01014 to disposition the Engineering recommendations." | | | | THE NRC RESIDENT INSPECTOR HAS BEEN INFORMED OF THIS CONDITION BY THE | | LICENSEE. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35007 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: PEACH BOTTOM REGION: 1 |NOTIFICATION DATE: 11/06/1998| | UNIT: [2] [3] [] STATE: PA |NOTIFICATION TIME: 16:00[EST]| | RXTYPE: [2] GE-4,[3] GE-4 |EVENT DATE: 11/06/1998| +------------------------------------------------+EVENT TIME: 15:00[EST]| | NRC NOTIFIED BY: PAT NAVIN |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: BOB STRANSKY +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |CLIFFORD ANDERSON R1 | |10 CFR SECTION: | | |HFIT 26.73 FITNESS FOR DUTY | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ |2 N Y 100 Power Operation |100 Power Operation | |3 N Y 100 Power Operation |100 Power Operation | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | FITNESS FOR DUTY REPORT | | | | A NON-LICENSED SUPERVISOR TESTED POSITIVE FOR ILLEGAL DRUG (NOT SPECIFIED) | | USE DURING A RANDOM TEST ADMINISTERED ON 11/3/98. THE SAMPLE WAS CONFIRMED | | POSITIVE ON 11/6/98. THE NRC RESIDENT INSPECTOR HAS BEEN INFORMED BY THE | | LICENSEE. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35008 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: COOK REGION: 3 |NOTIFICATION DATE: 11/06/1998| | UNIT: [] [2] [] STATE: MI |NOTIFICATION TIME: 19:00[EST]| | RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 07/01/1995| +------------------------------------------------+EVENT TIME: [EST]| | NRC NOTIFIED BY: MARY BETH DePUYDT |LAST UPDATE DATE: 11/06/1998| | HQ OPS OFFICER: BOB STRANSKY +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |RONALD GARDNER R3 | |10 CFR SECTION: | | |ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ | | | |2 N N 0 Cold Shutdown |0 Cold Shutdown | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | HISTORICAL REPORT REGARDING POTENTIAL FOR LOSS OF FEEDWATER DUE TO HELB | | | | "It has been determined that in July 1995, the fire door/High Energy Line | | Break (HELB) door for the startup blowdown flash tank room had been blocked | | open for approximately 39 hours to allow for draining of the ESW header. At | | the time the condition was evaluated, it was concluded that it was | | permissible to block the door open for less than 72 hours based on the | | assumption that any HELB would be detected and stopped within 10 minutes and | | there would be no automatic actions, such as a reactor trip. By engineering | | evaluation, it was also concluded that the Motor Control Centers (MCCs) in | | the hallway outside the room would not be adversely affected in the 10 | | minutes that the leak existed. As part of the recent Auxiliary Feedwater | | (AFW) Safety System Functional Inspection (SSFI), the assumption that the | | leak would be terminated in 10 minutes was revisited and determined not to | | be valid. In parallel, a review of Westinghouse WCAP 10961 revealed that | | automatic action - a reactor trip - would occur at 12 minutes into the | | leak. | | | | "Today, based on this new information, additional review determined that a | | HELB in the startup blowdown flash tank room would potentially expose the | | MCCs to higher temperatures and higher humidity than they are qualified for. | | If the components on these MCCs were inoperable due to this, the potential | | exists for a loss of all feedwater. AFW would potentially be lost as the | | valves which supply Essential Service Water to the AFW pumps are on these | | MCCs, and main feedwater would have been lost on the reactor trip. This | | would represent an unanalyzed condition. | | | | "Both units are currently in Mode 5, cold shutdown. This new information | | will be evaluated and appropriate actions taken prior to startup." | | | | THE NRC RESIDENT INSPECTOR HAS BEEN INFORMED OF THIS REPORT BY THE LICENSEE. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Fuel Cycle Facility |Event Number: 35009 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 11/07/1998| | RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 22:00[EST]| | COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 11/06/1998| | 6903 ROCKLEDGE DRIVE |EVENT TIME: 23:00[EST]| | BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 11/07/1998| | CITY: PIKETON REGION: 3 +-----------------------------+ | COUNTY: PIKE STATE: OH |PERSON ORGANIZATION | |LICENSE#: GDP-2 AGREEMENT: N |RONALD GARDNER R3 | | DOCKET: 0707002 |MIKE BELL NMSS | +------------------------------------------------+ | | NRC NOTIFIED BY: KURT SISLER | | | HQ OPS OFFICER: BOB STRANSKY | | +------------------------------------------------+ | |EMERGENCY CLASS: N/A | | |10 CFR SECTION: | | |NCFR NON CFR REPORT REQMNT | | | | | | | | | | | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | 24-HOUR REPORT DUE TO ALARM OF SMOKE DETECTORS | | | | AT 2316 ON 11/6/98, TWO SMOKEHEADS IN THE TAILS WITHDRAWAL AREA OF THE | | X-330 PROCESS BUILDING ACTUATED. THE 30WB1 TAILS WITHDRAWAL COMPRESSOR WAS | | IMMEDIATELY ISOLATED AND VENTED TO SUBATMOSPHERIC PRESSURE, AND THE | | SMOKEHEADS WERE RESET. PERSONNEL RESPONDING TO THE AREA OBSERVED NO SMOKE, | | AND A PRELIMINARY INVESTIGATION BY HEALTH PHYSICS PERSONNEL FOUND NO | | EVIDENCE OF OUTGASSING. FURTHER INVESTIGATION IS ONGOING TO DETERMINE THE | | CAUSE OF THE ACTUATION. | | | | THIS EVENT WAS REPORTED AS AN ACTUATION OF A SAFETY SYSTEM (CADP SMOKEHEAD | | ALARM) DUE TO A POTENTIAL OUTGASSING. THE NRC RESIDENT INSPECTOR HAS BEEN | | INFORMED. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35010 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: PALO VERDE REGION: 4 |NOTIFICATION DATE: 11/08/1998| | UNIT: [] [] [3] STATE: AZ |NOTIFICATION TIME: 10:00[MST]| | RXTYPE: [1] CE,[2] CE,[3] CE |EVENT DATE: 11/08/1998| +------------------------------------------------+EVENT TIME: 08:00[MST]| | NRC NOTIFIED BY: STEVE BANKS |LAST UPDATE DATE: 11/08/1998| | HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |CHARLES CAIN R4 | |10 CFR SECTION: | | |DDDD 73.71 UNSPECIFIED PARAGRAPH | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ | | | | | | |3 N N 0 Hot Standby |0 Hot Standby | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | - PHYSICAL SECURITY EVENT - | | | | SECURITY WEAPON UNATTENDED FOR APPROXIMATELY 25 MINUTES. IMMEDIATE | | COMPENSATORY MEASURES TAKEN UPON DISCOVERY. THE LICENSEE INFORMED THE NRC | | RESIDENT INSPECTOR. REFER TO THE HOO LOG FOR ADDITIONAL DETAILS. | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ |Power Reactor |Event Number: 35011 | +------------------------------------------------------------------------------+ +------------------------------------------------------------------------------+ | FACILITY: PRAIRIE ISLAND REGION: 3 |NOTIFICATION DATE: 11/09/1998| | UNIT: [] [2] [] STATE: MN |NOTIFICATION TIME: 02:00[CST]| | RXTYPE: [1] W-2-LP,[2] W-2-LP |EVENT DATE: 11/09/1998| +------------------------------------------------+EVENT TIME: 00:00[CST]| | NRC NOTIFIED BY: DOUG SMITH |LAST UPDATE DATE: 11/09/1998| | HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+ +------------------------------------------------+PERSON ORGANIZATION | |EMERGENCY CLASS: N/A |RONALD GARDNER R3 | |10 CFR SECTION: | | |ARPS 50.72(b)(2)(ii) RPS ACTUATION | | | | | | | | | | | +-----+----------+-------+--------+-----------------+--------+-----------------+ |UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE | +-----+----------+-------+--------+-----------------+--------+-----------------+ | | | |2 A/R Y 22 Power Operation |0 Hot Standby | | | | +------------------------------------------------------------------------------+ EVENT TEXT +------------------------------------------------------------------------------+ | AUTO REACTOR TRIP FROM 22% POWER DUE TO MAIN TURBINE TRIP WHILE SHUTTING | | DOWN | | | | DURING A PLANNED PLANT SHUTDOWN FOR REFUELING, THE UNIT 2 MAIN TURBINE | | TRIPPED FROM 22% POWER FOR UNKNOWN REASONS. THIS CAUSED THE REACTOR TO AUTO | | TRIP. ALL CONTROL ROD BOTTOM LIGHTS INDICATED THAT ALL CONTROL RODS HAD | | COMPLETELY INSERTED INTO THE CORE. HOWEVER, THE INDIVIDUAL ROD POSITION | | INDICATORS (IRPIs) INDICATED ONE CONTROL ROD AT 10 STEPS AND A SECOND | | CONTROL ROD AT 19 STEPS. THE STEAM GENERATORS ARE BEING FED BY THE MAIN | | FEEDWATER SYSTEM AND STEAM IS BEING DUMPED TO THE MAIN CONDENSER. NO SAFETY | | OR RELIEF VALVES LIFTED. UNIT 2 IS STABLE IN MODE 3 (HOT STANDBY). | | | | THE LICENSEE IS INVESTIGATING THE CAUSE OF THE MAIN TURBINE TRIP AND THE | | IRPI CONDITION AND WILL PROCEED WITH THE REFUELING OUTAGE. | | | | THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. | +------------------------------------------------------------------------------+