Event Notification Report for August 2, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
07/30/1999 - 08/02/1999
** EVENT NUMBERS **
35790 35890 35958 35973 35974 35975 35976 35977 35978 35979
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35790 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 06/03/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 14:44[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 06/02/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 16:30[CDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 07/31/1999|
| CITY: PADUCAH REGION: 3 +-----------------------------+
| COUNTY: McCRACKEN STATE: KY |PERSON ORGANIZATION |
|LICENSE#: GDP-1 AGREEMENT: Y |DAVID HILLS R3 |
| DOCKET: 0707001 |DON COOL NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: THOMAS WHITE | |
| HQ OPS OFFICER: FANGIE JONES | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|OCBA 76.120(c)(2)(i) ACCID MT EQUIP FAILS | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| THREE SPRINKLER SYSTEMS DECLARED INOPERABLE DUE TO CORRODED HEADS (24-hour |
| report) |
| |
| The following text is a portion of a facsimile received from Paducah: |
| |
| "On 06/02/99 at 1630 CDT, the Plant Shift Superintendent (PSS) was notified |
| that numerous sprinkler heads were corroded, affecting 16 sprinkler systems |
| in C-337 and one system in C-333, such that the ability of the sprinklers to |
| flow sufficient water was called into question. Subsequently, these |
| sprinkler systems were declared inoperable, and TSR-required actions |
| establishing roving fire patrols were initiated. This deficiency was |
| detected during scheduled system inspections conducted by Fire Protection |
| personnel. Currently, functionality of the sprinkler heads has not been |
| fully evaluated by Fire Protection personnel, and the remaining cascade |
| buildings are currently being inspected, and if necessary, this report will |
| be updated to identify any additional areas. |
| |
| "It has been determined that this event is reportable under |
| 10CFR76.120(c)(2) as an event in which equipment is disabled or fails to |
| function as designed." |
| |
| The NRC resident inspector has been notified of this event. |
| |
| * * * UPDATE AT 1022 ON 06/04/99 FROM CAGE TO TROCINE * * * |
| |
| The following text is a portion of a facsimile received from Paducah: |
| |
| "Two sprinkle heads on system D-1 in C-337 and two sprinkler heads on system |
| 27 in C-335 were identified to also be corroded. These were identified to |
| the PSS on 06/03/99 at 1600 CDT and 1601 CDT, respectively, and determined |
| to require an update to this report by the PSS. |
| |
| "It has been determined that this event is reportable under |
| 10CFR76.120(c)(2) as an event in which equipment is disabled or fails to |
| function as designed." |
| |
| Paducah personnel notified the NRC resident inspector of this update. The |
| NRC operations officer notified the R3DO (Hills) and NMSS EO (Combs). |
| |
| * * * UPDATE AT 2152 ON 06/17/99 FROM WALKER TO POERTNER * * * |
| |
| The following text is a portion of a facsimile received from Paducah: |
| |
| "Three sprinkler heads on system C-15 and five sprinkler heads on system B-8 |
| in C-333 were identified to also be corroded. The PSS was notified of this |
| condition at 1300 CDT on 06/17/99 and determined that an update to this |
| report was required." |
| |
| "It has been determined that this event is reportable under |
| 10CFR76.120(c)(2) as an event in which equipment is disabled or fails to |
| function as designed." |
| |
| Paducah personnel notified the NRC resident inspector of this update. The |
| NRC operations officer notified the R3DO (Madera). |
| |
| * * * UPDATE 1440 6/18/1999 FROM UNDERWOOD TAKEN BY STRANSKY * * * |
| |
| "Two sprinkler heads on system C-15 and one sprinkler head on system B-8 in |
| C-333 were identified to have corrosion. The PSS was notified of this |
| condition at 1350 CDT on 06/18/99. The area of the fire patrol for system |
| C-15 was expanded to include the two heads identified as corroded. The one |
| head on system B-8 was in the area already being patrolled. The PSS |
| determined that an update to this report was required." |
| |
| The NRC resident inspector has been informed of this update. Notified R3DO |
| (Madera). |
| |
| * * * UPDATE 1315 6/25/1999 FROM WALKER TAKEN BY STRANSKY * * * |
| |
| "Two sprinkler heads on system D-8 and three sprinkler heads on system D-7 |
| in C-337 were identified to have corrosion. The PSS was notified of the |
| condition on system D-8 at 0125 CDT on 06/25/99 and at 1019 CDT on 06/25/99 |
| for system D-7. Both systems were immediately declared inoperable and LCO |
| fire patrol actions were implemented. It was determined that an update to |
| this report was required." |
| |
| The NRC resident inspector has been informed of this update. Notified R3DO |
| (Jordan). |
| |
| * * * UPDATE 2119 7/30/1999 FROM CAGE TAKEN BY STRANSKY * * * |
| |
| "Five sprinkler heads and one sprinkler piping tee on C-337 system D-7 were |
| identified to have corrosion. The PSS was notified of these corroded system |
| parts and declared the system inoperable at 0931 CDT on 07/30/99. LCO |
| required fire patrols of the affected area were initiated. The PSS |
| determined that an update to this event report was required." |
| |
| The NRC resident inspector has been informed of this update. Notified R3DO |
| (Wright). |
| |
| * * * UPDATE 1655 7/31/1999 FROM WHITE TAKEN BY STRANSKY * * * |
| |
| "Two sprinkler heads on C-337 System D-1 were identified to have corrosion. |
| The PSS was notified of these corroded system parts and declared the system |
| inoperable at 1155 on 7/31/99. LCO required fire patrols of the affected |
| area were initiated. The PSS determined that an update to this report was |
| required." |
| |
| The NRC resident inspector has been informed of this update. Notified R3DO |
| (Wright). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35890 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: VERMONT YANKEE REGION: 1 |NOTIFICATION DATE: 07/02/1999|
| UNIT: [1] [] [] STATE: VT |NOTIFICATION TIME: 16:58[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 07/02/1999|
+------------------------------------------------+EVENT TIME: 16:34[EDT]|
| NRC NOTIFIED BY: MIKE EMPY |LAST UPDATE DATE: 07/30/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |LAWRENCE DOERFLEIN R1 |
|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
+------------------------------------------------------------------------------+
| - PRIMARY CONTAINMENT COULD BE OUTSIDE ITS DESIGN BASIS UNDER CERTAIN |
| CONDITIONS - |
| |
| During validation of the Vermont Yankee (VY) Containment Pressurization |
| System Design Basis Document, the licensee determined that a design analysis |
| which envelopes the design operating conditions of the torus and drywell |
| does not exist. Specifically, no design analysis exists which verifies the |
| ability of the torus-to-drywell and torus-to-reactor building vacuum |
| breakers to limit depressurization of the containment to less than the |
| design basis value of - 2 psig (22A1265, Rev. 1). |
| |
| The specific event in question involves the effects of an inadvertent |
| drywell spray actuation occurring during conditions when the torus water is |
| at a minimum temperature 50�F (VYAPF 0150.03). |
| |
| Present design evaluations (VYC-236, Rev 0, "Torus-Reactor Building Vacuum |
| Breaker Conditions" and VYC-315, Rev 0, "Primary Containment Vacuum") |
| calculated the resulting torus and containment pressure caused by |
| inadvertent spray actuation, but used a spray water temperature of 83.7�F. |
| These analyses concluded that vacuum breaker operation was not necessary to |
| ensure that the containment remained within its external design pressure. |
| These analyses evaluated this event at normal operating conditions (100�F |
| torus water temperature, 33�F service water temperature, and 165�F drywell |
| temperature). No evaluation has been performed for temperatures below these |
| values. |
| |
| The General Electric design basis for the VY vacuum breaker sizing is based |
| on an evaluation for Monticello. This evaluation assumed a minimum spray |
| water temperature of 50�F and assumed that all vacuum breakers operated |
| within one second. The VY vacuum breaker design differs from this design |
| assumption in that operation of the torus-to-reactor building vacuum |
| breakers requires the opening of air operated valves (AOVs) #SB-16-19-11A & |
| B as part of the vacuum breaker operation. These AOVs require more than 5 |
| seconds to operate (VYOPF 4115.01, 03/23/99). As a result, it cannot be |
| assured that the vacuum breaker system will function adequately to prevent |
| the containment from exceeding its design basis external design pressure for |
| low spray water temperature conditions. |
| |
| A plant operability evaluation, based on the information included in |
| VYC-315, Rev 0, has concluded that a minimum drywell spray water temperature |
| of 70�F would be required to approach the design basis containment external |
| pressure limit of - 2.0 psig without effective vacuum breaker operation. In |
| order to achieve this low spray water temperature, a combination of low |
| torus water temperature and/or low service water temperature would be |
| required to exist. Current operating conditions indicate that the torus |
| water temperature is being maintained at ~80�F and has been maintained at |
| this temperature during the months of May and June, 1999. The current |
| service water temperature of 79�F ensures that, in the event of an |
| inadvertent drywell spray event, spray water temperature will not be lower |
| than 70�F. |
| |
| This 79�F service water temperature corresponds to the maximum 20 year |
| average for river water temperature. Based on this temperature, river water |
| temperature would not be anticipated to decrease below 70�F until |
| mid-to-late September, 1999. |
| |
| Based on the current high service water temperature, in combination with |
| existing torus water temperature, this condition does not effect the |
| operability of the primary containment or the operability of the primary |
| containment vacuum breakers. There is no operability concern providing the |
| torus water temperature remains above 70�F and river water temperature |
| remains above 33�F. |
| |
| The license plans to immediately issue standing orders to plant operators |
| regarding this situation and to perform necessary design analyses prior to |
| September, 1999. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| * * * UPDATE AT 0956 ON 07/30/99 BY SORTWELL TO WEAVER * * * |
| |
| The licensee is retracting this event report based on the following |
| explanation: |
| |
| GE Design Specification #22A2753 sizes the torus-to-reactor building vacuum |
| breakers to cope with an inadvertent containment spray initiation. This |
| specification is not part of VY's current licensing basis. Rather, this |
| specification provides guidance to be applied when determining the size of |
| the subject vacuum breakers. The postulated scenario (the inadvertent spray |
| initiation) requires multiple operator errors. The VY plant design basis |
| requires that the licensee postulate any SINGLE failure, including single |
| operator errors. Scenarios that assume multiple operator errors are beyond |
| the design basis of the VY plant. |
| |
| At the time of the original ENS notification, VY had in place, analyses |
| supporting containment spray operations. Those analyses demonstrated that |
| the actuation of containment sprays, consistent with VY plant procedures, |
| would have effects consistent with the design basis of the plant. |
| Additionally, at the time of discovery, VY had in place, calculations that |
| bounded the inadvertent spray scenario under the plant conditions present. |
| |
| More recently, an analysis was performed to quantify the possible effect of |
| an inadvertent initiation of containment spray under the off-normal |
| conditions identified in the GE design specification, including the assumed |
| multiple operator errors. The analysis assumes initial plant conditions |
| that are conservative. |
| |
| Using this approach, it was determined that an inadvertent initiation of |
| containment sprays could, if unmitigated, achieve the drywell design |
| pressure differential of 2 psid (drywell external pressure greater than |
| internal pressure) approximately 20 seconds after initiation. |
| |
| The subject vacuum breakers would be fully open approximately 13 seconds |
| into the postulated event. Either one of the two sets of vacuum breakers |
| has adequate capacity to limit the pressure transient to less than 2 psid. |
| |
| Therefore, this event is being retracted. |
| |
| The licensee notified the NRC resident inspector and the NRC operations |
| officer notified the R1DO (Kinneman). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35958 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: LASALLE REGION: 3 |NOTIFICATION DATE: 07/25/1999|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 22:32[EDT]|
| RXTYPE: [1] GE-5,[2] GE-5 |EVENT DATE: 07/25/1999|
+------------------------------------------------+EVENT TIME: 20:21[CDT]|
| NRC NOTIFIED BY: WILLIAMS |LAST UPDATE DATE: 07/30/1999|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |PATRICK HILAND R3 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(1)(ii)(A) UNANALYZED COND OP | |
|NLCO TECH SPEC LCO A/S | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 87 Power Operation |87 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PLANT ENTERED TECH SPEC 3.2.3 LCO ACTION STATEMENT |
| |
| AT 2021 CDT ON 07/25/99, THE "1B21-N500" PRESSURE TRANSMITTER INDICATION |
| DROPPED FROM 1000 PSIG TO INDICATED 600 PSIG REACTOR PRESSURE. THIS |
| TRANSMITTER IS THE PRIMARY PRESSURE TRANSMITTER TO THE ELECTROHYDRAULIC |
| CONTROL (EHC) SYSTEM. DUE TO THIS CHANGE IN PRESSURE INDICATION, THE BACKUP |
| TRANSMITTER TOOK CONTROL. IN THIS LINEUP (BACKUP TRANSMITTER IN CONTROL OF |
| THE EHC SYSTEM), THERE IS NO OTHER BACKUP TRANSMITTER AVAILABLE AND |
| PROCEDURE "LOA-EH-101" STATES THAT THIS CONDITION IS AN UNANALYZED CONDITION |
| AND THE PLANT IS TO ENTER TS 3.2.3. TS 3.2.3 REQUIRES THE REACTOR TO BE |
| <25% POWER WITHIN 4 HOURS. THIS CONDITION WAS NOT DETERMINED TO BE AN |
| UNANALYZED CONDITION UNTIL 2045 CDT ON 07/25/99 UTILIZING THE ABOVE |
| PROCEDURE AND THE UFSAR. THE PROBLEM WITH THE TRANSMITTER IS BEING |
| INVESTIGATED TO DETERMINE WHY IT DOES NOT INDICATE 1000 PSIG. |
| |
| THE LICENSEE WILL INFORM THE NRC RESIDENT INSPECTOR. |
| |
| * * * RETRACTION AT 2253 ON 07/29/99 FROM GRANWALD TO STRANSKY * * * |
| |
| "It was subsequently determined from a detailed evaluation that was |
| performed in May, 1999, which clearly shows that operation with a pressure |
| regulator out of service at LaSalle is bounded by the thermal limits |
| calculated for the slow closure of one or more turbine control valves |
| (TCVs). The use of the thermal limits reported in the LaSalle Unit 1 Cycle |
| 8 Core Operating Limits Report for the slow closure of one or more TCVs for |
| operations with a pressure regulator out of service does not result in |
| operation of the plant in an unanalyzed condition. The thermal limits were |
| adjusted to be in line with the TCV slow closure analysis and plant thermal |
| limits were declared operable and Tech Spec 3.2.3 exited within 4 hours. |
| Therefore, LaSalle Unit 1 was not in an unanalyzed condition as reported in |
| Event No. 35958." |
| |
| The licensee will notify the NRC resident inspector and the NRC Operations |
| Officer notified the R3DO Wright. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35973 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NINE MILE POINT REGION: 1 |NOTIFICATION DATE: 07/30/1999|
| UNIT: [] [2] [] STATE: NY |NOTIFICATION TIME: 00:02[EDT]|
| RXTYPE: [1] GE-2,[2] GE-5 |EVENT DATE: 07/29/1999|
+------------------------------------------------+EVENT TIME: 23:16[EDT]|
| NRC NOTIFIED BY: ANTHONY PETRELLI |LAST UPDATE DATE: 07/30/1999|
| HQ OPS OFFICER: DOUG WEAVER +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JOHN KINNEMAN R1 |
|10 CFR SECTION: | |
|AINA 50.72(b)(2)(iii)(A) POT UNABLE TO SAFE SD | |
|NLCO TECH SPEC LCO A/S | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| HIGH PRESSURE CORE SPRAY SYSTEM INOPERABLE |
| |
| Check valve #2CSH*V16 on the High Pressure Core Spray (HPCS) System pump |
| suction from the suppression pool is not in the inservice testing program |
| plan for reverse flow testing. A preliminary review indicates that this |
| check valve should be reverse-flow tested. The HPCS System was declared |
| inoperable and Unit 2 entered a 14 day LCO. Steps are being taken to retest |
| the valve. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35974 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DRESDEN REGION: 3 |NOTIFICATION DATE: 07/30/1999|
| UNIT: [] [2] [3] STATE: IL |NOTIFICATION TIME: 14:01[EDT]|
| RXTYPE: [1] GE-1,[2] GE-3,[3] GE-3 |EVENT DATE: 07/30/1999|
+------------------------------------------------+EVENT TIME: 11:50[CDT]|
| NRC NOTIFIED BY: BRIAN SAMPSON |LAST UPDATE DATE: 07/30/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |GEOFFREY WRIGHT R3 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 89 Power Operation |89 Power Operation |
|3 N Y 89 Power Operation |89 Power Operation |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| OFFSITE NOTIFICATION TO ILLINOIS ENVIRONMENTAL PROTECTION AGENCY. |
| |
| THE LICENSEE NOTIFIED THE ILLINOIS ENVIRONMENTAL PROTECTION AGENCY THAT ON |
| 07/29/99, STATION COOLING WATER EFFLUENT EXCEEDED DISCHARGE EFFLUENT |
| TEMPERATURE LIMITATIONS. |
| |
| THE LICENSEE WILL INFORM THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35975 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PILGRIM REGION: 1 |NOTIFICATION DATE: 07/30/1999|
| UNIT: [1] [] [] STATE: MA |NOTIFICATION TIME: 20:24[EDT]|
| RXTYPE: [1] GE-3 |EVENT DATE: 07/30/1999|
+------------------------------------------------+EVENT TIME: 19:30[EDT]|
| NRC NOTIFIED BY: DAVE NOYES |LAST UPDATE DATE: 07/30/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JOHN KINNEMAN R1 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(1)(ii)(A) UNANALYZED COND OP | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| SALT SERVICE WATER SYSTEM EXCEEDED MAXIMUM DESIGN TEMPERATURE |
| |
| The 'A' train of the salt service water system was declared inoperable when |
| the intake structure temperature exceeded the maximum design temperature of |
| 105�F. The highest actual temperature reached was 106�F, which occurred for |
| approximately 15 minutes, until operators were able to reduce the |
| temperature by realigning the salt service water system. The current intake |
| structure temperature is 100�F. |
| |
| The NRC resident inspector has been informed of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35976 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BRUNSWICK REGION: 2 |NOTIFICATION DATE: 07/31/1999|
| UNIT: [1] [2] [] STATE: NC |NOTIFICATION TIME: 15:22[EDT]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 07/31/1999|
+------------------------------------------------+EVENT TIME: 11:41[EDT]|
| NRC NOTIFIED BY: J. REINSBURROW |LAST UPDATE DATE: 07/31/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| CONTROL BUILDING VENTILATION ISOLATION |
| |
| "On 7/31/99 at 11:41 during change out of the Brunswick site chlorine tank |
| car an isolation of the control building ventilation system occurred. Two |
| chlorine detectors actuated at the Service Water building adjacent to the |
| chlorine tank car. These detectors isolate the control building ventilation |
| system on detection of chlorine. The control building ventilation system |
| components functioned as designed. |
| |
| "The evolution in progress during the isolation was the disconnection of the |
| spent chlorine tank car. Personnel in the area of the tank car with portable |
| chlorine detection equipment did not detect the presence of chlorine gas. |
| Subsequent inspections of areas adjacent to the tank car did not identify |
| chlorine gas with portable monitors." |
| |
| The chlorine detectors were reset, and the ventilation system was restored |
| to its normal lineup. The NRC resident inspector has been informed of this |
| event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35977 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NINE MILE POINT REGION: 1 |NOTIFICATION DATE: 08/01/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 15:43[EDT]|
| RXTYPE: [1] GE-2,[2] GE-5 |EVENT DATE: 08/01/1999|
+------------------------------------------------+EVENT TIME: 14:20[EDT]|
| NRC NOTIFIED BY: ROBERT KIRCHNER |LAST UPDATE DATE: 08/01/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JOHN KINNEMAN R1 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 0 Startup |0 Cold Shutdown |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUTOMATIC REACTOR SCRAM DURING STARTUP |
| |
| An automatic reactor scram occurred due to spurious level spikes of |
| intermediate range monitor (IRM) neutron detector channels 12, 15 and 16. At |
| the time of the event, the unit was critical, but just below the point of |
| adding heat. All control rods inserted following the scram. |
| |
| The licensee reported that the spurious IRM spikes occurred when the |
| selector switch for IRM channel 11 was rotated from range 2 to range 3. The |
| licensee is currently investigating the cause of this event. The NRC |
| resident inspector has been informed of this notification by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35978 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HATCH REGION: 2 |NOTIFICATION DATE: 08/01/1999|
| UNIT: [1] [2] [] STATE: GA |NOTIFICATION TIME: 23:38[EDT]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 08/01/1999|
+------------------------------------------------+EVENT TIME: 23:00[EDT]|
| NRC NOTIFIED BY: AL DEES |LAST UPDATE DATE: 08/01/1999|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG R2 |
|10 CFR SECTION: | |
|AENS 50.72(b)(1)(v) ENS INOPERABLE | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 98 Power Operation |98 Power Operation |
|2 N Y 98 Power Operation |98 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| LOSS OF NOAA RADIO |
| |
| NOAA radio communications were lost for 22 minutes due to a problem offsite. |
| The radio has been restored to operation. |
| |
| The licensee has notified the state and local government agencies and will |
| notify the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35979 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: WATERFORD REGION: 4 |NOTIFICATION DATE: 08/01/1999|
| UNIT: [3] [] [] STATE: LA |NOTIFICATION TIME: 23:45[EDT]|
| RXTYPE: [3] CE |EVENT DATE: 08/01/1999|
+------------------------------------------------+EVENT TIME: 21:49[CDT]|
| NRC NOTIFIED BY: DAVID LITOLFF |LAST UPDATE DATE: 08/01/1999|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BILL JONES R4 |
|10 CFR SECTION: | |
|ASHU 50.72(b)(1)(i)(A) PLANT S/D REQD BY TS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3 M/R Y 100 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| MANUAL REACTOR TRIP DUE TO LOSS OF CONTROLLED BLEEDOFF FLOW TO RCP 2B |
| |
| The licensee performed a manual reactor trip due to the loss of Reactor |
| Coolant Pump (RCP) 2B seal controlled bleedoff flow. The loss of seal |
| coolant flow resulted in a high seal temperature which requires tripping the |
| reactor by procedure. RCP 2B was secured immediately following the reactor |
| trip. The plant is in Hot Standby and stable. The loss of seal controlled |
| bleedoff flow is under investigation. |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021
Page Last Reviewed/Updated Thursday, March 25, 2021