Event Notification Report for May 24, 2000
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/23/2000 - 05/24/2000
** EVENT NUMBERS **
36958 36980 37015 37018 37019 37020 37021 37022 37023
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36958 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: KEWAUNEE REGION: 3 |NOTIFICATION DATE: 05/02/2000|
| UNIT: [1] [] [] STATE: WI |NOTIFICATION TIME: 17:10[EDT]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 05/02/2000|
+------------------------------------------------+EVENT TIME: 07:40[CDT]|
| NRC NOTIFIED BY: GARY HARRINGTON |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BRUCE BURGESS R3 |
|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 N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - 'A' TRAIN EMERGENCY SAFEGUARDS BUS UNEXPECTEDLY DEENERGIZED DURING |
| MAINTENANCE - |
| |
| At 0740 CDT on 05/02/00, while electrical bus maintenance was in progress, |
| the 'A' train emergency safeguards bus unexpectedly deenergized. |
| DEENERGIZING the bus initiated an ESF start signal for the associated 'A' |
| emergency diesel generator (EDG). At the time, the 'A' EDG had been removed |
| from service for refueling outage scheduled maintenance and no EDG start |
| occurred. |
| |
| In response to the loss of power to the 'A' train safeguards bus, the |
| licensee manually started the 'B' train residual heat removal pump to |
| reestablish decay heat removal There was no temperature rise in the |
| primary system. |
| |
| The licensee is determining the cause of the bus deenergization. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| * * * UPDATE ON 5/23/00 @ 1216 BY HARRINGTON TO GOULD * * * RETRACTION |
| |
| "For the event that occurred on May 2, 2000, there was no actual ESF |
| equipment (pumps, valves, etc.) actuated directly as a consequence of the |
| event. Therefore, no ESF equipment operated to mitigate the event. As such, |
| the event would not be reportable. However, there are implications that the |
| reporting requirements apply equally to ESF signals that are generated as |
| part of an event, regardless of whether the event caused equipment to |
| operate or not. For instance, NUREG-1022 contains a paragraph that states, |
| in part, that, " [t]he Statement of Considerations also indicates that |
| "actuation" of multichannel ESF actuation systems is defined as actuation |
| of enough channels to complete the actuation logic." Accordingly, the May 2 |
| event was evaluated in greater detail considering that the loss of power to |
| the bus in itself could be understood to be an ESF actuation. |
| |
| "At the time of the event, work on the emergency bus relays was in process. |
| During the work the bus was unexpectedly de-energized when the breaker |
| providing power to the bus from the off-site power source opened. From our |
| investigation, it appears a relay which was not being directly worked on |
| actuated. Since power was removed from the relay, the relay appears to have |
| been bumped or jarred which manually actuated it. The relay that was |
| actuated was a trip relay for the breakers providing power to the affected |
| bus. |
| |
| "The ESF function that could be interpreted to be actuated as a result of |
| the relay being actuated is the loss of power to the safeguards bus start |
| signal to the associated diesel. However, the diesel was tagged out of |
| service and the bus voltage restoring control circuit was defeated as part |
| of the bus work that was in progress. |
| |
| "In order to defeat the voltage restoring circuit the bus voltage restoring |
| control switch was placed in "manual," and the voltage restoring relays were |
| de-energized. With the switch in manual, a voltage search signal is not |
| generated. As a result, the diesel does not receive a start signal if a loss |
| of power to the bus occurs. Additionally, as part of the voltage restoring |
| circuit, once power is lost to the bus, a power search is initiated whereby |
| the circuit electrically seeks an available off-site power source and then |
| would seek the diesel if no off-site source were available. With the bus |
| voltage restoring circuit in manual there was no power source search |
| initiated. The bus power was not automatically restored, even though the |
| power to the bus was available. |
| |
| "Included in NUREG-1022 are a number of examples of situations where NRC has |
| described reportable events. Of those that are described, all either |
| involved equipment (pumps, valves, etc.) that actuated or the condition that |
| generated a signal needed the ESF function to mitigate the event whether |
| equipment actuated or not. In the event reported on May 2, no equipment |
| operated and there was no reliance on any accident mitigation function as |
| well as no need for any accident mitigation feature. Therefore, the event |
| should not have been reported as an ESF actuation simply because the |
| condition that occurred could have resulted in an ESF signal being |
| actuated. |
| |
| "According to the reporting criteria, if the actuation is invalid, and the |
| system was properly removed from service a report need not be filed. |
| According to NUREG-1022, "[v]alid ESF actuations are those that result from |
| "valid signals" or from intentional manual initiation, unless it is part of |
| a preplanned test. Valid signals are those signals that are initiated in |
| response to actual plant conditions or parameters satisfying the |
| requirements for ESF initiation. Note this definition of "valid" requires |
| that the initiation signal must be an ESF signal. This distinction |
| eliminates actuations which are the result of non-ESF signals from the class |
| of valid actuations. Invalid actuations are, by definition, those that do |
| not meet the criteria for being valid. Thus invalid actuations include |
| actuations that are not the result of valid signals and are not intentional |
| manual actuations." |
| |
| "The ESF signal of concern for the start of the diesel generator is that |
| which is generated in response to a loss of off-site power to the affected |
| bus. During the subject event, off-site power was not lost. Although the |
| off-site power was not automatically restored according to normal system |
| operational design, it remained available. Consequently, there was no need |
| for the diesel to supply the bus and as such no valid signal was generated. |
| Additionally, the power restoration circuit was properly removed from |
| service during the event; the voltage restoring switch was in manual. |
| Therefore, no ESF signal was generated. |
| |
| "In summary, the event described is not reportable based on 1) there not |
| being a need for any ESF feature to mitigate the event, and 2) the event not |
| causing a valid (or any) ESF signal along with the related ESF equipment |
| being properly removed from service." |
| |
| |
| The NRC Resident Inspector was notified. Reg 3 RDO (Hiland) was |
| informed. |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36980 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: KEWAUNEE REGION: 3 |NOTIFICATION DATE: 05/07/2000|
| UNIT: [1] [] [] STATE: WI |NOTIFICATION TIME: 13:53[EDT]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 05/07/2000|
+------------------------------------------------+EVENT TIME: 11:25[CDT]|
| NRC NOTIFIED BY: TERRY GENCIUS |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BRUCE BURGESS R3 |
|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 N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNEXPECTED LOAD SHED OCCURRED DURING ELECTRICAL MAINTENANCE TESTING |
| |
| "On May 7, 2000 at 0933, electrical testing was being performed on the |
| emergency diesel generator A load sequencer. The diesel generator was |
| out-of-service at the time [for annual maintenance]. An unexpected load |
| shed signal was developed, which removed charging pump C, service water pump |
| A2, containment fan coil A, and residual heat removal (RHR) pump A from the |
| train A emergency safeguards bus (Bus 5). In response to the load shed, |
| equipment was manually restarted from the control room. RHR pump A was |
| restarted in 90 seconds, restoring shutdown cooling. The cause of the load |
| shed was the post modification testing of the load sequencer, and is |
| continuing to be investigated." |
| |
| There was no increase in reactor coolant temperature during the 90 seconds |
| without RHR flow. Also, the RHR B pump was available if it had been |
| needed. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| * * * UPDATE ON 5/23/00 @ 1216 BY HARRINGTON TO GOULD * * * RETRACTION |
| |
| After further review, appears this event is NOT reportable based upon the |
| initiation signal was invalid and the system was properly removed from |
| service. Also the Load Shed Signal is a component of the Diesel Generator |
| engineered safety feature (ESF) and not the actuation of the ESF train. |
| |
| "NUREG 1022, "Event Reporting Guidelines: 10 CFR 50.72 and 50.73" Revision I |
| (NUREG 1022) provides guidance on what the NRC wants the industry to report. |
| Under section 3.3.2, "Actuation of an Engineered Safety Feature or the RPS" |
| is a definition of "valid signals." Valid signals are those signals that |
| are initiated in response to actual plant conditions or parameters |
| satisfying the requirements for ESF initiation. The actual plant condition |
| that actuates this ESF is an undervoltage condition on Bus 5, this did not |
| occur therefore it was an invalid signal. |
| |
| "During the test for DCR 3002 initial conditions were in place to prevent |
| the ESF from providing its intended feature. Preventive Maintenance |
| Procedure (PMP) 42-14, "DGE-Train "A" Auto Sequencing Test with Diesel A in |
| Pullout (Degraded Grid)" disables the ESF by placing the control switch (ES |
| 46641) for Emergency Diesel Generator A in "PULLOUT." To start the test the |
| control room operator places the Bus 5 Voltage Restoring Logic Test switch |
| to the "TEST" position, this disables the rest of the ESF from actuating |
| unless a valid signal is generated. |
| |
| "By looking at the actions required to mitigate the consequences of a |
| significant event, the load shed is only one component of the safety |
| feature. The shedding of loads from bus 5 does not, of itself, mitigate the |
| consequences of any significant event. |
| |
| "In addition, during the test sequence the load shed relays are in a state |
| which, if not blocked, would cause the same equipment to trip. This event |
| occurred because the block was removed prior to the load shed relays |
| resetting to their normal state. |
| |
| "Therefore, one component, Load Shed, of the Diesel Generator ESF actuated |
| due to an invalid signal while the ESF was properly removed from service. |
| Thus the event is not reportable." |
| |
| The NRC Resident Inspector was notified. Reg 3 RDO (Hiland) was informed. |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37015 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PERRY REGION: 3 |NOTIFICATION DATE: 05/22/2000|
| UNIT: [1] [] [] STATE: OH |NOTIFICATION TIME: 04:48[EDT]|
| RXTYPE: [1] GE-6 |EVENT DATE: 05/22/2000|
+------------------------------------------------+EVENT TIME: 02:53[EDT]|
| NRC NOTIFIED BY: BOB KIDDER |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JOHN MADERA R3 |
|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 |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - 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. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37018 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: QUAD CITIES REGION: 3 |NOTIFICATION DATE: 05/23/2000|
| UNIT: [] [2] [] STATE: IL |NOTIFICATION TIME: 00:11[EDT]|
| RXTYPE: [1] GE-3,[2] GE-3 |EVENT DATE: 05/22/2000|
+------------------------------------------------+EVENT TIME: 21:59[CDT]|
| NRC NOTIFIED BY: JOHN LECHMAIER |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |PATRICK HILAND R3 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 A/R Y 100 Power Operation |0 Hot Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - 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. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37019 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HOPE CREEK REGION: 1 |NOTIFICATION DATE: 05/23/2000|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 02:52[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 05/22/2000|
+------------------------------------------------+EVENT TIME: 23:38[EDT]|
| NRC NOTIFIED BY: NICK CONICELLA |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAN HOLODY R1 |
|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 4 Startup |4 Startup |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - 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. |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37020 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SAN ONOFRE REGION: 4 |NOTIFICATION DATE: 05/23/2000|
| UNIT: [] [2] [3] STATE: CA |NOTIFICATION TIME: 06:10[EDT]|
| RXTYPE: [1] W-3-LP,[2] CE,[3] CE |EVENT DATE: 05/23/2000|
+------------------------------------------------+EVENT TIME: 02:00[PDT]|
| NRC NOTIFIED BY: JACK FITCH |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CLAUDE JOHNSON R4 |
|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 |
|3 N Y 100 Power Operation |100 Power Operation |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PHYSICAL SECURITY REPORT - |
| |
| UNSECURED/UNATTENDED SECURITY GUARD WEAPON AND AMMUNITION INSIDE PLANT |
| PROTECTED AREA FOR ABOUT 25 MINUTES. COMPENSATORY MEASURES WERE IMMEDIATELY |
| TAKEN UPON DISCOVERY. THE LICENSEE PLANS TO NOTIFY THE NRC RESIDENT |
| INSPECTOR. CONTACT THE NRC OPERATIONS OFFICER FOR ADDITIONAL DETAILS. |
| |
| * * * UPDATE ON 5/23/00 @ 1342 BY PLUMLEE TO GOULD * * * RETRACTION |
| |
| LICENSEE IS RETRACTING THIS EVENT SINCE NO WEAPON WAS LOST. |
| |
| THE NRC RESIDENT INSPECTOR WILL BE INFORMED. THE REG 4 RDO (JOHNSON) WAS |
| NOTIFIED. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37021 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HOPE CREEK REGION: 1 |NOTIFICATION DATE: 05/23/2000|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 08:44[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 05/23/2000|
+------------------------------------------------+EVENT TIME: 05:05[EDT]|
| NRC NOTIFIED BY: ART BREADY |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAN HOLODY R1 |
|10 CFR SECTION: | |
|AINC 50.72(b)(2)(iii)(C) POT UNCNTRL RAD REL | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 4 Startup |4 Startup |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - HIGH PRESSURE COOLANT INJECTION SYSTEM INOPERABLE DUE TO STUCK OPEN CHECK |
| VALVE - |
| |
| At 0505 on 05/23/00, the High Pressure Coolant Injection (HPCI) System was |
| determined to be inoperable as a result of the discharge check valve being |
| stuck partially open. This condition was discovered during investigation of |
| a low injection header pressure alarm, and subsequent attempts to fill and |
| vent the discharge header were unsuccessful. It is believed that the check |
| valve stuck partially open when the system was secured after a low pressure |
| surveillance test at about 0305. The discharge check valve was mechanically |
| agitated at 0700, and reseated as evidenced by an audible sound and rise in |
| injection header pressure. |
| |
| At the time of discovery, the plant was in Operational Condition 2 with |
| reactor power at 4% and reactor pressure at approximately 500 psig. All |
| other safety related equipment was operable at the tune of discovery, with |
| the exception of the 'A' Residual Heat Removal Pump, which was aligned for |
| suppression pool cooling mode of operation. There was no significant impact |
| to overall plant safety as a result of this condition. |
| |
| Plant maintenance and engineering personnel are currently evaluating the |
| failure of the HPCI System discharge check valve. lnjection header fill and |
| vent is in progress to determine the amount of air that is present and |
| restore the system to an available condition. This information will be used |
| to determine if the safety function of the HPCI System was unavailable as a |
| result of the discharge check valve malfunction. |
| |
| A root cause investigation team has been assembled, and evaluation of system |
| and personnel performance is in progress. |
| |
| The licensee notified the NRC Resident Inspector and plans to notify local |
| officials. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37022 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: COMANCHE PEAK REGION: 4 |NOTIFICATION DATE: 05/23/2000|
| UNIT: [1] [2] [] STATE: TX |NOTIFICATION TIME: 11:07[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/23/2000|
+------------------------------------------------+EVENT TIME: 09:40[CDT]|
| NRC NOTIFIED BY: CHRIS ALEXANDER |LAST UPDATE DATE: 05/23/2000|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CLAUDE JOHNSON R4 |
|10 CFR SECTION: | |
|AARC 50.72(b)(1)(v) OTHER ASMT/COMM INOP | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| DISCOVERY OF A LOOSE CONNECTION WHICH CAUSED INOPERABLE PROCESS COMPUTER |
| SYSTEM (PCS) SITE DATA SYSTEM (SDS) COMMUNICATION IN THE EMERGENCY |
| OPERATIONS FACILITY (EOF) |
| |
| The following test is a portion of a facsimile received from the licensee: |
| |
| "During troubleshooting of an unrelated LAN problem, Telecommunications |
| [personnel] apparently bumped a cable causing the connection to become |
| loose. This loose connection caused the communication to TT10 and TT11 PCS |
| SDS in the EOF to become inoperable. The SDSs were inoperable from 05/22/00 |
| [at] 1523 [CST] to 05/23/00 [at] 0739 [CST] for a total of 16 [hours and] 16 |
| [minutes]." |
| |
| The licensee stated that the control room was notified of this condition at |
| 0940 CST on 05/23/00. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Other Nuclear Material |Event Number: 37023 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: LAKEHEAD PIPELINE |NOTIFICATION DATE: 05/23/2000|
|LICENSEE: LAKEHEAD PIPELINE |NOTIFICATION TIME: 11:23[EDT]|
| CITY: SUPERIOR REGION: 3 |EVENT DATE: 09/30/1997|
| COUNTY: DOUGLAS STATE: WI |EVENT TIME: [CDT]|
|LICENSE#: 22-26732-01 AGREEMENT: N |LAST UPDATE DATE: 05/23/2000|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |PATRICK HILAND R3 |
| |SCOTT MOORE NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: ROBERT POLLOCK | |
| HQ OPS OFFICER: LEIGH TROCINE | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|IBBF 30.50(b)(2)(ii) EQUIP DISABLED/FAILS | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| HISTORICAL EVENT REGARDING THE LOSS OF SHUTTER POSITION INDICATION ON A |
| BERTHOLD DENSITY GAUGE |
| |
| A nuclear gauge shutter failed during a routine semi-annual shutter test on |
| 09/30/97. The manufacturer of the nuclear gauge (Berthold Instruments) was |
| notified immediately, and a repair date of 10/07/97 was established. The |
| shield mechanism was replaced on 10/07/97. |
| |
| It was assumed that the shutter was in a partially closed position, but |
| there was no way to tell for sure. The gauge has a shaft control rod that |
| runs from the shutter to the turn knob. Apparently, the shaft control rod |
| broke, and the licensee could no longer determine the position of the |
| shutter. |
| |
| The licensee stated that survey meter readings near the gauge are normally |
| quite low when the shuttle is fully open. The licensee also stated that |
| this failure held no danger to employees or the public because normal |
| operation of the gauge is with the shutter fully open at all times. |
| |
| The nuclear gauge is located on a pipeline, and it is used for density |
| determination of the fluid flowing through the pipe. It has a 1,000-mCi of |
| cesium-137 source located on one side of the pipe and a detector located on |
| the other side of the pipe. |
| |
| The licensee stated that this issue was determined to be reportable during a |
| recent NRC inspection/audit. The licensee notified the NRC Region 3 office |
| (Mike Lafranzo). (Call the NRC operations officer for a site contact |
| telephone number.) |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021
Page Last Reviewed/Updated Thursday, March 25, 2021