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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 THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED  !!!!!!!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.)                                                           |
+------------------------------------------------------------------------------+