| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/28/2004 - 05/31/2004 ** EVENT NUMBERS ** |
 | General Information or Other | Event Number: 40645 | Rep Org: MISSISSIPPI DIV OF RAD HEALTH
Licensee: LAMPKIN CONSTRUCTION COMPANY
Region: 1
City: VICKSBURG State: MS
County:
License #: MS-964-01
Agreement: Y
Docket:
NRC Notified By: Mark Langston, MS DRH
HQ OPS Officer: DICK JOLLIFFE | Notification Date: 04/05/2004
Notification Time: 16:41 [ET]
Event Date: 04/05/2004
Event Time: 13:30 [CDT]
Last Update Date: 05/28/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
LAWRENCE DOERFLEIN (R1)
SCOTT MOORE (NMSS)
DONNA-MARIE PEREZ (TAS)
| Event Text MS AGREEMENT STATE REPORT - STOLEN TROXLER MOISTURE DENSITY GAUGE
At 1641 EDT on 04/05/04, a MS Division of Radiological Health representative reported that at 1330 CDT on 04/05/04, a representative of Lampkin Construction Company, Vicksburg, MS, reported that between 1700 CDT on 04/02/04 and 0800 CDT on 04/05/04 a Troxler Moisture Density Gauge, locked in its transport case was stolen from a storage building at their office in Vicksburg, MS. A window had been broken and a door was forced open at the storage building. The Troxler Gauge, Model #3430, Serial #22349 contains an 8 milliCurie Cs-137 source and a 40 milliCurie Am-241-Be source. The licensee notified the MS Emergency Management Agency and the Warren County Sheriff. The licensee is planning to issue a press release.
* * * UPDATE VIA EMAIL FROM B. J. SMITH, MS DIVISION OF RADIOLOGICAL HEALTH ON 5/28/04 AT 0956 EDT TAKEN BY CROUCH * * *
The following information was received via e-mail:
"Troxler gauge recovered by Warren County Sheriff's department on 5-26-04 at approximately 3:30 PM. Gauge was located on the Mississippi River after water level had receded. Gauge was picked up by Division of Radiological
Health and returned to our office for leak test. No damage was observed to the gauge however it was apparent it had been under water. The gauge was determined to be free of contamination. Licensee Lampkin Construction RSO
was coming to pick up gauge at DRH office this morning. Licensee will be required to send gauge for repair and leak tests and furnish DRH with copy of results. This incident is closed as of 5-28-04."
* * * UPDATE VIA EMAIL FROM B. J. SMITH, MS DIVISION OF RADIOLOGICAL HEALTH ON 5/28/04 AT 1225 EDT TAKEN BY RIPLEY * * *
The following information was received via e-mail:
The location of the gauge recovery was clarified to be near a road close to Delta, Louisiana. The balance of the information remains the same.
Notified R1DO (Conte), NMSS (Torres), and TAS (Hahn). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 40665 | Facility: SALEM
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF MARTIN
HQ OPS Officer: CHAUNCEY GOULD | Notification Date: 04/11/2004
Notification Time: 04:43 [ET]
Event Date: 04/11/2004
Event Time: 00:23 [EDT]
Last Update Date: 05/30/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
LAWRENCE DOERFLEIN (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text PLANT HAD A VALID ESF SIGNAL TO START 1C EMERGENCY DIESEL GENERATOR
"THIS IS AN 8-HOUR NOTIFICATION TO REPORT A VALID ESF SIGNAL TO START 1C EMERGENCY DIESEL GENERATOR THAT OCCURRED ON 4/11/04 AT 0023. Salem Unit 1 is defueled. The spent fuel pool cooling system is providing decay heat removal. Spent fuel pool temperature is being maintained at 106 degrees. RCS temperature is 75 degrees. The RCS is vented to atmosphere. Reactor level is 97.5 feet. 1C emergency diesel generator is cleared and tagged for maintenance. 13 station power transformer was returned to service on 4/10/04 at 2347. The operating crew briefed expected response and abnormal procedures if the 4kv vital bus did not transfer during the retest of 13 station power transformer. Operations successfully retested 1A and 1B 4kv vital bus transfer from 14 station power transformer to 13 station power transformer and back to 14 station power transformer in accordance with station operating procedures. Operations attempted to retest the 1C 4kv vital bus transfer from 14 station power transformer to 13 station power transformer. When the operator attempted to close 13CSD in feed breaker, the 14CSD in feed breaker opened as designed, but 13CSD breaker failed to close. The 1C 4kv vital bus deenergized due to both offsite power source in feed breakers opening and 1C emergency diesel generator unavailable. An ESF signal to start the 1C emergency diesel generator was provided by the safeguards equipment cabinet {SEC}. Since the 1C emergency diesel was cleared and tagged for maintenance, it did not start. The cause of the failure to transfer is not known at this time. The operating crew implemented the appropriate abnormal operating procedures for the de-energized 1C 4kv vital bus. Operator and plant response was as expected. Decay heat continues to be removed by the spent fuel pool cooling system. Outage incident response team is evaluating and troubleshooting the cause of the loss of power to 1C 4kv vital bus. ORAM risk remains yellow. There were no personnel injured during the event."
The licensee will inform Lower Alloways Creek (LAC) Township and the NRC Resident Inspector.
* * * RETRACTED ON 5/30/04 AT 1103 EDT BY STEVE SAUER AND TAKEN BY GERRY WAIG * * *
The licensee provided the following information via facsimile:
"On April 11, 2004 at 0023 PSEG made an 8 hour notification to report a valid ESF actuation of the 1 'C' Emergency Diesel Generator (Event Number 40665). At the time of the event Salem Unit 1 was defueled. The spent: fuel pool cooling system was providing decay heat removal and spent fuel pool temperature was being maintained at 106 degrees. RCS temperature was 75 degrees and the RCS was vented to atmosphere. The 1 'C' EDG was cleared and tagged for maintenance.
"The 13 Station Power Transformer (SPT) had been returned to service on 4/10/04 at 2347. The operating crew was briefed on the expected response and reviewed the abnormal procedures if the 4kv vital bus did not transfer during the retest of 13 SPT. Operations successfully retested 1 'A' and 1 "B" 4kv vital bus transfer from 14 SPT to 13 SPT and back to 14 SPT in accordance with station operating procedures. When Operations attempted to retest the 1 'C' 4kv vital bus transfer from 14 SPT to 13 SPT, the 14 SPT in feed breaker opened as designed, but 13 SPT breaker failed to close and the 1 'C' 4kv vital bus de-energized. A signal to start the 1 "C" EDG was provided by the safeguards equipment cabinet (SEC), because the 1 'C' EDG was cleared and tagged for maintenance it did not start.
"Decay heat continued to be removed by the spent fuel pool cooling system.
"Subsequent Investigation into this event and further review of NUREG 1022 has determined that the condition described above is not reportable under the requirements of I0CFR50.72(b)(3)(iv) or 50.73(a)(2)(iv).
"NUREG-1022, section 3.2.6 states that valid signals are those signals that are initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. In this case, plant conditions were such that it did not require the 1 'C' EDG to be capable of starting and loading automatically in response to an undervoltage signal. The reactor was defueled and spent fuel cooling system providing fuel cooling was unaffected by the event.
"NUREG-1022 states that train level actuations are reportable. However, in this instance, the EDG did not actuate because it was removed from service and was not required to be operable. NUREG-1022 states that single component actuations are typically not reportable because single components of complex systems, by themselves, usually do not mitigate the consequences of significant events. With the 1 'C' EDG removed from service, the 1 'C' bus undervoltage signal is not sufficient to complete the full actuation logic to mitigate the event.
"Therefore this event was not reportable under 10CFR50.72 or 50.73, as per the guidance provided in NUREG 1022."
The licensee will notify the NRC Resident Inspector.
Notified R1DO (Richard Conte) |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 40714 | Facility: BRUNSWICK
Region: 2 State: NC
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: KEN HILL
HQ OPS Officer: RICH LAURA | Notification Date: 04/29/2004
Notification Time: 01:40 [ET]
Event Date: 04/28/2004
Event Time: 22:30 [EDT]
Last Update Date: 05/28/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
MIKE ERNSTES (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 96 | Power Operation | 96 | Power Operation | Event Text HPCI SYSTEM INOPERABLE FOLLOWING PLANNED MAINTENANCE
"During post maintenance testing following a High Pressure Coolant Injection (HPCI) System outage, the HPCI System was not declared operable due to unstable operation - oscillations of turbine speed (300-400 RPM), pump flow (600 GPM) and discharge pressure (300-550 psig) were seen in both automatic and manual flow control during the System Operability Periodic Test. HPCI had been declared inoperable at 0410 on 4/28/04 and placed under clearance to support planned maintenance on the Flow Controller, Flow Transmitter, system valves and condensate pump. The cause of the unstable operation is currently being investigated."
The licensee notified the NRC Resident Inspector.
* * * * RETRACTION FROM S. TABOR TO M. RIPLEY AT 1423 ET ON 5/28/04 * * * *
"On April 28, 2004, at 0410 hours, the High Pressure Coolant Injection (HPCI) system was declared inoperable to support scheduled maintenance on the HPCI system. To satisfy post maintenance test requirements and support restoring the HPCI system to an operable status, surveillance test, OPT-09.2, "HPCI System Operability Test," was performed. During this testing at 2230 hours, oscillations in pump flow, pressure, and turbine speed were observed in both the automatic and manual flow control modes of operation. Based on the test results, HPCI remained inoperable until the cause of the oscillations could be identified, corrective actions implemented, and the system satisfactorily tested. On April 29, 2004, at 0140 hours, the NRC was conservatively notified (Event Number 40714), in accordance with 10 CPR 50.72(b)(3)(v)(D), of a condition that at the time of discovery could have prevented the fulfillment of the HPCI safety function.
"Troubleshooting determined that the HPCI flow controller to the 2-E41-C002-CNV Ramp Generator Signal Converter (RGSC) was subject to spurious deviations. The RGSC was removed from its installed position and bench tested. Circuit review and testing determined that electronic component degradation was the most likely cause of the RGSC output signal perturbations. A new RGSC was calibrated and installed. After installation, in place testing showed that the new ramp generator did not exhibit signal variations. A final HPCI system operability test was performed and verified that the HPCI system was responding normally. On May 1, 2004, at 1220 hours, the HPC1 system was restored to service.
"Reportability Discussion:
NUREG-1022, Rev. 2, Section 3.2.7 (page 56) lists types of events or conditions that are generally not reportable under 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v) criteria. The list of not-reportable conditions includes: Removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function).
"On April 28, 2004, the HPCI system was removed from service to support a planned maintenance system outage. In addition, surveillance testing was performed to support system restoration following the maintenance outage for testing in accordance with an approved surveillance procedure. Based on the post maintenance test results, the HPCI system was declared inoperable and since the HPCI system is a single train safety system, an ENS notification was made. However, further evaluation of the condition determined that no condition was discovered that could have prevented the HPCI system from performing its functions. The following information provides the basis for that determination:
"Review of all applicable operating data collected during HPCI system testing performed from the time of discovery of the oscillation concern indicates that (1) the resulting HPCI speed spikes were in the positive direction and therefore no concern related to the inability of HPCI to provide adequate vessel level makeup existed, (2) the RGSC output perturbations followed a consistent pattern and the magnitude of the associated control problem would not have become more severe for a period longer than the assumed HPCI mission time, and (3) none of the excursions experienced were high enough to cause a system trip at the 5000 plus/minus 100 rpm overspeed trip setpoint. Even if a spurious speed spike resulted in a HPCI overspeed trip, the HPCI overspeed trip is designed to automatically reset and allow the system to ramp back up to operating speed. Given these facts, there was no observed performance that represented a loss of system function. Had HPCI been called upon to inject during the time that the condition resulting in HPCI flow oscillations existed, the system would have met all functional requirements.
"Carolina Power & Light Company, doing business as Progress Energy Carolinas, Inc., has determined that this event does not meet the 10 CFR 50.72 or 10 CFR 50.73 reporting criteria and the notification for Event Number 40714, is retracted. The resident inspector has been notified."
Notified R2 DO (R. Ayres). |
| General Information or Other | Event Number: 40783 | Rep Org: ROTORK CONTROLS INC
Licensee: ROTORK CONTROLS INC
Region: 1
City: ROCHESTER State: NY
County:
License #:
Agreement: Y
Docket:
NRC Notified By: KAREN BLACK
HQ OPS Officer: MIKE RIPLEY | Notification Date: 05/28/2004
Notification Time: 15:51 [ET]
Event Date: 05/28/2004
Event Time: [EDT]
Last Update Date: 05/28/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
RICHARD CONTE (R1)
DAVID AYRES (R2)
SONIA BURGESS (R3)
MIKE RUNYAN (R4)
NRR PART 21 CONTACTS (NRR)
NMSS PART 21 CONTACT (NMSS)
| Event Text PART 21 NOTIFICATION - VALVE ACTUATOR SWITCH MECHANISMS
"Basic Component Affected
Rotork NA1 switch mechanism assemblies manufactured between 1978 (post 78 build) and November 2001 supplied either as a spares item or fitted in an NA1 type Electric Valve Actuator. Customers supplied with potentially affected actuators manufactured between January 1998 and November 2001, were previously notified individually of this condition and may have completed the risk assessment and corrective action detailed below. This report extends the affected time frame, potentially affecting actuators not identified on the original notifications.
"Rotork NA4, NA5, NA1E and NAE5 type Electric Valve Actuators are not affected.
"Nature of the Defect and Associated Safety Hazard
It has recently been identified that the molded (PPS) components within the switch mechanism assembly have a low level of crystallinity and it cannot be confirmed that they are to the same specification as those originally tested and qualified at Wyle in 1978 (refer test report 43979-1 Rev A).
Effect on Functionality
The report provides a detailed explanation of the effect on function depending on the valve position and open/close action.
"NA4 and NAS type Electric Valve Actuators have a maximum ambient temperature rating of 70 deg C (160 deg F) and are not affected. NA1 type Electric Valve Actuators have the same ambient temperature rating but can be subject to a loss of coolant accident (LOCA). The condition reported will only affect Valve Actuators in plant locations where: 1) the LOCA and Operational temperature specifications, as defined in the Design Basis Document for each facility and location, could result in the switch mechanism components exceeding 80 deg C (176 deg F) and 2) The switch mechanism fitted was manufactured between 1978 and November 2001 and 3) The actuator is exposed to the two conditions outlined in section 2.0 of the report." |
| Power Reactor | Event Number: 40784 | Facility: NORTH ANNA
Region: 2 State: VA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PAGE KEMP
HQ OPS Officer: ARLON COSTA | Notification Date: 05/29/2004
Notification Time: 09:55 [ET]
Event Date: 05/29/2004
Event Time: 02:02 [EDT]
Last Update Date: 05/29/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
DAVID AYRES (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Hot Standby | 0 | Hot Standby | Event Text CONTROL ROD POSITION INDICATION FAILURE DURING HOT ROD DROP TEST
"During performance of Hot Rod Drop Testing [and] when withdrawing 'D' Control banks, a failure of Group 1 Position Indication was identified. Entered action of Technical Requirement Manual [TRM] 3.1.3 and opened the Reactor Trip Breakers within 15 minutes per action [statement] of TRM 3.1.3."
After the Reactor Trip Breakers were opened all Group 1 "D" Control Rods fully inserted into the core. There were no reactivity concerns since the reactor was borated with adequate shutdown margin. The failure of the position indicator has been identified and repaired. There were no other issues associated with this incident and the licensee will proceed with Hot Rod Drop Testing while at 0% reactor power and Mode 3.
The licensee will notify the NRC Resident Inspector. |
| Power Reactor | Event Number: 40785 | Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [ ] [ ]
RX Type: [2] GE-5
NRC Notified By: MIKE BRANDON
HQ OPS Officer: MIKE RIPLEY | Notification Date: 05/30/2004
Notification Time: 18:40 [ET]
Event Date: 05/30/2004
Event Time: 11:37 [PDT]
Last Update Date: 05/30/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
MIKE RUNYAN (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 65 | Power Operation | 65 | Power Operation | Event Text HIGH PRESSURE CORE SPRAY SYSTEM INOPERABLE
"The High Pressure Core Spray (HPCS) system was declared inoperable due to the pump's failure to meet the flow requirement specified in TS Surveillance Requirement 3.5.1.4. This surveillance is normally performed on a quarterly basis in accordance with the plant's In-service Testing (IST) Program. The flow values measured during the performance of this surveillance were below both the normal and alert ranges. This test had also been run on 5/27/04 with results in the alert range; HPCS system instruments had been vented between the two tests to rule out the possibility that the results were due to measurement errors.
"Upon declaring the HPCS pump inoperable, TS 3.5.1 Action B was entered. In accordance with Action B, the Reactor Core Isolation Cooling (RCIC) system was verified to be operable. With the RCIC system verified operable, Action B provides a 14-day completion time to restore HPCS to an operable status.
"All other Emergency Core Cooling Systems (ECCS) were operable during this event. This event is being reported as an event or condition that could have prevented the fulfillment of a safety function credited for mitigating the consequences of an accident. The HPCS system is a single train system at Columbia."
The licensee notified the NRC Resident Inspector. | |