| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/06/2006 - 11/07/2006 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 42966 | Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: EDWARD HARRISON
HQ OPS Officer: PETE SNYDER | Notification Date: 11/06/2006
Notification Time: 04:57 [ET]
Event Date: 11/06/2006
Event Time: 01:10 [CST]
Last Update Date: 11/06/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
MONTE PHILLIPS (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 98 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC REACTOR SCRAM DURING TURBINE TESTING
"On 6 November 2006 at 0110 an automatic reactor scram was received. This report is being made under 10 CFR 50.72 (b)(2)(iv)(B), 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical.' The plant also received Group 2, 3 and 4 isolations from the low level signal as reactor water dropped due to the reactor scram. The isolations are being reported under 10 CFR 50.72 (b)(3)(iv)(A), 'any event or condition that results actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section.' The scram is believed to be the result of a turbine trip but is currently being investigated further. Main Turbine surveillance testing was in progress at the time but it is not known if the testing caused the turbine trip and subsequent Reactor Scram. Reactor pressure is currently being controlled by the turbine bypass valves and steamline drains. No actuations of Safety Relief Valves were required or occurred. Both Reactor Recirculation pumps tripped as part of RPT. The 'B' Recirc pump has been restarted. Reactor level dropped following the scram resulting in the isolations but was recovered and is currently being maintained by the normal feedwater systems. All isolations went to completion and have been reset at this time. The plant electrical buses transferred without incident and are currently in a normal shutdown alignment.
"The current plan is to remain in Mode 3 while investigations determine the cause of the turbine trip and reactor scram."
All control rods fully inserted on the trip. Currently the plant is using offsite power but diesel generators are available. The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42967 | Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: TIMOTHY PAGE
HQ OPS Officer: PETE SNYDER | Notification Date: 11/06/2006
Notification Time: 11:51 [ET]
Event Date: 09/08/2006
Event Time: 18:28 [EST]
Last Update Date: 11/06/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
NEIL PERRY (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 98 | Power Operation | 98 | Power Operation | Event Text HALF SCRAM SIGNAL DUE TO EQUIPMENT FAILURE
"On September 8, 2006 at approximately 1828, with the James A. FitzPatrick Nuclear Power Plant (JAF) operating at 98% reactor power, an equipment malfunction occurred, resulting in the tripping of one of the two 'B' Reactor Protection System (RPS) motor generator Electrical Protection Assembly (EPA) breakers. This caused a corresponding loss of power to the 'B' RPS bus, a half scram and primary containment isolation valve closures. Appropriate actions were taken in accordance with Abnormal Operating Procedure (AOP)-60, 'Loss of RPS Bus B Power.' All equipment operated as designed as a result of the loss of power to the 'B' RPS bus.
"At 1859, the 'B' RPS bus was placed on the alternate power supply. At approximately 1905, the RPS half scram and Primary Containment Isolation System (PCIS) isolation signal was reset. Restoration of the various isolated systems commenced shortly thereafter. The reactor remained at 98% power throughout the event and Technical Specifications (TS) Limiting Condition for Operation (LCO) Actions were entered for the affected equipment as required.
"The above event meets the reporting criteria of 10 CFR 50.73(a)(2)(iv)(a) since the loss of RPS bus resulted in primary containment isolation signals affecting containment isolation valves in more than one system. The following systems or components isolated as a result of the loss of 'B' RPS bus: Reactor Water Cleanup, Reactor Building Ventilation, 'B' Containment Atmosphere Dilution, Torus Vent and Purge, Drywall Equipment and Floor Drain Sumps, 'B' Drywall Containment Atmospheric Monitors, Recirculation System Sample Line, Main Steam Line Drains, Residual Heat Removal drain valve to radwaste, Standby Gas Treatment (auto initiated).
"Since the signal was invalid (the result of loss of power vs. an actual containment isolation condition), this event meets the criteria in 10 CFR 50.73(a)(1) for being reported as a 60 day telephone notification in lieu of a written LER.
"The loss of the 'B' RPS bus was caused by the tripping of one of the two in-series 'B' RPS motor generator EPA breakers. A Failure Modes and Effects Analysis was completed for the EPA breaker. Potential failure modes were evaluated and ruled out with the exception of failure of a subcomponent on the EPA breaker logic card. The logic card and breaker were replaced, restoring the 'B' RPS motor generator to service."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42968 | Facility: RANCHO SECO
Region: 4 State: CA
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP
NRC Notified By: BOB JONES
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/06/2006
Notification Time: 17:20 [ET]
Event Date: 11/06/2006
Event Time: 09:30 [PST]
Last Update Date: 11/06/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
JEFFREY CLARK (R4)
MELVYN LEACH (IRD)
EDWIN HACKETT (NMSS)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Decommissioned | 0 | Decommissioned | Event Text ONE AND POSSIBLY FIVE OTHER CLASS "C" FUEL ASSEMBLIES WERE SINGLED OUT AS HAVING THE POTENTIAL OF HAVING FLAWS ABOVE THE HAIRLINE CRACK OR PINHOLE LEAK CRITERIA.
This event is a 24 hour notification per Technical Specification 2.2.1.
Based upon the differences in the definition of damaged fuel presented in RSAP-0112, and the commitments made in various licensing documents, it was determined that a potential existed that damaged assemblies may have been placed in a fuel canister licensed only for intact assemblies. A potential problem exists with fuel assemblies, which had been categorized as class "C" assemblies in the RSAP-0112 inspection results, and therefore would not have had any restrictions on the type of canister to be placed. Class "C" assemblies were those assemblies not meeting the requirements of F2 for gross cladding defects, did not have cracks which exceeded dimensional of approx. 0.34 inches across and 0.7 inches high.
Based on further review of the RSAP-0112 inspection results, six fuel assemblies were upgraded as potentially damaged fuel assemblies because the written inspection reports indicated the existence of a crack or missing cladding greater than the class "C" limit.
With the exception of assembly 2G6, the video portion of the tape is not 100% conclusive that a crack greater than a hairline exists on the other five fuel assemblies. Based upon the images available of assembly 2G6, the crack on this assembly was scaled to be approx. 0.04 inches wide by 0.25 inches long. The film quality of the remaining 5 assemblies is not sufficient to provide such detailed scaling but is estimated for discussion purposes.
NRC Project Manager (Randy Hall) was notified of this by the licensee. | |