| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
11/03/2006 - 11/06/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42915 | Facility: SUMMER
Region: 2 State: SC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: PETE SNYDER | Notification Date: 10/17/2006
Notification Time: 17:50 [ET]
Event Date: 10/17/2006
Event Time: 14:15 [EDT]
Last Update Date: 11/04/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
JAY HENSON (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text PRESSURE BOUNDARY LEAKAGE IDENTIFIED DURING AN OUTAGE
"On October 14th V.C. Summer Nuclear Station shut down for a routine refueling outage. At 1415 on October 17, 2006 during an inspection of the reactor vessel upper head, personnel identified a small leak in a weld in the Reactor Vessel Level Instrumentation System (RVLIS). This weld is located in a 3/4 inch stainless steel instrumentation line. The line is attached to, and is approximately three feet from the reactor vessel head. This leak resulted in a small amount of boron accumulation in an isolated area. There is no indication that any safety systems were affected."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED BY ARNIE CRIBB TO JEFF ROTTON AT 1521 ON 11/04/06 * * *
"EN #42915 was submitted by South Carolina Electric & Gas Company based on the initial indication of a pressure boundary leak in the Reactor Vessel Level Instrumentation System (RVLIS). South Carolina Electric & Gas Company retracts EN #42915 based on the following discussion.
"During subsequent cleaning and inspection of the suspect weld, which was preserved in its as-found condition for root cause analysis, no flaw in the analyzed RVLIS piping or pipe coupling was found. Further investigation was conducted of the area under the Control Rod Drive Mechanism (CRDM) shroud near the RVLIS line. The source was determined to be leakage through the lower canopy seal weld of CRDM No. 25, which resulted in accumulation of boric acid on the suspect RVLIS line.
"Since the RVLIS line was not leaking and the CRDM canopy connection is a mechanical (screwed) fitting with a seal weld, there has been no pressure boundary leakage as defined by Technical Specifications. The canopy seal weld is neither a structural weld nor a pressure retaining weld comprising the reactor coolant system pressure boundary. A permanent plant design was developed and the canopy seal was repaired using a standard canopy seal clamping process.
"A 100 % bare metal visual inspection of the reactor vessel head was completed during this refueling outage. The results confirmed that there has been no degradation."
The licensee will notify the NRC Resident Inspector.
Notified the R2DO (Evans) |
| Power Reactor | Event Number: 42957 | Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: JOE SHOFFNEN
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/02/2006
Notification Time: 14:53 [ET]
Event Date: 11/02/2006
Event Time: 13:34 [EST]
Last Update Date: 11/03/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
JOHN WHITE (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC TURBINE TRIP/ REACTOR TRIP DUE TO INVALID LOW CONDENSER VACUUM SIGNAL
"At 1334 on 11-2-06 an Automatic Reactor Trip occurred from 100% power. All systems functioned as required. One safety valve stuck open on both OTSGs. They subsequently re-seated. An employee working on the roof at the time of the trip fell off a ladder and injured his leg. Emergency medical was contacted to assist with the injured worker. Two fire trucks and an ambulance was dispatched to the site to remove the injured worker. The worker was not contaminated. There is no indication of any OTSG tube leaks. Initial investigation indicates the reactor tripped, due to a turbine trip due to an invalid low vacuum signal."
State and local officials will be notified of this event by the licensee.
I&C Techs were performing maintenance on one of the low pressure vacuum switches. An electrical fault fed to the other two low pressure vacuum switches causing a 2 out of 3 signal which resulted in a turbine trip followed by a reactor trip signal, as expected. All rods fully inserted into the core. One safety valve (9 safety valves on each OTSG) on each Once Through Steam Generator stuck open. OTSG "B" safety relief valve was open less than one minute. There are no leaking OTSG tubes. A condensate relief valve located in the turbine building opened/shut - nobody injured. The ICS (Integrated Control System) operated as expected. All emergency core cooling systems and the emergency diesel generators are fully operable plus the electrical grid is stable.
A licensee working on the industrial coolers on top of the industrial building, standing on a ladder, fell off the ladder when OTSG relief valve opened. Licensee either broke or badly sprained his leg.
The NRC Resident Inspector was informed of this event by the licensee.
* * * UPDATE ON 11/03/06 AT 1607 EST FROM ADAM MILLER TO MACKINNON * * *
"Post trip evaluation determined that the Main Steam safety valves were not stuck open. The safety valves were operating within their tolerance band. The "B" OTSG Main Steam safety valve reseated with no operator action as steam pressure decreased. The "A" Main Steam safety valve was reseated when operators lowered OTSG pressure in accordance with Plant Operating Procedures. TMI-1 issued a press release on this event at 15:13 on 11/2/06." R1DO (John White) notified.
The NRC Resident Inspector was notified of this update by the licensee. |
| General Information or Other | Event Number: 42958 | Rep Org: ANDREWS ENVIRONMENTAL ENGINEERING
Licensee: ANDREWS ENVIRONMENTAL ENGINEERING
Region: 3
City: INDIANAPOLIS State: IN
County:
License #: 13-32079-01
Agreement: N
Docket:
NRC Notified By: STEVE REUTER
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/02/2006
Notification Time: 16:24 [ET]
Event Date: 11/02/2006
Event Time: 16:00 [EST]
Last Update Date: 11/03/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
INFORMATION ONLY
| Person (Organization):
MONTE PHILLIPS (R3)
SCOTT MOORE (NMSS)
| Event Text DAMAGED TROXLER MOISTURE DENSITY GAUGE.
The outer case and the key pad to a Troxler, Model number 3440, was damaged when a vehicle struck the case. The rod to the gauge still operates properly. The location of the incident occurred at the Newton County landfill near Brook, IN. The RSO will go to the site to take radiation surveys. The serial number of the gauge is 21041. The gauge contains 8 millicuries of Cesium-137 and 40 millicuries of Am-241/Be.
* * * UPDATE FROM STEVE REUTER TO JOE O'HARA ON 11/3/06 AT 1152 * * *
Licensee reported that radiation surveys of the gauge indicated less than 3 milliRem/hr approximately six inches from the device, and the carry case isn't damaged. The licensee has recovered the gauge and will perform a leak test prior to forwarding it to the vendor via FedEx.
Notified R3DO(Monte Phillips) and NMSS ( Sandra Wastler). |
| Power Reactor | Event Number: 42961 | Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [ ] [ ]
RX Type: [2] GE-5
NRC Notified By: A J MOORE
HQ OPS Officer: JOHN MacKINNON | Notification Date: 11/03/2006
Notification Time: 12:04 [ET]
Event Date: 11/03/2006
Event Time: 03:09 [PST]
Last Update Date: 11/03/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
MICHAEL SHANNON (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text PROCEDURAL ERROR CAUSED RESIDUAL HEAT REMOVAL SHUTDOWN COOLING INBOARD CONTAINMENT ISOLATION VALVE TO CLOSE.
"This event notification is being made to report an event that could have prevented the fulfillment of the safety function to remove residual heat. On November 3, 2006, the plant was shutdown in Mode 4 (Cold Shutdown) with the reactor vessel level being maintained between 60-80 'inches' and temperature was being maintained 110-120F. Operators were performing a procedure to transfer RPS B to its alternate power supply. During the procedure, RHR-V-9 (RHR shutdown cooling inboard containment isolation valve) closed and shutdown cooling was interrupted. Reactor Recirculation Pump A remained in service providing forced flow through the core. Operators reopened RHR-V-9 and restored shutdown cooling. Reactor temperature reached 148F and vessel level reached 95 'inches' (bottom of main steam lines is 116 inches) while shutdown cooling was not in service. Shutdown cooling was restored in 45 minutes (120 minutes time to boil from 120F) after RHR-V-9 went closed. Reactor temperature and level were restored to their previous operating bands by 0451 PST. Initial indications are that RHR-V-9 closed due to an error in the procedure which failed to remove control power from RHR-V-9 during the RPS transfer. Without removing control power, a motor-closed demand would have been created in the circuit during the transfer and upon restoration of power to the valve motor, the valve would motor close. Further investigation is ongoing."
The NRC Resident Inspector was notified of this event by the licensee. |
| Power Reactor | Event Number: 42963 | Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: STEVEN OAKLEY
HQ OPS Officer: JEFF ROTTON | Notification Date: 11/04/2006
Notification Time: 20:37 [ET]
Event Date: 11/04/2006
Event Time: 16:30 [EST]
Last Update Date: 11/04/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
MONTE PHILLIPS (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Startup | 0 | Startup | Event Text AUXILIARY FEEDWATER NOT ALIGNED PROPERLY IN STARTUP MODE
"On 11/3/2006 at 1834 hours, it was discovered that, with the Plant in Mode 2, both trains of Auxiliary Feedwater were not in the required pump automatic start configuration. One pump was operating and performing the heat removal function. Both Auxiliary Feedwater train pump controls were immediately placed into the Automatic configuration. It was determined that the pump start controls had been placed in Manual on 11/1/2006 at about 2152 hours. Upon discovery, it was believed that the Auxiliary Feedwater heat removal function was satisfied because one train was operating and had been in operation since before 2152 hours on 11/1/2006.
"Subsequently, on 11/4/2006 at 1630 hours it was determined that, with the pump controls in manual, coincident with a postulated Loss of Offsite Power and an Auxiliary Feedwater Actuation Signal, the operating Auxiliary Feedwater pump would trip and neither train pump would automatically start.
"This situation would constitute a condition that could have prevented the fulfillment of the safety function to mitigate the consequences of a Loss Of Offsite Power event and is reportable under 50.72(b)(3)(v)D.
"Had such an event occurred, the Operators would have restored Auxiliary Feedwater flow in accordance with station procedures."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42964 | Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: ROBERT BRUNS
HQ OPS Officer: JEFF ROTTON | Notification Date: 11/04/2006
Notification Time: 21:01 [ET]
Event Date: 11/04/2006
Event Time: 18:10 [EST]
Last Update Date: 11/04/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JOHN WHITE (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 20 | Power Operation | 20 | Power Operation | Event Text HPCI INOPERABLE
"During HPCI testing with reactor pressure at 975 psig, HPCI flow and turbine speed oscillations occurred. The HPCI turbine was run for approximately one minute and was manually tripped by the operator when system conditions did not improve.
"A cause investigation is underway.
"All other ECCS and RCIC remain operable.
"The unit is in a 14 day LCO for this event.
"The NRC Resident Inspector has been notified." |
| 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. | |