| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/15/2006 - 06/16/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42622 | Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: MIKE HIMEBAUCH
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/06/2006
Notification Time: 23:09 [ET]
Event Date: 06/06/2006
Event Time: 16:00 [EDT]
Last Update Date: 06/15/2006 | 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 | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text FAILURE TO MEET EMERGENCY EQUIPMENT SERVICE WATER SURVEILLANCE REQUIREMENT
"On June 6, 2006 at 1600 EDT the Division 2 Emergency Equipment Service Water System (EESW) was in service for a planned surveillance test when the system failed to achieve required flows as specified in the surveillance. These flow rates are acceptance criteria and therefore resulted in system inoperability. EESW cools the Emergency Equipment Cooling Water (EECW) System which in turn cools various safety related components including the High Pressure Coolant Injection (HPCI) System Area Cooler. Unplanned HPCI inoperability occurred due to the Division 2 EECW/EESW inoperability based on loss of the HPCI System Area Cooler. A 14 day Limiting Condition for Operation (LCO) was entered for HPCI per LCO 3.5.1. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident, based on loss of a single train safety system."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 08:14 ON 6/15/2006 FROM JEFF GROFF TO ABRAMOVITZ * * *
"On June 6, 2006 at 1600 EDT, during the performance of the quarterly pump and valve operability surveillance test on Division 2 of the Emergency Equipment Service Water System (EESW), the minimum pump flow required by the procedure to perform the test could not be established. Because minimum pump flow could not be established, Division 2 of EESW was declared inoperable. EESW cools the Emergency Equipment Cooling Water (EECW) System which in turn cools various safety related components including the High Pressure Coolant Injection (HPCI) System Area Cooler. HPCI was declared inoperable based on loss of the HPCI System Area Cooler due to the Division 2 EECW/EESW inoperability. A 14 day Limiting Condition for Operation (LCO) was entered for HPCI per LCO 3.5.1. A report was made to the NRC pursuant to 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident, based on loss of a single train safety system.
"On June 7, 2006, the test was re-performed with the valve in the bypass line around the heat exchanger Temperature Control Valve (TCV) throttled open. The required pump flow was established and the surveillance was successfully completed at 1815 EDT.
"Further Engineering evaluation concluded that minimum pump flow could not be established on June 6, 2006 due to normal pump wear and heat exchanger fouling. The pump flow required for performing the pump and valve operability surveillance test was established to monitor pump degradation and is higher than the flow required for the EESW system to perform its safety function. It has been verified that the measured flow exceeds the system design basis required flow with an adequate margin and that the pump and heat exchanger remain adequate to support the HPCI room cooling operation. The HPCI safety function was maintained throughout this period; therefore, this event is being retracted."
The licensee notified the NRC Resident Inspector. Notified the R3DO (Louden). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42641 | Facility: OCONEE
Region: 2 State: SC
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: RANDY TODD
HQ OPS Officer: PETE SNYDER | Notification Date: 06/14/2006
Notification Time: 19:08 [ET]
Event Date: 06/14/2006
Event Time: 14:00 [EDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
BRIAN BONSER (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 LEAKING DECAY HEAT REMOVAL ISOLATION VALVE BYPASS LINE
"On 2-21-06, during a tour of containment during normal operation at 100% power, a small leak (one (1) to three (3) drops per second) was noted a 1/2 inch line connected to the decay heat removal (DHR) drop line. It was identified as being a body-bonnet leak on valve 1LP-167 subject to a TS limit of 10 gpm.
"At approximately 1400 hours on 6-14-06 following a shutdown for an unrelated issue, the source was identified as a leak at a weld in a "tee" joint adjacent to 1LP-167. This is considered RCS pressure boundary leakage, subject to a TS limit of zero leakage. The leak was isolated by closing a normally open valve in the 1/2 inch line and the leakage stopped.
"Initial Safety Significance: The leak is in a 1/2 inch line which provides over pressure protection from thermal expansion in the volume between 1LP-1 and 1LP-2 (the main pressure boundary isolation valves between the high pressure RCS and the LPI (DHR) system). The leak rate (1 to 3 drops per second) was not significant, except that it was RCS pressure boundary leakage. 1LP-1 is normally closed, but must be opened to establish a DHR path. Valve 1LP-167 is a 1/2 inch check valve which would have limited RCS leakage. Thus, if the leak had grown, it would have been limited to the amount of seat leakage past either 1LP-167 or 1LP-1. It would also have been limited by the 1/2 inch size of the line containing the leak."
Technical Specification LCO 3.4.13 applies to RCS leakage in modes 1 to 4. The licensee plans to fix the leak prior to entry into mode 4.
The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 00:15 ON 6/16/2006 FROM SAM LARK TO ABRAMOVITZ * * *
"On 6-14-06 at 1908 hours Oconee reported an RCS pressure boundary leak in a 1/2 inch line connected to the decay heat removal (DHR) line near valve 1LP-1 inside containment. Oconee has reviewed the event in greater detail and has concluded that the event is not reportable. The Basis for TS 3.4.13 states that RCS LEAKAGE includes leakage from connected systems up to and including the second normally closed valve (or outermost isolation valve for systems penetrating containment). However TS 1.1 contains a definition of LEAKAGE which includes 'Pressure Boundary LEAKAGE: LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.' The leakage in this event was isolable, and therefore does not meet the definition of Pressure Boundary LEAKAGE. Therefore the zero leakage criterion of TS 3.4.13 does not apply to this leak. The applicable criterion is 10 gpm identified LEAKAGE. Since the leak does not meet the criterion as Pressure Boundary LEAKAGE, the leak was isolable, and the applicable TS LEAKAGE limit was not exceeded, this event does not meet the reportability criteria for 10 CFR 50.72 or 50.73 and event notification 42641 is hereby RETRACTED.
"Additional information and clarification: "During normal operation the leak was isolated by one barrier (valves 1LP-167 and 1LP-1, closed in parallel). The leakage observed on 2-21-06 during a containment tour at Mode 1 was recorded as 1 drop per second. As stated in the initial notification, at that time the leak was believed to be a body-bonnet leak. It was observed at Mode 1 again on 5-25-06 and recorded as 3 drops/second. On 6-14-06, the leakage was recorded as one drop/second while at reduced pressure in Mode 4, before the DHR systems was placed in service. At that point, the leak was isolated by closing an additional valve (1LP-166, normally open), and the leak stopped. The Low Pressure Injection system was placed in service for DHR, which opened 1 LP-1. Later, with system pressure at approximately 285 psig in Mode 5 (outside the applicability of TS 3.4.13), 1LP-166 was reopened to allow additional verification of the leak location. At that time the leak was described as a 'spray' but no leak rate was measured before 1LP-166 was reclosed. The leak rate at that time was estimated as well less than 10 GPM.
"Corrective Action: The affective section of 1/2 inch pipe and associated fittings have been removed for transfer to a Duke laboratory for analysis. Repairs will be completed prior to return to mode 4."
The licensee notified the NRC Resident Inspector. Notified the R2DO (Bonser). |
| Power Reactor | Event Number: 42642 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GRANT FERNSLER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/15/2006
Notification Time: 05:05 [ET]
Event Date: 06/15/2006
Event Time: 03:00 [EDT]
Last Update Date: 06/15/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
DAVID SILK (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 REACTOR SCRAM WHILE SHIFTING POWER SUPPLIES
"At approximately 0300 hours on 15 June, the Susquehanna Unit One reactor automatically scrammed due to an apparent neutron monitoring trip while transferring Reactor Protection System power supplies. All rods [fully] inserted, and both reactor recirculation pumps tripped. Reactor water level lowered to -38" causing level 3 (+13") and level 2 (-38")isolations, and was restored to normal level (+35") by RCIC and subsequently the feedwater system. All isolations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected.
"A reactor recirculation pump was restarted to re-establish forced core circulation. The reactor is currently stable in condition 3. An investigation into the cause of the shutdown is underway. Unit Two continued power operation.
"The NRC resident inspectors were notified. A press release will occur."
After the scram, HPCI automatically started but was manually shut down with RCIC maintaining vessel level. Decay heat removal is being maintained with main feedwater and the turbine steam dumps. The electrical grid is stable. No major LCOs were in affect at the time of the event. |
| Power Reactor | Event Number: 42643 | Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: JEFF GROFF
HQ OPS Officer: BILL GOTT | Notification Date: 06/15/2006
Notification Time: 13:38 [ET]
Event Date: 06/15/2006
Event Time: 10:53 [EDT]
Last Update Date: 06/15/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
PATRICK LOUDEN (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | A/R | Y | 100 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC REACTOR SCRAM DUE TO TURBINE TRIP
"At 10:53 on 6/15/06, a reactor scram, occurred due to a Turbine/Generator Trip. All control rods fully inserted into core. The lowest vessel water level reached was 134 inches. Water level is now being controlled in the normal water level band using Condensate/Feedwater system. No SRVs lifted. RPV pressure is being controlled by the Turbine Pressure Regulator.
"At the time of the scram, 2B Main Transformer cleaning was taking place. The initial alarm was 'Main Transformer 2B Oil Temp Hi' followed by Generator Differential Relaying and a Turbine Trip. Transformer Deluge also initiated. An investigation is in progress to determine the specific cause for the initiating event.
"Group 13 'Drywell Sumps' isolated on Level 3 as expected. At the time of the scram, all ECCS systems and Emergency Diesel Generators were Operable."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42644 | Facility: SAN ONOFRE
Region: 4 State: CA
Unit: [1] [2] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: LINDA CONKLIN
HQ OPS Officer: MIKE RIPLEY | Notification Date: 06/15/2006
Notification Time: 15:31 [ET]
Event Date: 06/13/2006
Event Time: 22:05 [PDT]
Last Update Date: 06/15/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
ANTHONY GODY (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Decommissioned | 0 | Decommissioned | | 2 | N | Y | 99 | Power Operation | 99 | Power Operation | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFSITE NOTIFICATION - AUTO ACCIDENT FATALITY
"On Tuesday, June 13, 2006, at approximately 2205 PDT, an on duty SONGS security officer driving an SCE vehicle struck a pedestrian on a public road outside the Owner Controlled Area. At that time, the officer was performing routine rounds and was driving between the plant and SCE's MESA facility.
"SCE notified local authorities including the California Highway Patrol and the Orange County Fire Authority.
"The individual sustained serious injuries and was transported to a local hospital. SCE has been notified unofficially that the individual passed away on June 14, 2006. The identity of the individual is unknown by SCE. SCE is reporting this occurrence in accordance with 10CFR50.72(b)(2)(xi). No news release is planned by SCE at this time.
"At the time of this report, Units 2 and 3 were operating at about 99 and 100 percent power, respectively. The NRC Resident Inspectors have been notified of this occurrence and will be provided with a copy of this report." |
| Power Reactor | Event Number: 42647 | Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: SHAWN ELLIOTT
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/16/2006
Notification Time: 02:18 [ET]
Event Date: 06/15/2006
Event Time: 22:23 [EDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
BRIAN BONSER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | M/R | Y | 45 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP DUE TO DIGITAL ELECTRO-HYDRAULIC LEAK
"On 6/15/06 at 2223 hrs an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 45% power due to severe DEH Leak on the #1 Throttle Valve. DEH Leak ceased upon Turbine Trip. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed without contingencies and Unit 2 was stabilized in Mode 3. All control rods fully inserted and no S/G Safety Valves Lifted. Feedwater to the S/G was supplied by the main FW Pumps. All safe shutdown equipment operated as expected. There were no major equipment failures."
Decay heat is being removed with main feedwater and dumping steam to the condenser. The grid is stable. The fire brigade was activated following the trip and set a fire watch (no fire).
The NRC Resident Inspector was notified of this event by the licensee. | |