| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/03/2006 - 03/06/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42252 | Facility: SAN ONOFRE
Region: 4 State: CA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: JULIE HOLT
HQ OPS Officer: JEFF ROTTON | Notification Date: 01/11/2006
Notification Time: 20:43 [ET]
Event Date: 01/11/2006
Event Time: 12:48 [PST]
Last Update Date: 03/03/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
REBECCA NEASE (R4)
MICHAEL MAYFIELD (NRR)
|
| 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 SHUTDOWN COOLING INOPERABILITY DUE TO SMALL DISCHARGE HEADER LEAK
"On Wednesday, January 11, 2006, at approximately 1248 PST, with San Onofre Unit 2 in Mode 5 (loops not filled), Southern California Edison (SCE) declared both trains of the Shutdown Cooling (SDC) System inoperable. This action was taken following discovery and evaluation of an approximate 1.5 inch crack and minor leak (about 10 drops per minute) in a line located at the Low Pressure Safety Injection (LPSI) common discharge header (the LPSI pumps are also used as the SDC system pumps). Although both trains of the SDC system remain functional and one train is in service, SCE declared both SDC trains inoperable because the affected pipe might not be ASME Code qualified as a result of this crack.
"SCE is conservatively reporting this event in accordance with 10CFR50.72(b)(3)(v)(B) as a condition that could prevent the remove residual heat. SCE is continuing to evaluate this condition.
"SCE is following the actions required by Technical Specifications 3.4.8, and will repair the pipe crack during the current refueling outage, after the reactor core has been off-loaded to the spent fuel pool. SCE expects to place the plant in Mode 6 tomorrow.
"At the time of this report, Unit 2 is in Mode 5 and Unit 3 is in Mode 1 at 100% power. The NRC Senior Resident Inspector, NRC Region IV, and the Nuclear Reactor Regulation Project Manager have been briefed of this event and will be provided with a copy of this report."
Shutdown cooling has been running for approximately one week and the leak was initially discovered sometime on 01/10/06 during an inspection in which Boric acid accumulation was discovered . A work request was written to perform a more thorough inspection that was performed on 01/11/06. The crack is on an 8 inch line that returns to RCS Loop 2A which is normally isolated during high pressure shutdown conditions and reactor power operation.
* * * RETRACTION FROM C. WILLIAMS TO W. GOTT AT 2039 EST ON 3/3/06 * * *
"On January 11, 2006, Southern California Edison (SCE) reported to the NRC that both trains of Shutdown Cooling System (SDC) were inoperable at San Onofre Unit 2 due to a through-wall crack in a pipe in the common discharge header. That occurrence was reported to the NRC in accordance with 10CFR50.72(b)(3)(v)(B) for a condition that could prevent the removal of residual heat.
"SCE has since analyzed the pipe integrity at the design basis conditions and determined the pipe would have remained intact and that system leakage through the crack would have remained below the allowable system leak rate. Based on these results, SCE concluded that the pipe was capable of performing its safety functions under design basis conditions. Because the system was always capable of performing its safety function, SCE is retracting the January 11, 2006 report. Nevertheless, SCE plans to submit a voluntary licensee event report to document this event and inform the NRC of the corrective actions taken.
"At the time of this phone call, Unit 2 was in the middle of the Cycle 14 refueling outage and Unit 3 was at about 100 percent power. SCE has notified the NRC resident inspectors about this retraction and will provide them with a copy of this report."
Notified R4DO (J. Whitten) |
| Power Reactor | Event Number: 42385 | Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: RICH KLINEFELTER
HQ OPS Officer: PETE SNYDER | Notification Date: 03/04/2006
Notification Time: 05:34 [ET]
Event Date: 03/04/2006
Event Time: 05:17 [EST]
Last Update Date: 03/04/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
CHRISTOPHER CAHILL (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Startup | 0 | Startup | Event Text CONTROL RODS DECLARED INOPERABLE DURING SHUTDOWN
"On 03/03/06, Susquehanna Unit 1 operators began the process of shutting down the unit for the 14th Refueling and Inspection Outage. It was expected that during this evolution, 'slow to settle' control rods would be encountered, and rather than delay the continuance of the shutdown to perform operability testing, it was determined that these control rods would conservatively be declared inoperable. Based on previous testing data and trending analysis of operability testing performed during the cycle, PPL believes that the subject control rods would have passed operability tests.
"Accordingly, at 0517 EDT, Technical Specification 3.1.3.f was entered which requires the unit to be taken to Mode 3 in 12 hours when nine or more control rods are inoperable. The inoperable control rods are fully inserted and disarmed, shutdown margin requirements are met, and the control rod system was always fully capable of performing its safety function.
"At the time when the ninth control rod was declared inoperable, Unit One was operating at 0% Power. Unit Two is continuing operation at 100% power.
"PPL is reporting this event as a Technical Specification Required Shutdown per 10CFR 50.72(b)(2)(i). The NRC Resident Inspector was notified by the licensee." |
| Power Reactor | Event Number: 42386 | Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [ ] [2] [ ]
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: CHRIS SKINNER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 03/04/2006
Notification Time: 14:51 [ET]
Event Date: 03/03/2006
Event Time: 18:00 [EST]
Last Update Date: 03/04/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
CHRISTOPHER CAHILL (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 91 | Power Operation | 91 | Power Operation | Event Text REACTOR BUILDING VENTILATION FLOWPATH INOPERABLE
"This notification is being made in accordance with License Condition 2.F for Nine Mile Point Unit 2 which states in part 'report any violations of the requirements contained in Section 2.C of this license in the following manner: initial notification shall be made within 24 hours to the NRC Operations Center via the Emergency Notification System, with written follow-up within 30 days in accordance with the procedures described in 10 CFR 50.73(b), ( c), and (e).' License Condition 2.C (2) states in part 'Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.' This notification describes a licensee identified condition where both redundant Standby Gas Treatment (SGT) trains were apparently inoperable in violation of Technical Specifications. The condition has been corrected.
"At 1800 on 3 March 2006, while operating at 91 % power (coast down to refueling), Nine Mile Point Unit 2 identified a condition in which both trains of SGT were apparently rendered inoperable for intermittent time periods of a few hours in length, starting from 17 February 2006 through about 1900 on 28 February 2006. This was not recognized at the time; as such the requirement to initiate a plant shutdown per LCO 3.0.3 was not performed. The condition was caused by use of a heavy-duty tarp and associated cargo net supporting it from underneath, installed across the Unit 2 Reactor Building Hoist Well. Installation of the tarp and net across the hoist well occurred between the above dates, for time periods of a few hours each, in order to support refueling preparations, thereby avoiding spread of contamination during rigging activities. The blockage of the Reactor Building Hoist Well would have obstructed or significantly degraded the design flow paths of both trains of SGT if called upon to perform their safety functions in a design basis accident. Therefore current information indicates this tarp installation configuration renders the SGT system inoperable.
"The tarp and net were permanently removed at about 1900 on 28 February 2006 when a supervisor questioned if tarp installation was allowed while the reactor was at power.
"Although removal of the tarp and net typically required only a few minutes effort by plant workers, its installation and continued blockage of the ventilation flow path would have resulted in declaration of the [SGT] safety systems to be inoperable if not removed. Workers were not sensitive to the safety function of the open ventilation flow path provided by the hoist well, and no programmatic training or administrative requirements were identified which prohibited the configuration at conditions other than cold shutdown.
"Nine Mile Point is in the process of taking compensatory measures to preclude installation of the tarp when SGT is required to be operable by briefing operations, radiation protection and refuel worker crews on ventilation requirements and sensitivity to safety functions. Operations took physical control (lock and key) of the tarp. Corrective actions to provide administrative controls on the tarp installation are in process. Detailed evaluation of the safety significance of the condition is ongoing.
"Initial review of plant records indicate that this configuration was also installed intermittently around July 2003. It was recognized as undesirable but was not identified as an operability or reportability issue at the time. Corrective action to prevent its recurrence was not effective. More detailed information on specific dates and durations when this configuration existed will be provided in the 30 day written LER report, after a detailed review.
"The instances noted above and any similar conditions identified will be explained in detail in the follow-up LER that will be submitted as required by 10CFR 50.73(a)(2)(i)(B) - 'Any operation or condition which was prohibited by the plants Technical Specifications .'"
Unit-1 is not affected since they have a different method to control use of this tarp. The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42387 | Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [ ] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: TIM GAFFNEY
HQ OPS Officer: ARLON COSTA | Notification Date: 03/05/2006
Notification Time: 11:12 [ET]
Event Date: 03/05/2006
Event Time: 07:10 [MST]
Last Update Date: 03/05/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
JACK WHITTEN (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR TRIP DUE TO LOW DEPARTURE FROM NUCLEATE BOILING RATIO.
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73.
"On March 5, 2006 at approximately 0710 MST Palo Verde Unit 3 experienced a reactor trip (RPS actuation) from 100% rated thermal power due to low departure from nucleate boiling [ratio] (DNBR) trips on all four channels of the core protection calculators (CPCs). The unit was at normal temperature and pressure prior to the trip.
"Prior to the reactor trip, at approximately 0704 MST, a CEAC [Control Element Assembly Calculator] #1 sensor fail alarm was received. While investigating the alarm, at 0710 MST, a control element assembly (CEA) deviation alarm for CEAC #1, all four CPC channel sensor fail alarms, and a CEA withdrawal prohibit alarm were received. The reactor tripped six seconds later. A CEA calculator (CEAC) fail alarm was received on CEAC #1. The apparent cause is presently suspected to be a failure of CEAC #1. An investigation has commenced to determine the root cause of the reactor trip.
"All of the control rods fully inserted into the core. Four of eight steam bypass control valves quick opened, per design, directing steam flow to the condenser. No main steam or primary relief valves lifted and none were required. There was no loss of heat removal capability or loss of safety functions associated with the event. Electrical buses transferred to offsite power as designed. The Shift Manager determined this event was an uncomplicated reactor trip. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event nor contributed to the event.
"Unit 3 is stable at normal operating temperature and pressure in Mode 3. No ESF actuations occurred and none were required. The event did not result in any challenges to the fission product barrier or resulted in any releases of radioactive materials. There were no adverse safety consequences or implications as a result of this event. The event did not adversely affect the safe operation of the plant or health and safety of the public."
The licensee notified the NRC Resident Inspector. | |