05000341/LER-2007-002
| Docket Number11 15 2007 2007 -0002 -000 01 14 2008 05000 | |
| Event date: | 11-15-2007 |
|---|---|
| Report date: | 01-14-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| ENS 43784 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| 3412007002R00 - NRC Website | |
Initial Plant Conditions:
Mode� 2 Reactor Power� 9 percent
Description of the Event
At 03:13 EST November 15, 2007, during the restart of the reactor from the twelfth refueling outage, an initiation of the alternate rod insertion/recirculation pump trip (ARI/RPT) system occurred due to an invalid reactor water level 2 (Level 2) signal from the Division 1 reactor water level instrumentation. A safety tag-out was needed on a valve external to the reactor water level reference leg backfill panel to ensure isolation of the backfill system during work on the panel. That valve was already closed as the Division 1 backfill system was not in operation at the time. The tag-out activity required an operator to verify that a system isolation valve, a small T-handle valve with no indication, was closed. When performing the tag-out, the assigned operator first verified that the isolation valve was closed by attempting to operate it in the closed direction to verify that the valve was closed. However, he was not confident the valve was closed. To assure himself that this valve was closed, the operator inappropriately cracked the isolation valve open and then immediately closed the valve. This action caused -a short pressure perturbation on the level instrumentation reference leg that initiated a false reactor water Level 2 Division 1 ARI/RPT system initiation, half-scram, and a reduction in feedwater flow [SJ].
The ARI/RP'T system functioned as designed. Within approximately 10 seconds of the ARI/RPT trip, the control room operators placed the reactor mode switch to shutdown inserting a manual scram in response to the trip of the recirculation pumps initiated by the ARI/RPT trip as directed by plant procedures. All control rods fully inserted into the core.
The main turbine [TA] was not in operation at the time of the event, and the turbine bypass valves were handling the reactor steam flow before and immediately following the reactor scram. At the time of the scram all emergency core cooling systems (ECCS) and Emergency Diesel Generators (EDGs) [DG] were operable. Due to the short duration of the false water level signal, the High Pressure Coolant Injection (HPCI) [BJ] and Reactor Core Isolation Cooling (RCIC) [BN] systems functioned normally and did not automatically initiate, nor was an automatic reactor scram initiated.
The lowest reactor vessel water level reached was 183 inches which is above reactor water level 3 (Level 3), and the standby feedwater pumps were placed in service to maintain reactor water level. With water being supplied by the standby feedwater system and with an increase in flow from the north reactor feedwater pump, a high level alarm and HPCl/RCIC reactor water level 8 (Level 8) trips were subsequently received. The operating north reactor feed pump was manually tripped to keep from overfeeding the vessel. Since there was little decay heat available coming out of the refueling outage, the main steam isolation valves (MSIV's) were manually closed at 03:49 hours to control the cool down rate.
Plant equipment was determined to respond as expected to the invalid low level signal and subsequent manual reactor scram.
This event is being reported under 50.73(a)(2)(iv)(A), as an event or condition that resulted in manual actuation of the reactor protection system (RPS) [JD] that resulted in a reactor scram. Immediate notification was made to the NRC in accordance with 10 CFR 50.72 at 06:34 ET on November 15, 2007 (EN 43784).
Cause of the Event
Two root causes have been identified for this event. The nuclear operator performing the tag-out of the isolation valve for the Division 1 reference leg backfill system departed from training and procedural requirements for checking valve position. By opening B2100F241A, even fractionally, a pressure transient in the Division 1 reactor water level instrumentation reference line led to an initiation of Level 2 ATWS/ARI, a trip of both recirculation pumps and insertion of control rods. It was also determined that the pre-job brief was not adequate in that a human performance assessment of the task was not performed.
Analysis of the Event
The reactor responded as designed to the false reactor water Level 2 ARI/RPT signal. The ARI signal sealed in and energized the ARI solenoid valves which vented the scram air supply lines as designed. Both recirculation pumps [AD] and recirculation pump field breakers tripped in response to the ARI/RPT signal as designed.
Expected primary containment isolation [JM] signals and actuations or partial actuations also occurred for the Division 1 Level 2 signal as expected, including isolation groups 2, 10, 12, 14, 16, 17, and 18.
In response to the false Level 2 signal, the control center heating and ventilation and air-conditioning (CCHVAC) systems shifted into recirculation mode, the reactor building ventilation system tripped, Division 1 standby gas treatment system automatically started, the Division 1 control air compressor automatically started, and secondary containment isolation dampers closed. All of these are expected actuations from a reactor water Level 2 signal.
The false water level transient was very short in duration. Only the RPS Al trip system sensed the false level transient. Due to differences in RPS trip system response times, it was determined that the pickup of the Al channel without the pickup of the corresponding B I channel in response to a very short water level transient can be expected. These channel response times meet the overall RPS trip system response time requirements.
Therefore, the reactor water Level 3 scram and related isolations did not occur because only RPS Al channel tripped which can be expected as a result of the short false water level transient.
The HPCI and RCIC systems functioned normally and did not start in response to the false Level 2 signal. The duration of the Level 2 signal was determined to be too short to ensure the pick up and seal in of the system actuation relays for HPCI and RCIC systems. A signal was also sent to low pressure coolant injection (LPCI) loop select logic, but the loop selection functioned normally and did not complete for the same reason. Therefore, as a result of the short false water level transient, automatic startup of these systems did not occur and was not expected.
The plant response to the false level 2 signal was as expected and was enveloped by the recirculation pump trip transient described in the UFSAR. There was no challenge to the integrity of the reactor coolant system or the main steam system. The lowest reactor water level during the transient was measured to be 183 inches above top of active fuel which is above the reactor water Level 3 isolation and trip setpoint.
The highest reactor pressure occurred prior to the event and at 934 psig is below the safety relief valve (SRV) setpoints, and no SRVs operated. Subsequent to the reactor trip, reactor pressure was adequately controlled using the main turbine bypass valves and trended downward. Reactor water level was controlled using the standby feedwater system. Since the reactor was only operating at 9% power prior to the event, and with little decay heat available due to the limited operating history after the refueling outage, the transient was mild. Since the ARI/RPT and reactor protection systems performed as designed, and the plant response was enveloped by the UFSAR transient analyses, there were no safety consequences as a result of this event.
Corrective Actions
The plant manager performed a human performance stand-down session for all plant employees to discuss this and other human performance events and expectations. The Operations Manager also met with each of the Operations shifts and set expectations for the use of human performance techniques as a means to reduce human errors.
This event is documented and evaluated in the Fermi 2 corrective action program. Other actions are being planned to address this event. These actions will be tracked and implemented by the corrective action program.
Additional Information
A. Failed Components: None.
B. Previous LERs on Similar Problems:
There have been no events involving perturbations in a reactor water level reference leg in the last 5 years.