05000346/LER-2003-013
| Event date: | |
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| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3462003003R00 - NRC Website | |
DESCRIPTION OF OCCURRENCE:
At the time of this event, the Davis Besse Nuclear Power Station (DBNPS) procedure DB-OP-06903, "Plant Shutdown and Cooldown," permitted plant cooldown evolutions with Control Rod Group 1 (AA) either fully withdrawn or inserted.
The operating crew desired to have Control Rod Group 1 withdrawn during cooldown because it provided instantaneous negative reactivity in the event a reactivity problem was identified. The design of the Reactor Protection System (RPS) (CC) is such that, in order to support this configuration at lower pressures, a Shutdown Bypass High Pressure Trip is inserted that will actuate the RPS and insert the rods at a Reactor Coolant System (RCS) LAB) pressure setpoint of approximately 1812 psig.
On September 30, 2003, with the plant in Mode 3 and following completion of a normal operating pressure test using non-nuclear heat, operators were performing a plant cooldown. The RCS was at approximately 1750 psig and 532 degrees F.
Control Rod Group 1 (trippable reactivity) was withdrawn at approximately 2113 hours0.0245 days <br />0.587 hours <br />0.00349 weeks <br />8.039965e-4 months <br />. At approximately 2125 hours0.0246 days <br />0.59 hours <br />0.00351 weeks <br />8.085625e-4 months <br /> cooldown of the RCS commenced by opening the Turbine Bypass Valves (TBVs) (.711. Over the next several minutes, a decrease in RCS pressure and Pressurizer (PZR) [PZR1 level was observed. The TBVs were throttled back but not closed, and PZR heater bank 3 was energized. The RCS responded by increasing in pressure. At approximately 2134 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.11987e-4 months <br />, this increasing pressure resulted in an unplanned Reactor Trip Signal on Shutdown Bypass High Pressure which inserted the withdrawn Control Rod Group 1.
The RPS performed as designed by opening all four reactor trip breakers and subsequently de-energizing the control rod drive mechanisms. RCS inventory and pressure/temperature limits were maintained within allowable values. There were no post-trip response equipment issues identified. There were no equipment safety concerns identified. There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.
The immediate operational actions in response to the event included:
Stabilizing plant conditions and ceasing RCS cooldown, Returning to normal operating temperature and pressure, and Performing a crew stand down to address Operations plant control.
APPARENT CAUSE OF OCCURRENCE:
The apparent causes of this event were less than adequate operator performance, and less than adequate procedure guidance.
Prior to commencing the plant cooldown, simulator training and a pre-job briefing were conducted. The simulator training was conducted with Control Rod Group 1 inserted, and reactor core decay heat levels were programmed higher APPARENT CAUSE OF OCCURRENCE (continued):
than those that actually existed because the extended outage had allowed more time for core heat to decay. The actual cooldown was performed with Control Rod Group 1 withdrawn. The pre-job briefing did not include a discussion of the RPS Shutdown Bypass High Pressure trip or limiting setpoints.
Prior to RPS actuation, plant cooldown was commenced by increasing the TBV demand by about five percent over the steady state demand. The demand increase was based on the experience from the simulator training that was performed to support the cooldown evolution. However, the decay heat load programmed into the simulator was higher than the actual decay heat load of the RCS. The repositioning of the TBVs resulted in a higher than expected cooldown rate.
During the transient, the operator lowered the demand signal on the TBVs but did not completely shut the valves. The higher than expected cooldown rate lowered RCS pressure and PZR level. To compensate for this system response, PZR heaters were energized, the PER level controller setpoint was increased, and demand on the TBVs was reduced. This resulted in increasing RCS pressure and ultimately an RPS actuation occurred on Shutdown Bypass High Pressure.
During the cooldown, control room operators were not specifically monitoring the highest indicated RCS pressure. The focus of operator attention was on the Safety Features Actuation System (SFAS) [41E] inputs for margin to the SPAS RCS Low Pressure trip, not the RPS. The Shift Engineer did not check the readings for the RPS pressure points to determine control margin.
Subsequent to the event, it was determined that procedure DB-OP-06903, "Plant Shutdown and Cooldown," did not contain adequate guidance to aid operator knowledge of the RPS Shutdown Bypass High Pressure trip.
ANALYSIS OF OCCURRENCE:
Prior to the RPS actuation, a normal operating pressure test was being conducted using non-nuclear heat. The RPS performed as intended by opening all four reactor trip breakers and subsequently de-energizing the control rod drive mechanisms. Reactor coolant inventory and pressure/temperature limits were maintained within allowable values. There were no post-trip response equipment issues identified. There were no equipment safety concerns identified.
Therefore, this event was of minimal safety significance.
An 8-hour notification of this event was made to the NRC on September 30, 2003, pursuant to 10 CFR 50.72(b)(3)(iv)(A) (Event No. 40208). This report is submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A), an event that resulted in actuation of the Reactor Protection System.
CORRECTIVE ACTIONS:
The individual human performance issues associated with this event were addressed in accordance with company policy and procedures.
Procedure DB-OP-06903, "Plant Shutdown and Cooldown," was revised on October 2, 2003, to provide more prescriptive guidance for the plant cooldown operation. A further revision was issued on November 18, 2003, to incorporate a requirement for a pre-job briefing that includes specific operating experience gained from this event.
This event was discussed with on-shift operating personnel and included guidance on management expectations when differences between training and actual plant conditions are encountered. This discussion was conducted by the Operations Manager between October 23 and November 7, 2003.
A case study of this event is to be developed with a focus on constant cognizance of proper operating envelope and the need to continually monitor multiple indications for each parameter. This case study will be used as part of the initial and continuing training programs. The development of this case study will be completed by January 31, 2004.
With the occurrence of this event, senior management had several recent events evaluated for extent of condition. This evaluation identified that organizational weaknesses and Operations Department shortcomings existed. As a result of this recognition, an Operations Improvement Action Plan has been developed and restart designated items are being implemented prior to restart to improve individual, program and procedure, management, and independent oversight.
FAILURE DATA:
The DIME'S has experienced no unintended actuation of the RPS resulting in a reactor trip in the last three years. However, on September 15, 2003, while heating up the RCS for the NOP test, Core Flood Valve CF1B unexpectedly opened and discharged to the Decay Heat System lifting one or more relief valves which discharged to the Reactor Coolant Drain Tank. While this event was not reportable, its root causes were similar. They were procedure deficiencies and less than adequate pre-job briefing. Corrective actions included discussing the event with Operations personnel and future inclusion of the event in training lesson plans This event is documented in CR 03-07746. The corrective actions for the CF1B event were too specific to the event to have reasonably precluded the RPS actuation and plant trip event. However, as noted above, an Operations Improvement Action Plan has been developed and is being implemented to improve operational performance from a programmatic perspective.
Energy Industry Identification System (Elis) codes are identified in the text as [Xa].
NP-33-03-013-00 � CR 03-08374