05000482/LER-2010-012
| Wolf Creek Generating Station | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4822010002R00 - NRC Website | |
A reactor trip occurred during the startup on October 17, 2010. Wolf Creek Generating Station (WCGS) was coming out of a two week forced outage. All major equipment was available for the startup. Due to the length of the shutdown, the reactor core was xenon free and had low decay heat. Criticality was achieved at 1816 CDT on October 16, 2010. Main feedwater pump 'B' [EllS Code: SJ-P] was placed in service at 2320 CDT and power was raised to 7% power. Turbine warming was completed at 0854 CDT on October 17, 2010, which allowed reactor power to be increased to 15%. At 0912 CDT, the turbine [El IS Code: TA] was rolled to 1800 rpm in preparation kir synchronization to the grid. At 0922 CDT, the reactor was at approximately 15% power.
Though feedwater pre-heating was in service, feedwater temperature decreased steadily during the power increase, initiating divergent Steam Generator (SG) level [EllS Code: JB] oscillations. The Control Room operators took manual control of the 'B' bypass feed regulating valve [EllS Code: JB- LCV] but were unable to dampen the level oscillations. SG level in the 'B' generator reached the High-High level setpoint of 78% at 0952 CDT, which resulted in a turbine trip and feed water isolation signal (FWIS). Motor-driven auxiliary feedwater pumps (AFW) [EllS Code: BA-P] started, but their combined feed capacity was insufficient to maintain the current power level. The Control Room operators began to reduce reactor power to stay within the capacity of the Auxiliary Feed Water System by inserting control rods in manual. Recognizing that the power reduction could not be accomplished in time, the Control Room supervisor ordered the reactor be manually tripped. At 0953 CDT on October 17, 2010 the reactor automatically tripped [EllS Code: JE] on SG 'A' Low-Low level of 23.5%, one second earlier than the manual trip.
Emergency response procedures were entered after the reactor trip. Due to a low initial decay heat load, full AFW flow, and pre-heating steam loads, Reactor Coolant System (RCS) average temperature fell below 550 degrees Fahrenheit. Emergency Boration [EllS Code: CB] was initiated at 1001 CDT. The Control Room operators took additional action to terminate the cooldown by isolating major steam loads and reducing AFW flow. At 1048 CDT, operators took additional action to eliminate the cooldown and closed the Main Steam Isolation Valves [El IS Code: SB-ISV]. With all steam loads isolated, RCS temperature recovered and the operating crew stabilized the plant at hot standby. The minimum RCS temperature reached was 538.7 degrees Fahrenheit. Emergency boration was secured at 1059 CDT. No safety limits were challenged or exceeded during the event. The unit was successfully restarted on October 19, 2010 after a 27-hour delay due to the reactor trip.
_a The reactor trip and actuation of ESFAS instrumentation actuation described in this event is reportable per 10 CFR 50.73(a)(2)(iv)(A), which requires reporting of "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section.
Paragraph (B)(1) of 10 CFR 50.73(a)(2)(iv) includes "Reactor protection system (RPS) including:
reactor scram or reactor trip." Paragraph (B)(6) of 10 CFR 50.73(a)(2)(iv) includes "PWR auxiliary or emergency feedwater.
ROOT CAUSE
Control room operators were unable to maintain SG levels at low poWer using the main feed regulating bypass valves in automatic or manual control eventually over feeding the 'B' SG. As a result the turbine tripped on High-High SG level, which initiated a FWIS. The basis for the operators inability to control SG levels is provided below.
Guidance in plant operational procedures was not aligned with the required plant design parameters for low power operations, specifically in controlling feed water preheating, SG level control and response to a FWIS. As a result, the operation of the plant during power ascension was outside the main feedwater bypass valve optimum operating region and the feedwater preheating limitations.
Main feedwater bypass valve characteristics and SG level process control settings did not provide stable (convergent) operating characteristics during low power operations. As a result, there was an over reliance on manual feedwater control and individual operator experience to mitigate SG level oscillations.
CORRECTIVE ACTIONS
Procedures SYS AE-121, "Turbine Driven Main Feedwater Pump Startup," and GEN 00-003, "Hot Standby to Minimum Load," were revised to incorporate the optimal range for operating the main feedwater regulating valves and regulating bypass valves. This range will provide the most consistent valve flow response for a given change in valve position.
A more appropriate startup and shutdown sequence will be determined to remain within the design operational parameters of the feedwater preheating system and Feedwater bypass valves.
Appropriate procedures will be revised as needed.
The safety significance of this event is low. This event is analyzed as reported in WCGS Updated Safety Analysis Report (USAR) Section 15.2.7, "Loss of Normal Feedwater Flow." Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the reactor coolant system, or the steam system, since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves.
There were no adverse effects on the health and safety of the public.
OPERATING EXPERIENCE/PREVIOUS SIMILAR OCCURRENCES
Mode 3. The cause of the event was inadequate monitoring of critical operating parameters.
feedwater (MFW) pump. The failed transfer of an inverter to its alternate power supply caused the MFW pump trip.
room operators manually tripped the reactor due to decreasing SG levels. The cause of the MFW pump trip was a failed servo in the MFW control circuitry.