05000425/LER-2015-001
| Vogtle Electric Generating Plant - Unit 2 | |
| Event date: | 03-14-2015 |
|---|---|
| Report date: | 05-13-2015 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4252015001R00 - NRC Website | |
Reported lessons learned are incorporated into the licensing process and fed back to industry Send comments regarding burden estimate to the FOIA, Pnvacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection
A. REQUIREMENT FOR REPORT
This report is required per 10CFR50.73(a)(2)(iv)(A) due to an unplanned automatic actuation of the Reactor Protection System (RPS), Containment Isolation Valves, Emergency Core Cooling System (ECCS), Auxiliary Feedwater System (AFW), and Emergency Diesel Generators (EDGs).
B. UNIT STATUS AT TIME OF EVENT
Mode 1, 100 percent power.
C. DESCRIPTION OF EVENT
On March 14, 2015, Unit 2 reactor tripped at approximately 0429 EDT due to a Safety Injection (SI) / Steam Line Isolation (SLI) actuation. The initiating event was closure of the Loop 3 outboard Main Steam Isolation Valve (MSIV). The valve closed as indicated on the Integrated Plant Computer (IPC) with steam flow lowering from normal to off scale low in approximately 4 seconds. The SI / SLI actuation was initiated due to lowering Loop 2 steam generator pressure. As the MSIV closed, steam flow from loop 3 decreased and the other three steam generator's steam flows increased to continue supplying steam to the main turbine. Loop 2, which is closest in physical proximity to Loop 3, had the largest steam flow increase. The increased steam flow resulted in a reduction in steam generator pressure. The reduction in steam generator pressure in Loop 2 over a few seconds was sufficient to initiate a rate compensated Low Main Steam Line Pressure SI and SLI.
The SI and SLI resulted in a Reactor Trip, Turbine Trip, Containment Isolation, Emergency Core Cooling System (ECCS) actuation, AFW actuation, and start of the EDGs. Due to the Steam Line Isolation the condenser was not available to remove decay heat. As designed, the Atmospheric Relief Valves (ARVs) opened and decay heat was discharged to the atmosphere, with feedwater supplied from the Condensate Storage Tanks (CSTs) and through the AFW pumps to the Steam Generators and to the main steam lines.
Operators responded immediately using emergency response procedures to evaluate the event and take appropriate actions. The SI was terminated in an appropriate time frame and the unit was stabilized in Mode 3 at normal Reactor Coolant System no-load temperature and pressure.
D. CAUSE OF THE EVENT
The cause of the event was failure of the Loop 3 outboard MSIV hydraulic dump valve and subsequent loss of hydraulic control pressure. The loss of hydraulic control pressure to the MSIV resulted in the valve closure.
E. SAFETY ASSESSMENT
The MSIV closing created a low pressure condition in the unaffected Steam Generators. This pressure reached the rate compensated setpoint for a Low Steam Line Pressure Safety Injection and Steam Line Isolation and RPS actuation. All systems responded as required, including all control rods fully inserting into the core, Containment Isolation signal generation and subsequent closure of all valves, startup of ECCS systems, start of the EDGs, and AFW system start up and supply to the Steam Generators with excess heat rejecting to the environment through the ARVs.
This system response to the event is as designed for SI and SLI actuation.
Because all systems responded as required and there were no actual adverse effects on the health and safety of the public, the safety significance of this event is considered very low.
F. CORRECTIVE ACTION
The solenoid valve was replaced and post maintenance testing was performed satisfactorily.
G. ADDITIONAL INFORMATION
1) Failed Components:
Loop 3 Outboard Main Steam Isolation Valve hydraulic dump valve solenoid 2) Previous Similar Events:
None found 3) Energy Industry Identification System Code:
[JC] — Reactor Protection System [BA] — Auxiliary Feedwater [SB] — Main Steam System