عنوان مقاله [English]
The loss of coolant accident is due to the reduction of coolant fluid volume in the primary circuit. The direct cause of this accident is the mechanical failure or fatigue of material of the components of the primary circuit components during power plant operation. The accident, which is a design-based incident, is an important factor in assessing the safety of a nuclear power plant. If the break occurs in the main circuit of the primary circuit with a diameter greater than 25% of the cross-section area, it shall be referred to as a large break. In this paper, this accident with break diameter of 850 mm is modeled and analyzed using the TRACE code in a VVER-1000 reactor. The TRACE is specifically designed for coolant loss accidents. By this analysis, in contrast to conservative assumptions in the reactor safety evaluation, the best estimation of the reactor safety can be achieved and significant economic considerations can be obtained. Finally, the results of TRACE code are compared with the final safety analysis report of the power plant as well as previous research by the RELAP5. The results indicate the accuracy of the TRACE in modeling the large break accident.