Abstract
No matter how the nuclear reactors are designed for safety. The accidend still occur. After the Three Mile Island accident, the nuclear industry performed a variety of serve accident research to understand the severe accident phenomena and to use severe accident code for improving the severe accident sequences prediction capability. The loss of coolant accident (LOCA) is one of the most important design basis accidents in the nuclear power plant. If accidents occur in the NPP, the operators must try to find out the abnormal plant conditions by observing some parameters. With the partial information and short time, the operators must to use the severe accident sequences prediction which accurately predict the progression of the accidents in the initial stage. During a LOCA, it is important for the operators to find out where the break location is and how large the break size is in order to effectively compare behavior for LOCA. In this paper, pipe breaks inside primary containment (recirculation line break, LOCA) and pipe breaks outside containment (Steam line break accident, LOCA) are analysis from PCTran Chin Shan,which are compare with Final safety analysis report (FSAR). .This research work, use PCTran ChinShan Model to compare behaviors for LOCA of different break sizes and the transient characteristics of the plant parameters, inculding the system pressure, temperature, and water level, etc.. are obtained by performing the PCTran simulations. Pipe breaks inside primary containment which the accident under control early by the biggest break occur in the NPP, ECCS are normal start and down to temperature and pressure which is smilliar to FSAR. Pipe breaks outside containment for LOCA of different break sizes in the NPP, MSIV is closed at the same time which is smilliar to FSAR, but the behavior for LOCA is not enough to ECCS are normal start. Another, in this paper, The effect which MSIV is opened, the heat was brought form the NPP.