Investigation Of Rod Control System Reliability Of Pwr Reactors

Abstract

The Rod Control System is employed to adjust the position of the control rods in the reactor core which corresponds with the thermal power generated in the core as well as the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive employs magnetic stepping-type mechanism. This type of mechanism consists of a pair of circular coils and latch-style jack with the armature. When the electric current is supplied to the coils sequentially, the control-rods, which are held on the drive shaft, can be driven up ward or down ward in increments. This sequential current c ontrol drive system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or known also as the Rod Control System (RCS). The p urpose of this paper is to investigate the RCS reliability of APWR using the Fault Tree Analysis (FTA) method since the analysis of reliability which considers the FTA for common CRDM can not be found in any public references. The FTA method is used to model the system reliability by developing the fault tree diagram of the system. The results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.