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 upward or downward in increments. This sequential current control drive system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or known also as the Rod Control System (RCS). The purpose 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.
References
Tanaka, A., et al., Development of a 3-D simulation analysis system for PWR control rod drive mechanism. International Journal of Pressure Vessels and Piping, 2008. 85(9): p. 655-661.
Rajan Babu, V., G. Thanigaiyarasu, and P. Chellapandi, Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action. Nuclear Engineering and Design, 2014. 278(0): p. 601-617.
[3] Shirazi, S. A. M., The simulation of a model by SIMULINK of MATLAB for determining the best ranges for velocity and delay time of control rod movement in LWR reactors. Progress in Nuclear Energy, 2012. 54(1): p. 64-67.
[4] Yoon, K. H., et al., Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant. Nuclear Engineering and Design, 2009. 239(10): p. 1857-1861.
[5] Singh, N. K., D. N. Badodkar, and M. Singh, Numerical and experimental study of hydraulic dashpot used in the shut-off rod drive mechanism of a nuclear reactor. Nuclear Engineering and Design, 2014. 273(0): p. 469-482.
[6] Rajan Babu, V., et al., Testing and qualification of Control & Safety Rod and its drive mechanism of Fast Breeder Reactor. Nuclear Engineering and Design, 2010. 240(7): p. 1728-1738.
[7] Chyou, Y. P., D. D. Yu, and Y.-N. Cheng, Performance validation on the prototype of control rod driving mechanism for the TRR-II project. Nuclear Engineering and Design, 2004. 227(2): p. 195-207.
[8] Ramesh, E. and S. Usha. Reliability analysis of control rod drive mechanisms of FBTR for reactor startup and power control. in Reliability, Safety and Hazard (ICRESH), 2010 2nd International Conference on. 2010.
[9] Iida, H., et al., Long-term stability of Sm2Co17-type magnets for control rod drive mechanism (CRDM) in a nuclear reactor. Magnetics, IEEE Transactions on, 1995. 31(6): p. 3653-3655.
[10] Mitsubishi Heavy Industries, L. US-APWR Technical Report. FMEA of Control Rod Drive Mechanism. Control System. 2007.
[ 11] Institute, E. P. R. Nuclear Maintenance Applications Center: Westinghouse Full-Length Rod Control System - Life Cycle Management Planning Sourcebook. 2006.
[12] Roslund, C. J., et al., Digital nuclear control rod control system. 2012, Google Patents.
[13] Ericson, C. A., Hazard Analysis Techniques for System Safety. 2005, New Jersey: John Wiley & Sons.