Control Rod Drop Accident Analysis Based on Three-dimensional Coupling Analysis Code

The challenge for future nuclear fuel management research is to deal with the problems caused by the high nuclear fuel burnup, high power factor, and high enrichment of fuel assemblies. On the other hand, the increasingly strict license safety review of design methods and design limits also place ever more demanding requirements. This makes the traditional lowdimensional analysis method that introduces too conservative assumptions seem inadequate. Therefore, it is necessary to develop a new three-dimensional method so that all global and local parameters of the nuclear steam supply system (NSSS) can be coupled in the simulation of the accident transient, including neutrons parameters, thermal hydraulic and system parameters of primary and secondary loop, etc. As an essential physical and thermal part of the numerical reactor, a three-dimensional coupling analysis code was developed. Through the dynamic link library technology and based on the independently developed system thermal hydraulic code GINKGO and three-dimensional core code COCO, coupling code GINKGO/COCO was developed. The coupling code is verificated by benchmark TMI-MSLB exercise 2. By comparing the parameters of steady-state core parameters, transient power and reactivity, the results show that the coupled calculation model of the developed code is correct, and the calculation has good accuracy and reliability. In this paper, three dimension of the rod drop accidents were analyzed by coupling code. Based on the developed GINKGO/COCO coupling code, the effect of two sets of rod drop calculation on the core power distribution of the control rod group was studied, and the stable power of the two sets of rod drop after the automatic adjustment of the R rod was compared and analyzed. Through the numerical analysis, it can be observed that the noncenter symmetric rod groups drop accident will lead to the core power center asymmetry, and the center asymmetric rod group will lead to the core power center asymmetry, which makes the core outlet loop temperature different. The rod drop value of R rod group is small, but the adjustment range of R rod group is large. The larger reactivity value of the rod drop is, the larger the steady-state power recovery after adjustment is compared with the initial steadystate difference. The variation trend of DNBR calculated by W3 coorelation formula in the rod drop accident example shows that the opposite law with the power, and the local fine DNBR calculation can be carried out in the future combined with the more fine sub-channel coupling analysis.