A lumped-parameter numerical model was constructed based on the conservation laws of mass and energy and the point neutron kinetics with 6 groups of delayed neutron to represent the dynamics of primary loop of a pressurized water reactor (PWR) core. On the viewpoint of control theory, the coupled phenomenon of neutron kinetics and thermohydraulics can be recognized as a dynamic system with feedback loops which is caused by the Doppler effect and the coolant temperature difference. Scilab was implemented to representing the equivalent transfer functions and associated feedback loops of a PWR core. The dynamic responses were performed by the perturbations of coolant inlet flow, coolant inlet temperature, and reactivity insertion.

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.

Nuclear fuel elements assemblies are generally consists of fuel rod bundles; each bundle is a concentric cylinder with three layers. Taking GE-8X8 for example, there is pellet, gap, and cladding from inside to outside. The diameter for each concentric cylinder is 9.26cm, 9.47cm, and 10.71cm respectively. In reality, a structure deformation may happen to those components due to the reason of radiation result in the high temperature of the bundles system. For the space of gap decreases by the expansion of pellet, the thermal conductivity might be under predicted and there is not enough study about this topic yet. To improve the accuracy of PRAs, more studies of the shrink phenomena on the gap between pellet and cladding are necessary. In this study, we had developed a program on the purpose of processes improvement for CFD simulation about spent fuel dry storage system. The program can adjust the dimension for each part of formation very friendly. We think it can also do some help on the needs if we want to compare the performance on heat transfer for different fraction on each part of bundle. In addition, the axial power distributions of the rod were also defined file by the user very easy, the results shown no obviously temperature difference between the full gap and 90% reduction of the gap.

After Fukushima nuclear power plant (NPP) event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the failure of spent fuel pool cooling for Chinshan NPP by TRACE. In this study, by using the above TRACE results, we focused on the application of FRAPCON-3.4 in the spent fuel pool safety analysis of Chinshan NPP. FRAPCON-3.4 can calculate the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. There are two steps considered in this study. The first step is the verification of the FRAPCON-3.4 by using IFA-431 experimental data. The next step is the fuel analysis of Chinshan NPP spent fuel pool by using FRAPCON-3.4 and the TRACE results.