Thermal hydraulics analysis of a nuclear reactor transient involves complicated gas-liquid two-phase flows appearing in many different flow patterns under various operating and accident conditions. For example, the system pressure drops, within a minute, from 150 bars to the atmospheric pressure within a minute following a loss of coolant accident of a pressurized water reactor (PWR), resulting a dramatic phase change. Furthermore, the nuclear reactor system is very big and complicated. The reactor vessel is about 10 meters in height and having around 50 thousand fuel rods for a typical PWR. Thus, a multi-scale analysis approach is widely used where the analysis length scales are usually defined as local (~10^-3 m), component (~10^-1 m), and system (~10⁰ m) scales. The conventional safety analysis model of a nuclear reactor usually adopts the system scale with simplified one-dimensional models.
Recently, with the advances in high performance computers (HPC), more advanced model and simulation technology is being applied for the nuclear reactor performance and safety analysis. In this paper, a state of the art multi-scale and multi-physics simulation method is presented to enhance the safety analysis of nuclear reactors based on CUPID which is a three-dimensional two-phase flow analysis code developed by Korea Atomic Energy Research Institute (KAERI). Advances in two-phase flow analysis for PWRs are discussed for the local and component scales using the CUPID code. This includes a wall boiling heat transfer at the fuel rod surface and a radiation heat transfer in the containment at the local scale. Then the component scale applications are introduced for the reactor core and the steam generator of PWRs. A multi-scale coupling method is also presented for the application to the PWR transient analysis, where an implicit method is applied for the coupling.
Nuclear reactor simulation also involves different physics such as the neutron kinetics for the reactor core power generation, the fuel structure mechanics for the heat transfer from the fuel rod to the coolant, and the thermal hydraulics for the reactor coolant flow. Simultaneous multi-physics simulation is available with the advanced HPC, and this can help greatly reduce the prediction uncertainty. Practical application of the multi-physics simulation to the nuclear reactor safety analysis is demonstrated and the benefit of this simulation is discussed. Special focus is given to the limitation and prospect of the heat transfer models important for the nuclear reactor application.