Yuanzineng kexue jishu (Aug 2022)
Challenge, State-of-art and Future of Two-phase Flow in Light-water Nuclear Reactor
Abstract
The two-phase flow is a vital topic of thermal-hydraulic research in light-water nuclear reactors. The fluid dynamics and heat transfer characteristics of two-phase flow impact the safety and efficiency of nuclear power plants. The two-phase flow results in various transients in the nuclear reactor: The change of the pressure drop redistributes the flow rate of the channels, and the generation of the void in the core alters the reactivity and consequently the power. The two-phase flow is among the most complex flows since the interface structure evolves and the mass/momentum/energy transfers across the interface. The two-phase flow research of the nuclear system started in the 30s and 40s of the last century, when the pioneer scientists studied the flow instability and the pressure drop. Along with the rise of commercial nuclear reactors in the 50s, the interests lay in the critical flow, the critical heat flux, and the void distribution. In the 70s, the industry shifted its interest to the blow and the flooding, which are compound flow phenomena in specific geometries of the reactor system. In the 90s, the focus was to improve the ability of the nuclear safety codes, such as developing the constitutive models and conducting benchmark experiments. Since this century, the growth of the measuring techniques has supported fundamental research, like the interface evolution and the behavior under extreme conditions. Also, the development of the computers enables the multi-dimensional study using computational fluid dynamics codes and enables the multi-physics study using the codes-coupling platform. This paper aims to understand the research trend of the two-phase flow in the nuclear system by reviewing the history, summarizing the critical problems, and presenting the status. Five phenomena, which lack knowledge and need to be investigated, are described in the current review: the two-phase flow structure, the interfacial drag, the heat transfer crisis, the interphase heat/mass transfer, and the counter-current flow limitation. Three research directions are suggested: advanced modeling, advanced measuring techniques, and cross-cutting fields. Concluding remarks, with research recommendations from the local scale to the system scale, are given to the two-phase flow in the nuclear system.
