Research on the effects of local throttling on thermal-induced two-phase flow instabilities in the typical coolant open-channel system in the nuclear reactor core

Thermal-induced two-phase flow instabilities could badly slow the runtime performance of systems, such as boiling water reactor core, and even endanger safe operation. Hence, many researchers have carried out numerous researches on instability characteristics, but few publications cover effects local throttling in open-channel systems coolant rod bundle subchannel with spacer grids nuclear core. This paper provides a numerical study simplified typical system core by using NUSOL-SYS code. The ratio, position, other parameters stability channel were out. results show that usually uniformly distributed conditions, high ratio are positively correlated. In fixed throttle is correlated distance from region to entrance void fraction. If regions arranged along heating certain value, number will enhance system. Besides, this preliminarily proposed criterion function f(n, RT, ?) complicated which provide technical reference for design operation especially,