Numerical Study of Thermal-Hydraulics in CANDU-typed Moderator

The moderator plays an important role to sustain a robustness of the fuel channel during a hypothetical loss of coolant accident (LOCA) in the Canada Deuterium Uranium (CANDU) reactor vessel. Since the pressure tube stain to contact its surrounding Calandria tube (PT/CT contact) could lead a CT dryout, it is important to estimate a local subcooling of moderator inside the Calandria vessel. However, it is possible to measure the local temperature only in the inlet/outlet region. Thus, it is needed to estimate the local subcooling of moderator at arbitrary region inside the Calandria vessel for the accident such as LOCA. In order to measure the local temperature quantitatively, numerous experimental and numerical researches have been investigated. Over all the numerical studies, the subcooling in the moderator and local velocity and temperature fields have been the only purpose rather than the boiling and corresponding two phase phenomena for the LOCA accident. In this study, the CUPID code is used to validate the existing mixed and natural convection regime as well as two phase phenomena feasibility