Assessment of the CUPID code applicability to the thermal-hydraulic analysis of a CANDU moderator system

The CUPID code has been developed for a transient analysis of two-phase flows in nuclear reactor components. The primary objective of this study is to assess the applicability of the CUPID code to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor. At first, the CUPID code is validated against the Stern experiments, which were carried out to investigate the flow in a Calandria vessel. To represent the complicated internal structure of the Calandria vessel, a porous media approach is adopted for the tube bundle region of the Calandria vessel, and an open media approach is used for the outer region. Then, the two regions are modeled using a three-dimensional grid system with polyhedral meshes and bent-structured meshes, respectively. The calculation results of single-phase flow experiments showed good agreement with the experimental data. Thereafter, a hypothetical two-phase flow transient is simulated to assess the CUPID code applicability to two-phase flows analyses.