Numerical investigation of the CANDU moderator thermal-hydraulics usingn the CUPID code

摘要

The CUPID code has been developed for a component-scale thermal-hydraulic analysis of single- and two-phase flows in light and heavy water reactors. As an application to CANDU nuclear reactor, the single- and two-phase natural circulation flow inside the moderator tank has been analyzed by assessing the experiments conducted at the 1/4-scaled test facility at the Stern laboratory. A porous media approach was applied for the Calandria tube bundles to avoid computational complexity. This resulted