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

Abstract

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 in a good agreement with the experimental data as well as a cost-effective analysis. In this work, the analysis is extended to the prototype geometry of a CANDU reactor. A similar coarse mesh model was adopted for the Calandria tube bundle region. However, because of the complicated shape of the inlet nozzle, the flow distribution at the inlet nozzles was complicated. In this analysis, each inlet nozzles was not modeled in detail and, instead, three different boundary conditions for the inlet nozzle flow were examined. The results showed that, by using an appropriate inlet flow modeling and the porous media approach, the CUPID code can cost-effectively predict the thermal-hydraulics in the CANDU moderator tank.