Abstract
A direct reactor auxiliary cooling system (DRACS) under natural circulation (NC) conditions with a dipped-type direct heat exchanger (D-DHX) in the upper plenum of a reactor vessel (RV) has been investigated for enhancing the safety of sodium-cooled fast reactors. Studies of the past have revealed that core-plenum interactions, which consists of penetration of the coolant from D-DHXs into the subassemblies and the narrow gap between them (IWF: inter-wrapper flow), and the heat transfer through a wrapper tube among subassemblies (radial heat transfer), occurred and increased core cooling performance during the DRACS operation. Therefore, a multidimensional thermal-hydraulics analysis model in the RV using a computational fluid dynamics (CFD) code (RV-CFD model) was developed to evaluate the core cooling performance. For the design study, the RV-CFD model must simulate reasonable calculation costs while maintaining accuracy. In this study, the subchannel analysis method using the CFD code for fuel subassemblies (subchannel CFD model) was applied to the RV-CFD model. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in a diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Two numerical simulations were compared to the experimental data obtained from the sodium experimental apparatus PLANDTL-1. In the first case, the focus was only the radial heat transfer without the D-DHX operation. In another case with the D-DHX operation, the IWF noticeably occurred, and the focus was on the core-plenum thermal interaction. The calculated sodium temperature in the core correlated well with the experimental results. The RV-CFD with subchannel CFD model was validated for core-plenum interactions during the DRACS with the D-DHX operation under NC conditions.
