In nuclear power plants, the occurrence of a three-dimensional mixing phenomenon in two-phase flow can help mitigate the peak cladding temperature of fuels in a core. Simulating this phenomenon necessitates subchannel-scale analysis. CUPID is a three-dimensional thermal hydraulic analysis code developed by KAERI. This paper presents the validation results of the code against rod bundle tests conducted using SIRIUS-3D, a 5 × 5 rod bundle test facility. One test utilized subchannel void sensors, while the other employed X-ray CT. To validate the SIRIUS-3D test data, subchannel-scale nodes were constructed for CUPID calculation. For improved convergence, the implicit friction calculation was conducted with a small CFL ratio. The drag coefficient interfacial friction model without the EVVD model yielded the best prediction. The void fraction was over-predicted at a low velocity of 0.1 m/s, which could be improved by reducing the interfacial friction.