29th Symposium of AER on VVER Reactor Physics and Reactor Safety (2019, Energoland, Mochovce NPP, Slovakia)
nuclear applications of computational fluid dynamics (CFD)
Vít Doleček1*, Ladislav Vyskočil1 and Václav Železný1,2
1: UJV Rez a. s., Dept. of Safety Analyses, Hlavni 130, 250 68 Husinec-Rez, Czech Republic
2: Czech technical University in Prague, Faculty of Mechanical Engineering, Technicka 4,
166 07 Praha 6, Czech Republic
The goal of this work was to evaluate crossflows in a mixed core of VVER reactor. The mixed core consists of old and new fuel assemblies with different geometry. These fuel assemblies have different spacer grids and mixing grids at different elevations. The old and new fuel assemblies also have slightly different total pressure loss. This causes crossflows between neighboring fuel assemblies. Crossflows were evaluated with VIPRE-01 subchannel code and with Ansys Fluent CFD code. In both codes, the computational domain covers an interface among fuel assemblies of different type along the whole core height. VIPRE-01 domain is modelled pin-by-pin and it includes corner plates. Spacer and mixing grids are represented by local pressure losses. CFD model has much finer grid resolution. Spacer and mixing grids are modelled by thin walls and porous zones. Due to the enormous size and complexity of the CFD domain, the mixing vanes were not modelled. Transverse coolant velocities along the core height were calculated in both codes and compared. The results obtained with subchannel code are in a good agreement with results from CFD simulation.