Thermalhydraulic analysis of U-7Mo/Al and U-6Zr/Al experimental fuel Element in RSG-GAS by using COOLOD-N2, NATCON and CFD-3D

The Center for Nuclear Fuel Technology BATAN through research and development of low-enriched uranium fuel (<20% U235) has a plan for irradiating experimental fuel element plates of U-7Mo and U-6Zr/Al in RSG-GAS. G-7 irradiation position is therefore selected with a consideration to give minimum reactivity changes. Thermal hydraulics safety analysis at steady state and free convection is performed on the power of 30 MW and 15 MW. One-dimensional computational programs, COOLOD-N2 and NATCON, are used to calculate thermal hydraulics parameters in conditions of forced and free convection, while three-dimensional computational fluid dynamic FLUENT is used to provide visualization of thermal hydraulics parameters. The comparison of COOLOD-N2 and FLUENT calculation results shows a deviation between 2.78% to 13.92%. Safety margin of U-7Mo and U-6Zr/Al irradiated at 15 MW and 30 MW meet the safety margin required in the RSG-GAS Safety Analysis Report. (author)