CFD investigation on thermal-hydraulic behaviors of a wire-wrapped fuel subassembly for sodium-cooled fast reactor

Different from the square-arrayed fuel rod bundles in pressurized water reactors, the fuel subassembly for sodium-cooled fast reactors features with wire spacers helically wrapped on the fuel rod. This intricate geometry consequently imposes a significant challenge in numerical simulation. In the present study, a new marked and separated meshing method is developed to reduce the required mesh and modeling time. Instead of directly generating the complex geometry and mesh, the hexahedral wire wrapper mesh was dealt with locus equations. First, the numerical modeling method was validated by combining the experimental data of a 19-pin wire-wrapped rod bundle and the simplified Cheng and Todreas correlation. Furthermore, a comparative study of different meshing methods was conducted. Then, an investigation on the three-dimensional flow phenomenon and heat transfer characteristics of a 61-pin wire-wrapped fuel subassembly for the China Experimental Fast Reactor was carried out. Finally, complex axial and transverse flow behaviors and temperature field in different types of subchannels were quantitatively analyzed. Results indicate that the axial velocity exhibits periodic oscillation with a maximum normalized value of 1.54. The flow oscillation in the interior subchannels presents a smaller amplitude and higher frequency than that in the corner and edge subchannels. Local hot spots occur in the gaps beneath the spacer wires. Sweeping and cross flow patterns exist in the peripheral and interior subchannels, respectively. In addition, the secondary transverse flow induced by the spacer wires is stronger in the corner subchannel. The cross flow strength reaches a peak value equal to 30% of the average axial velocity, which promotes the thermal mixing.