Session: 08-05: Computational Fluid Dynamics (CFD) and Applications - V

Paper Number: 135052

135052 - Study of the Effect of Different P/d Ratios on the Thermal-Hydrodynamic Performance of Lead-Cooled Fast Reactor Fuel Assemblies. 

Abstract: 

The core is the area of the nuclear reactor where the fission reaction occurs, and a complete thermo-hydraulic study is necessary to offer vital research data to assure the safe and stable functioning of the nuclear reactor. The safety of the core structure is endangered in lead-cooled fast reactors because of the large number of fuel rods in each box of fuel assemblies and the high average heat flux density on the surface of the fuel rods. Additionally, the large temperature difference caused by incomplete coolant mixing in the fuel assembly's outlet area will result in greater thermal stress on the fuel rod cladding and related solid components near the fuel assembly's outlet area. According to the study, the ratio of the pitch of the fuel rod to the diameter of the fuel rod (P/D) and the width of the side channel of the fuel assembly have a great influence on the uniformity of the coolant temperature at the outlet of the fuel box. However, there are few studies on the temperature uniformity of coolant in the outlet area of fuel assemblies in lead-cooled fast reactors. Consequently,CFD numerical simulation is utilized to numerically simulate the fuel assembly structure of a 1/6 lead-cooled fast reactor based on fluent software and the SST k-ω turbulence model, based on the fuel assembly parameters and operating boundary conditions of the small and micro-innovative lead-cooled nuclear energy system (SMILE).Initially, the fast reactor fuel assemblies' thermal and hydraulic calculations are completed, and the viability of the numerical simulation calculation method is confirmed.Subsequently, the three 3D models with P/D values of 1.15, 1.20, and 1.24 were constructed, and their opposing side channel widths were subjected to sensitivity analysis. The numerical simulation results show that with the increase in P/D, the uniformity of coolant temperature at the outlet of the fuel tank is greatly improved. As the width of the opposite side channel increases, the fuel assembly's corner channel temperature gradually drops. However, the fuel assembly box's cold wall effect becomes more pronounced, causing more coolant to flow away from the side channel. This causes the side channel temperature to drop quickly and the center area's temperature to rise, increasing the coolant temperature difference at the outlet.When P/D = 1.24 and the width of the side channel is 1.95mm, the outlet temperature of the fuel assembly is effectively flattened, and the temperature difference is reduced to 36.9 °C, which improves the thermal safety margin and ensures the safety of the core structure.

Presenting Author: Qiong Cao North China Electric Power University

Presenting Author Biography: Associate professor at the School of Nuclear Science and Engineering, North China Electric Power University, engaged in the thermal hydraulics of nuclear reactors.

Authors:

Yangguang Zhang North China Electric Power University
Daogang Lu North China Electric Power University
Qiong Cao North China Electric Power University
Xiaotian Wang North China Electric Power University