Thermo-hydraulic study of the 19-pin sodium-cooled fast reactor fuel assembly clogging accident

Sodium-cooled fast reactors (SCRs) are one of the fourth-generation nuclear energy systems that can fully utilise nuclear resources and dispose of nuclear waste. Sodium-cooled fast reactors have hexagonal core fuel assemblies with tightly arranged rod bundles, which are prone to clogging accidents. In this paper, a CFD model is established for the fuel assembly of a sodium-cooled fast reactor, and computational fluid dynamics (CFD) is used to simulate the thermo-hydraulic characteristics of sodium flow in the prototype fuel assembly and the porous plugged prototype fuel assembly. From the simulation results, a CFD solution for an assembly model with 19 rod bundles was obtained. The CFD model has been validated against benchmark clogging experiments in the literature. In the study, the location of the clogging incident and the area of the clogging are considered, and comparative analyses of the flow velocity and temperature at the outlet and inside the assembly are carried out to draw conclusions. It can be seen that for the simulation of channel blockage accidents, the temperature field change at the outlet cross section is smaller and the flow velocity change at the component outlet is more sensitive. Inside the assembly, however, the return zone caused by the blocking block makes the maximum temperature change in the overall basin reach close to 100 k. This indicates that for the overall monitoring of the sodium temperature at the fuel assembly outlet is not able to detect the internal blocking in time, and that the change in the velocity field at the outlet face may be able to be used as a monitoring tool for the occurrence of the blocking flow accident.