Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.