Application of Direct Numerical Simulation to Determine the Correlation Describing Friction Losses during the Transverse Flow of Fluid in Hexagonal Array Pin Bundles

Yury E Shvetsov,Mikhail V Bayaskhalanov,Regina P Dichina,Yury S Khomyakov

doi:10.3390/fluids7010022

Yury E Shvetsov, Mikhail V Bayaskhalanov + Show 2 more

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https://doi.org/10.3390/fluids7010022

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Abstract

This paper presents the results of a numerical simulation to determine the hydraulic resistance for a transverse flow through the bundle of hexagonal rods. The calculations were carried out using the precision CFD code CONV-3D, intended for direct numerical simulation of the flow of an incompressible fluid (DNS-approximation) in the parts of fast reactors cooled by liquid metal. The obtained dependencies of the pressure drop and the coefficient of anisotropy of friction on the Reynolds number can be used in the thermal-hydraulic codes that require modeling of the flow in similar structures and, in particular, in the inter-wrapper space of the reactor core.

Highlights

Under development are various liquid metal cooled nuclear reactor projects related to Generation IV reactors
One of the design features of a typical fast reactor with fuel assemblies of cassette type in the core is the presence of an inter-wrapper space (IWS) in the core
This means the presence of, simultaneously, two coolant flows in the reactor core: one of them is the main one, passing inside the fuel assembly and responsible for heat removal from the fuel rods, and the other, passing through the IWS and usually making up a few percent of the total flow through the reactor

Summary

Introduction

Under development are various liquid metal cooled nuclear reactor projects related to Generation IV reactors. One of the design features of a typical fast reactor with fuel assemblies of cassette type in the core is the presence of an inter-wrapper space (IWS) in the core This means the presence of, simultaneously, two coolant flows in the reactor core: one of them is the main one, passing inside the fuel assembly and responsible for heat removal from the fuel rods, and the other, passing through the IWS and usually making up a few percent of the total flow through the reactor. Both streams, in the course of their movement, exchange heat through the fuel assembly. This influence was considered in [1]

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