Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—I: Two-Channel Analyses

Pressure loss coefficients are generally required by subchannel and system thermalhydraulics codes. These coefficients are not readily available for small modular reactors featuring non-conventional designs and novel coolants. In the present study, the pressure loss coefficients were obtained using three-dimensional (3D) computational fluid dynamics (CFD) modelling for an advanced water-cooled reactor. A representative light water fuel assembly used in the OECD/NRC pressurized water reactor Subchannel and bundle tests (PSBT) benchmark was selected for CFD modelling and simulation under various working conditions. The fuel assembly includes three types of PWR spacer grids: simple spacer grid (SG), non-mixing vane spacer grid (NMVG), and mixing vane spacer grid (MVG). Turbulent flow through subchannels of both non-heated and heated rod bundles was simulated to predict recoverable and non-recoverable pressure distribution along the length of the bundle. It was observed that vortices were generated at the tips of spacer grids, affecting the cross-flow in subchannels significantly. The estimated pressure loss coefficients were found to be influenced by the flow conditions (Reynolds number or the upstream flow history) and spacer grid configuration. Pressure loss coefficient values ranged from 1.14 to 1.80, depending on the spacer grid type, design, and flow conditions. The CFD methodology used in this study was demonstrated to have the potential to generate input parameters required for the subchannel analysis and optimization of fuel assembly designs and serve as a surrogate for empirical correlations.

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Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

Podila

Bromley

et al. 2022

Journal of Nuclear Engineering and Radiation Science

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Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—I: Two-Channel Analyses

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Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

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