Investigação numérica e experimental do escoamento de água em feixe de varetas representativo de elementos combustíveis nucleares de reatores do tipo PWR

Abstract

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The spacer grids cause a cross and swirl flow between and within the subchannels, enhancing heat transfer in the grid vicinity. Experimental and theoretical investigations, such asComputational Fluid Dynamics (CFD) analysis, have been carried out in the past years to study these important thermal and fluid dynamic features. Due to the detailed discretization of the flow characteristics obtained through CFD simulations, this analysis is quickly becoming the preferred tool for the improvement of spacer grid designs. However, before CFD can be considered as a reliable tool for the analysis of flow through rod bundles there is a need to establish the credibility of the numerical results. Procedures must be defined to evaluate the error and uncertainty due to aspects such as mesh refinement, turbulence model, wall treatment and appropriate definition of boundary conditions. These procedures are referred to as Verification and Validation (V&V) processes. With the intention to subsidize the development of the Brazilian nuclear industry with tools for the development of a totally national nuclear fuel assembly for PWR reactor, this work aims for the definition of an experimental and numerical methodology to investigate the water flow through spacer grids with mixing devices. The developed methodology employs the CFD tool and experiments in a 5 x 5 rod bundle fuel element segment to assess pressure loss and velocity profiles, obtained with an LDV (Laser Doppler Velocimetry) system. Data from literature was used for the initial development of the numerical procedure. A V&V process was performed according to the ASME V&V 20 standard. Numerical uncertainty was estimated and five turbulence models were evaluated. The verified and validated numerical procedure was applied to the simulation of the experimentally tested spacer grid. The numerical uncertainty values were extrapolated from the initial development. The numerical results were compared to the experimental data obtained in a test section developed in this work. The numerical results showed good agreement to the experimentally obtained data for pressure loss and the velocity profiles. The procedures presented and developed in this work can be very useful for future projects of spacer grids for nuclear fuel elements.