No abstract

In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels. The aim of this paper is to present a practical approach to perform coupled calculations between neutronic (Monte Carlo) and thermal-hydraulic (CFD) codes. For this purpose, is necessary to evaluate the influence of the water thermal-physical properties temperature variations on keff parameter. Besides that, Serpent Nuclear Code was used for the neutronics evaluation, while OpenFOAM was used for thermal-hydraulics. OpenFOAM si- mula-tions were made by using a modified chtMultiRegionFoam solver, developed to read Serpent output correctly. The neutronic code was used without any modifications. The results shows that this coupled calculations were consistent and that leads to encouraging further methodology development and its use for full core simulation. Also, the results shows good agreement with calculations performed using other version of OpenFOAM and Milonga as neutronic code.

In the present work, a numerical study of countercurrent flow limitation (CCFL) with air-water as working fluids was carried out using Computational Fluid Dynamics (CFD). The CFD code Star-CCM+ was used, presenting itself a robust and reliable tool for this type of problem. The used geometry was a scaled (1:14) PWR hot leg, with small deviations from the experimental facility built in the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN). The aim of this work was to study numerically the behavior of the system in a potential Loss of Coolant Accident (LOCA) coming from the cold leg of the reactor, which would cause an interruption in the core cooling capabilities. Experimental data from the CDTN experimental facility were used as input for Star-CCM+, which imply in a fixed water mass flow and raising air mass flow. The raising in air mass flow was performed by giving ramp increments through simulation time, providing numerical stability and reproducing the experimental conditions. A numerical tactic was employed regarding the air density variation in the experiment. As a result, the degree of freedom of the simulation were decreased, leading to an enhancement in the numerical stability. As a first assessment, only one mesh was used with a fixed time step, allowing the evaluation of the numerical stability using global Convective Courant Number. The numerical result was compared with experimental ones, showing consistency with the actual physical behavior of this kind of system.

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.