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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.

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.

CNPq - Conselho Nacional de Desenvolvimento Científico e Tecnológico FAPEMIG - Fundação de Amparo à Pesquisa do Estado de Minas Gerais CAPES - Coordenação de Aperfeiçoamento de Pessoal de Nível Superior FINEP - Financiadora de Estudos e Projetos, Financiadora de Estudos e Projetos Outra Agência As grades espaçadoras são importantes componentes presentes em elementos combustíveis de reatores nucleares à água pressurizada, PWR – Pressurized Water Reator. Essas grades são responsáveis por manter a integridade estrutural do elemento combustível e contribuem com a melhoria da eficiência termo-hidráulica do reator. Com o objetivo de desenvolver uma grade espaçadora nacional de alta eficiência, estudos com grades espaçadoras comerciais e protótipos fabricados em impressora 3D têm sido realizados no Laboratório de Termo-Hidráulica e Neutrônica – LTHN do Centro de Desenvolvimento da Tecnologia Nuclear - CDTN. Neste trabalho foi avaliado o desempenho termo-hidráulico de três grades espaçadoras, sendo dois tipos comerciais (aletadas e de canais) e de uma grade impressa do tipo de canais. As grades foram testadas em elementos combustíveis nucleares representativos de modelos comerciais distintos, porém constituídos por arranjos quadrados de 5x5 varetas. Parâmetros de turbulência foram determinados a partir de velocidades medidas a jusante das grades espaçadoras por meio da técnica de velocimetria laser Doppler (LDV – Laser Doppler Velocimetry). Para a grade aletada foram avaliadas cinco condições de escoamento para Reynolds (Re) na faixa de 18x10³ a 54x10³. Os resultados mostraram diferenças significativas no comportamento do escoamento para essa faixa de Re. Os testes de comparação entre as três grades foram feitos para Re = 27x10³. Os resultados de comparação entre as grades mostraram que sob o ponto de vista termohidráulico, a grade espaçadora aletada é superior à grade comercial de canais e também com relação à grade impressa avaliada. Entretanto a partir dos resultados obtidos para a grade espaçadora impressa comprovou-se a viabilidade do uso da técnica de prototipagem no desenvolvimento de novos elementos combustíveis nucleares. Foi criado e disponibilizado um banco de dados experimental da grade impressa que poderá ser utilizado como dados de entrada em códigos de simulação numérica e análise termo-hidráulica de subcanais e para validar simulações de fluidodinâmica computacional (CFD – Computational Fluid Dynamic). Spacer grids are important components in nuclear fuel assemblies of Pressurized Water Reactors – PWR. These grids are responsible for maintaining the structural integrity of the fuel assemblies and contribute to enhance the thermo-hydraulic efficiency of the reactor. In order to develop a high-efficiency national spacer grid, studies with commercial spacer grids and prototypes manufactured using a 3D printer have been carried out at the Thermo-Hydraulic and Neutronic Laboratory (LTHN) of the Nuclear Technology Development Center - CDTN. Thermo-hydraulic performance of three spacer grids was evaluated in this work: Two types of commercial (mixing vanes and channels) and one printed grid (channels). Spacer grids were tested in different representative nuclear fuel assemblies with a 5x5 square rod bundle. Turbulence parameters were determined from velocities measured downstream of the spacer grids using the Laser Doppler Velocimetry – LDV. For the mixing vanes spacer grid, five flow conditions were evaluated by Reynolds in range of 18x10³ to 54x10³. The comparison tests between these three spacer grids were made for Re = 27x10³. Results showed significant differences in flow behavior for the Re range selected. The comparison results between spacer grids showed that the mixing vane spacer grid is superior to subchannel type and printed grid based on the thermo-hydraulic performance. From the results obtained from the printed spacer grid it was possible to demonstrate the viability of 3D printed prototypes for assessment and development of nuclear fuel assemblies. An experimental benchmark data base was generated for a printed spacer grid and the obtained data can be used as input data in numerical simulation codes and sub-channel thermo-hydraulic analysis and can also be used to validate computational fluid dynamics (CFD) simulations.

The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5x5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8x104 to 5.4x104. This experimental research was carried out in thermohydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations.