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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.

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.

Pressurized water reactors (PWRs) are the most common types of electricity generating nuclear reactors. Within their cores, fuel rods generate heat by fission and the high-pressure water is used as a coolant. The heated water is then used as a heat source in a steam generator that boils water in the secondary loop. The steam is used to spin a turbine and generate electricity. Spacer grids are key components of a PWR’s core. Their objectives are to maintain the fuel rods at their positions and enhance the coolant mixing and heat exchange. Over the past decades, numerous experimental and numerical studies have been performed, to characterize the flow induced by different types of spacer grids. The recent advancements of the computational description of the above-mentioned flows, led to a shift to higher resolution models for turbulence like Large Eddy Simulations. Therefore, highly spatially resolved experimental data is needed for code validation. The comparison and usage of the data about the flow in rod bundles with spacer grids obtained by different research groups is challenging. One major reason comes from the fact that the majority of the spacer grids studied are proprietary, and their geometries cannot be disclosed. The data required for some purposes like numerical code validation do not necessarily have to originate from one specific type of spacer grid. It is important to provide a detailed overview of the geometry tested and the experimental conditions. Equally crucial is to possess high spatial resolution and well-quantified and low relative uncertainty. In this study, a new, non-proprietary spacer grid is designed and 3D-printed. The flow induced by the spacer grid is characterized in a 5 × 5 rod bundle facility with Matching of Index of Refraction (MIR) at a Reynolds number Re = 27 , 390 using particle image velocimetry techniques. A complete uncertainty analysis is performed accounting for a variety of uncertainty sources. A low relative uncertainty is obtained. Eight middle-of-subchannel planes are selected as the domain of interest for this paper. Highly spatially resolved statistical results for the velocities, root-mean-square (RMS) fluctuating velocities, and Reynolds stresses are obtained along the span-wise direction of the flow, starting at the edge of the spacer grid up to 4.3 hydraulic diameters. The cross-flow between subchannels induced by the spacer grid is sustained for all the domain assessed. The turbulence induced by the spacer grid is assessed with the use of two-points cross-correlations and with the comparison of the RMS fluctuating velocities at different elevations. An increase of the size of the turbulent vortices along the span-wise direction is related to the decrease in the rms fluctuating velocities to evidence the turbulent dissipation. Differences between the size of the turbulent structures found with integral length scale calculations in the stream-wise and span-wise directions are attributed to the turbulence anisotropy. The full dataset and spacer grid geometry design will be available on the research group website of the authors. The data obtained can be used by other groups according to their interests.

Perforated plates with thin orifices is one of the options under considerations to be used in the debris filtering bottom end pieces for the next generation of fuel assemblies for the Brazilian nuclear plants. The increase in pressure drop can be drastically reduced by optimizing the chamfer geometries. This paper describes the development and validation of a procedure devised to use Computational Fluid Dynamics (CFD) to accurately estimate the pressure drop of water flow through perforated plates with chamfered thin orifices. The procedure comprises a scaled down calculation domain, grid configuration and a set of numerical solution parameters. The validation was performed against experiments with a set of plates with different chamfer geometric features like position (inlet, outlet and both sides), angles and sizes. Three turbulence models were evaluated with the standard k–ɛ giving the best result. For the range of parameters evaluated here the pressure drop decrease of a two sided chamfer plate can be estimated as a sum of both inlet and outlet chamfers individually. In both cases the pressure drop decreases rapidly for small chamfers and more slowly for larger chamfers. In the flat regions the chamfer angle has little influence at the inlet and is more important at outlet

Thermal stratification and striping are observed in many piping systems including those of nuclear power plants. Periodic occurrences of these thermal transients lead to fatigue and may induce undesirable failures and deformations to the piping. The Thermal Hydraulic Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear (CDTN/CNEN) conducts an experimental and numerical project simulating the thermal stratified flows in piping systems of pressurized water reactors (PWR) to obtain some understanding on these phenomena. Experiments were carried out in a test section simulating the steam generator injection nozzle of a PWR. A numerical simulation of one experiment was performed with the commercial finite volume Computational Fluid Dynamic code CFX 13.0. A vertical symmetry plane along the pipe was adopted to reduce the geometry in one half, reducing mesh element size and minimizing processing time. The RANS two equations RNG k–ɛ turbulence model with scalable wall function and the full buoyancy model were used in the simulation. In order to properly evaluate the numerical model it was performed a Verification and Validation (V&V) process. Numerical uncertainties due to mesh refinement and time step were evaluated. This validation process showed the great importance of a proper quantitative evaluation of numerical results. In past studies qualitative evaluations were considered enough and numerical results like the one presented here could be considered satisfactory for the prediction of thermal stratified flows. However, with the present V&V study it was possible to identify objectively the strengths and weaknesses of the model.

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 presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 ?? 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k ??? ?? turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.

An experimental and numerical study was conducted on the pressure loss of flows through the bottom end piece of a nuclear fuel assembly. To determine an optimized numerical methodology using the commercial CFD code, CFX 10.0, a series of preliminary simulations of water flows through perforated plates in a square ducts were performed. A perforated plate is a predominant geometry of the bottom end piece, responsible for the majority of the flow’s pressure drop. The numerical pressure loss applying an optimized mesh and the k-e turbulence model showed good agreement when compared with a conventional methodology (Idelchik). Numerical results for the standard bottom end piece were obtained applying the previously determined mesh criteria and the k-e turbulence model with some geometric simplifications. The agreement between the numerical simulations and experimental results can be considered satisfactory but suggests further numerical investigations with the bottom piece under real conditions of the experiment, without the geometric simplifications and with a gap between the piece and the wall of the flow channel. Additionally, other turbulence models should be appraised for this complex geometry.