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In the present work, we have applied a combined dynamic large-eddy simulation (LES) and direct numerical simulation (DNS) approach for a three-dimensional planar jet in a turbulent forced convection regime (Re = 18000) with a heated co-flow. Results from LES are compared with Reynolds Averaged Navier-Stokes (RANS) simulations and experimental data. We have analyzed flow and heat transfer features for four values of the characteristic Prandtl numbers (Pr = 0.71, 0.2, 0.025, and 0.006), which are representatives of air, He-Xe gas mixture, Lead-Bismuth Eutectic (LBE), and sodium, respectively. The latter two low-Prandtl fluids have been considered because of their role as primary coolants in advanced fast pool-type reactor prototypes (such as the Multi-purpose Hybrid Research Reactor for High-tech Applications (MYRRHA) at SCK•CEN, Belgium). We have provided detailed insights into instantaneous and long-term time-averaged behavior of the velocity and temperature fields (the first- and second-order moments). Furthermore, we have analyzed profiles of characteristic velocity and temperature time scales and dissipation rates, as well as the power spectra of the streamwise velocity component and temperature at several characteristic locations. The mean temperature profiles demonstrated rather low sensitivity for various values of the Prandtl number. In contrast, profiles of the temperature standard deviation exhibited larger variations, decreasing in magnitude with lower Prandtl values. Here presented results of the high fidelity numerical simulations (dynamic LES/DNS) for the low-Prandtl working fluids can be used for further development, testing, and validation of the advanced RANS-type turbulence models.

To support the design of an external filtering and conditioning system for the lead-bismuth cooled MYRRHA reactor, the formation and transport of iron oxide particles from corrosion products in the reactor primary system have been investigated for normal operating conditions. The regions of the reactor with the highest probability of oxide formation are identified by a local chemical equilibrium model for magnetite formation. This analysis reveals that magnetite precipitation generally occurs in regions with large temperature gradients. For the specific case of the MYRRHA reactor, these regions correspond to the transition region between the barrel and the upper plenum, mainly at the location of the holes in the top part of the barrel. The transport behaviour of solid oxides from these regions is then investigated with a multi-phase Euler-Lagrange particle tracking model of the MYRRHA primary system. The simulations show that the majority of large oxide particles (above 100 µm) will eventually move to the free surface without passing through the reactor core, thereby allowing their removal by an external filtering system with surface extraction. This indicates that such large particles present a minimal risk for sudden core blockage, which does not compromise reactor safety. On the other hand, particles below a threshold diameter identified at 40 µm cannot be efficiently filtered out by an external system since the majority follows the carrier liquid and re-enters the core during each LBE flow-through cycle. The continuous purification of the coolant is therefore necessary to avoid undesired build-up of suspended particles in the primary system. A preliminary design value of the required mass flow rate through the filters is identified with the support of numerical simulations.

Corrosion of steels in lead bismuth eutectic (LBE) cooled reactors can be mitigated by forming a protective oxide layer on the steel surfaces. The amount of oxygen necessary to ensure continuous oxide layer formation on fuel cladding depends on the characteristics of the steel and on the local temperature, local oxygen concentration and velocity of the LBE in contact with the steel. The most critical areas from a corrosion point of view are high temperature and low oxygen concentration regions. Wire-wrapped fuel assemblies (FAs) which are foreseen to be used in LBE cooled reactors, are characterized by hot spots and quasi-stagnant areas where oxygen could be depleted. Experimental measurements to verify whether the oxygen concentration in those critical areas is sufficiently elevated for oxide layer formation, are practically impossible. This information can be however obtained by numerical modeling. This paper focuses on the development of a numerical model of oxygen mass transfer in a 19-pin scaled fuel assembly (FA) representative of the MYRRHA reactor core. Oxidation of steels and oxygen transport from the bulk of the LBE to the surface of steels were simulated simultaneously. The simulations provide a local oxygen concentration mapping at steel/LBE interface enabling to identify the regions of the core which could be prone to corrosion due to oxygen depleted LBE. Operation recommendations for the MYRRHA reactor were given based on the simulation results.

In the present work, we combine experiments and numerical simulations of a planar jet with heated co-flow with medium (air) and low-Prandtl (He-Xe gas mixture) fluids. Jets are recognized as representative test cases to be investigated in large components of pool-type liquid metal-cooled nuclear systems, like the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA), currently under design at SCK•CEN. The present planar jet configuration mimics a closed wind tunnel that is designed and operated at VKI to generate an experimental database for velocity and temperature fields of a turbulent forced-convection flow regime. The performed experiments combine the Particle Imaging Velocimetry (PIV) (in characteristic planes) and thermocouple (single point) measurements. In parallel with experiments, comprehensive numerical simulations have been performed within the RANS modeling framework. Next to the standard eddy-viscosity based two-equation k-ε model, an extended variant based on the low-Reynolds elliptic relaxation concept (so-called ζ-f model) has been applied too. To investigate the low-Prandtl effects on the heat transfer, series of the turbulent heat transfer models have been applied, ranging from a conventional constant turbulent Prandtl number to a more elaborate kθ-εθ model. The combination of the low-Reynolds ζ-f and kθ-εθ models was explored for the first time in the content of nuclear engineering applications. The focus of the numerical studies is to address in details the effects of low-Prandtl fluid in the strongly forced convection flow (central planar cold jet) in presence of a strong shear (hot co-flow). We demonstrate the importance of the proper specification of the inlet boundary conditions in numerical simulations to mimic correctly experimentally observed asymmetrical distributions of the cross-wise profiles of stream-wise velocity, turbulent kinetic energy and temperature. Finally, the minor differences in results between the assumed constant turbulent Prandtl number and more advanced kθ-εθ model of the turbulent heat flux confirmed the overly dominant mechanisms of the strong convection and molecular diffusion in the present configuration.

Numerical analysis of the thermohydraulic behavior of the innovative flexible fast spectrum research reactor, MYRRHA, under design by the Belgian Nuclear Research Center (SCK•CEN) is a very challenging task. The primary coolant of the reactor is Lead Bismuth Eutectic, LBE, which is an opaque heavy liquid metal with low Prandtl number. The simulation tool needs to involve many complex physical phenomena to be able to predict accurately the flow and thermal field in the pool type reactor. In the past few years, within the frame of a collaboration between SCK•CEN and the von Karman Institute, a new platform, MyrrhaFoam, was developed based on the open source simulation environment, OpenFOAM. The current tool can deal with incompressible buoyancy corrected steady/unsteady single phase flows. It takes into account conjugate heat transfer in the solid parts which is mandatory due to the expected high temperature gradients between the different parts of the reactor. The temperature dependent properties of LBE are also considered. MyrrhaFoam is supplemented with the most relevant thermal turbulence models for low Prandtl number liquids up to date.

As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

The European Scaled Pool Experiment (E-SCAPE) is a 1/6th-scale thermal hydraulic model of the MYRRHA reactor and has been developed within the European framework programs THINS and MYRTE. MYRRHA is a multi-purpose hybrid pool-type research reactor cooled with lead-bismuth eutectic (LBE) that is under design at the Belgian nuclear research center SCK•CEN. The experiments from the E-SCAPE facility are essential for the design and licensing of MYRRHA. On the one hand, the experimental data from E-SCAPE directly provide information to the designers on the flow patterns and heat transport in the LBE in the upper and lower plena of the reactor. On the other hand, the experimental data from E-SCAPE enable qualification of thermal hydraulic computer codes used for analyzing the flow and heat transport in the MYRRHA reactor. The accuracy of codes rely on the underlying modelling approaches, whose validity can only be assessed by comparing them with relevant experimental data. Given the rare conditions present in heavy liquid cooled nuclear reactors, the possibility to compare complete pool simulations with experimental data is a unique opportunity. This paper presents the modelling approaches and first results of the pre-test analyses performed for E-SCAPE by NRG, SCK•CEN and VKI using Computational Fluid Dynamics (CFD). The CFD analyses of NRG are performed with the STAR-CCM + code, at SCK•CEN with the ANSYS CFX code and at VKI OpenFOAM is used. A first comparison of the results obtained with three different CFD codes for flow and heat transport in the E-SCAPE pool is presented and, where possible, validated against experimental results.

Within the SESAME framework, the Computational Fluid-Dynamics (CFD) modeling of Heavy Liquid Metal (HLM) flows in pool configuration is (mainly) aimed at giving a technical support to the design of innovative GEN-IV nuclear reactors, such as ALFRED, for electricity production, or MYRRHA, as a multi-purpose research installation. This paper presents the CFD model of the primary coolant loop of MYRRHA version 1.4, built and operated with STAR-CCM+ versions 11–13. CFD is the only tool able to capture the 3D flow effects in the plena of a pool reactor. It is however still not feasible to simulate the detailed geometry of an entire reactor core by CFD. Fortunately, for many applications (safety studies, design) the exact geometrical representation of the core, heat exchangers, pumps and bypasses can be replaced by simplified models without loss of relevant information. These parts must be modeled such that they are functionally consistent with their role. Inspiration is taken from the system code techniques like homogenization procedures. A combination of solid and porous volumes is used within the CFD model. As with system code coupling, the homogenization can lead to loss of relevant information. In this case, the flow in the plena suffers from inadequate boundary conditions provided by the over-simplified modeled parts. To reach a sufficient level of maturity, the simplification procedure must be applied only where it is strictly necessary and at the right geometrical scale. First, an overview of the general features of the design and of the most important parameters that must be preserved by the CFD model is given. Then, we concentrate on the numerical model. Its topological structure is illustrated, based on a two-level geometrical and functional decomposition. Next, we focus on the core representation. Each sub-assembly (fuel assembly, dummy assembly, control rod and scram rod) has its own position. The core horizontal cross section is vertically propagated to represent the Above Core Structure (ACS). To obtain the right flow mixing the porosity of the various interfaces is calibrated. Due to the mixture of large and small length scales detailed meshing of the core and the ACS would result in very high mesh numbers. A specific procedure has been employed, to compensate cross sectional variations by physical property changes, maintaining the most important global characteristic. The ACS parameters have then been tuned in order to reduce the hot plenum surface temperature. Bypass flows have been implemented. Their role is to connect the fluid volumes and to avoid stagnation regions, which could cause potential corrosion issues. With the bypass flows and the simplified core/ACS structure models, a versatile CFD model for nominal operation is available. The Volume of Fluids (VoF) method used in the simulation allows the simulation of scenarios with variation of the free surface levels. Transient scenarios can be implemented to simulate operational incidents or accidents. To evaluate the safety of the reactor design, combinations of extreme parameter values are used in the simulations to demonstrate all safety criteria are met under worst case conditions. The simulations can also be used to define technical requirements of the system components (pumps and heat exchangers). Some examples are illustrated and partial conclusions are drawn.