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The increased use of electrothermal ice protection systems (ETIPSs) in various industries (manned and unmanned aviation, energy production) requires improvements to de-icing efficiency, ice shedding predictability, and energy consumption. To achieve these, a coupled numerical and experimental investigation of ETIPS’s ice removal (ice shedding) mechanism is presented in this paper. Idealized ETIPS de-icing experiments performed in the icing wind tunnel of the von Karman Institute show several ice shedding mechanisms. A one-dimensional phase change solver developed for ice melting simulations highlights the water layer thickness influence over the ice shedding process. Coupled numerical–experimental results are employed to develop an idealized ice shedding model. The model is validated in realistic de-icing experiments in a second experimental campaign in the icing wind tunnel of the LeClerc Icing Research Laboratory.

The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics...

This study considers the decay heat removal through passive systems (i.e., natural convection) in the pool-type experimental reactor MYRRHA (SCK•CEN). The low velocities associated with natural con...

In laser cutting processes, the air nozzle, used to evacuate the material molten by the laser, has a strong influence on the cutting quality and efficiency. The present paper proposes flow diagnostics to evaluate this efficiency based on the numerical evaluation of the flow aerodynamic around the nozzle and its interaction with the kerf. The computational domain includes the nozzle reservoir, where the pressure is imposed, the nozzle and the metal plate with its kerf. The flow diagnostics are based on Mach and momentum iso-contours, mass balancing and wall-shear stress on the kerf front. This allows discriminating between different nozzle configurations and to identify the most efficient aerodynamically. An application on four different nozzles is proposed, and results showed that the cut quality observed experimentally in real working conditions may be explained by the diagnostics put in place to evaluate the cutting efficiency.

The increasing interest in Small Air Transportation (SAT), to enhance global connectivity, is highlighting the need for reducing take-off distance: lift coefficient has to be increased without penalizing the configuration drag, weight and complexity. Within the EU Clean Sky 2 projects, a blown flap configuration has been developed to allow STOL capabilities of a future affordable and green commuter belonging to EASA CS 23. The blowing system design choice is aimed at keeping the inevitable associated increase in pitching moment very low, and not penalizing performance when blowing fails. In collaboration between Piaggio Aerospace and the MOTHIF consortium, a wing model was designed, built and tested in the VKI’s large subsonic L1-A wind tunnel, reproducing bi-dimensional preliminary results obtained during the design phase. In order to reduce and predict three dimensional and blockage effects, CFD has been used extensively to obtain a proper test chamber configuration and to reproduce some wind tunnel test results, so that both the accuracy of wind tunnel and design methodology can be assessed. This work presents the comparison between experimental data and CFD simulations to demonstrate the usefulness of simulations to reduce both the risk for erroneous results in experimental activities and their related cost.

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.

The E-SCAPE (European SCAled Pool Experiment) facility at the Belgian nuclear research center SCK•CEN is a thermal hydraulic 1/6th-scale model of the MYRRHA reactor with an electrical core simulator and cooled by LBE. The E-SCAPE facility provides experimental feedback to the designers on the forced and natural flow patterns, flow mixing and stratification during operational and accidental conditions. In addition, it provides valuable data for validation of the computational techniques for their use with LBE and in the design of MYRRHA. In terms of momentum field, one of the most important integral quantity to reproduce is the pressure drop in the system which has a direct consequence on the evolution of the mass flows and determines the free-surface levels in transient scenarios. In this paper, a detailed CFD study is performed on the flow in the E-SCAPE core region with the aim on predicting the correct pressure drop. Different numerical representation of the core geometry have been modeled and subjected to validation against experimental results from E-SCAPE. Special attention has been addressed to the choice of geometric simplifications and numerical models for accurate representation of the core of the E-SCAPE pool facility.

For the safety assessment of Heavy Liquid Metal nuclear reactors under seismic excitation, sloshing phenomena can be of great concern. The earthquake motions are transferred to the liquid coolant which oscillates inside the vessel, exerting additional forces on the walls and internal structures. The present study examines the case of MYRRHA, a multi-purpose experimental reactor with LBE as coolant, developed by SCK·CEN. The sloshing behavior of liquid metals is studied through a comparison between mercury and water in a cylindrical tank. Experimental investigation of sloshing is carried out using optical techniques with the shaking table facility SHAKESPEARE at the von Karman Institute. Emphasis is given on the resonance case, where maximum forces occur on the tank walls. The experimental cases are reproduced numerically with the CFD software OpenFOAM, using the VOF method to track the liquid interface. The non-linear nature of sloshing is observed through visualization, where swirling is shown in the resonance case. The complex behavior is well reproduced by the CFD simulations, providing good qualitative validation of the numerical tools. A quantitative comparison of the maximum liquid elevation inside the tank shows higher values for the liquid metal than for water. Some discrepancies are revealed in CFD results and the differences are quantified. From simulations it is verified that the forces scale with the density ratio, following similar evolution in time. Overall, water is demonstrated to be a valid option as a working liquid in order to evaluate the sloshing effects, for forcing frequencies up to resonance.

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.

This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for accident mitigation.

International audience; Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Sodium and Liquid lead (-alloys) are considered the short and long term solution respectively, as coolant in GEN-IV reactor. Thermal-hydraulics of liquid metals plays a key role in the design and safety assessments of these reactors. Therefore, this is the main topic of a large European collaborative program (the Horizon 2020 SESAME) sponsored by the European Commission. This paper will present the progress in the project with respect to liquid metal cooled reactor thermal-hydraulics (liquid metal heat transfer, fuel assembly thermal-hydraulics, pool thermal-hydraulics, and system thermal-hydraulics). New reference data, both experimental and high-fidelity numerical data is being generated. And finally, when considering the system scale, the purpose is to validate and improve system thermal-hydraulics models and codes, but also to further develop and validate multi-scale approaches under development.

Sloshing can be a great concern for the seismic safety of heavy liquid metal cooled nuclear reactors, such as the Gen IV prototype MYRRHA, currently under development by the Belgian Nuclear Research Center (SCK•CEN). Sloshing is studied using reduced scale laboratory experiments on the SHAKESPEARE shaking table facility of the von Karman Institute. Scaling of the experimental model is discussed through dimensional analysis, identifying the appropriate scaling factors which are then applied to the seismic excitation signals. Qualitative results of the liquid sloshing motions inside the model are obtained with flow visualization, while moments are measured on an instrumented rod that is partially immersed in the liquid. A two component moment-balance is constructed to measure the bending moments on the element about the horizontal axes. The results demonstrate that sloshing, as a non-linear phenomenon, is highly dependent on the frequency of forcing relative to the natural frequency of the liquid in the specific container. In the resonance case the sloshing response reaches the highest amplitude and maximum moments are measured, representing a worst case scenario for the reactor safety. Experiments with internal components in the sloshing model indicate that obstructions reduce the sloshing loads and prevent resonance type sloshing. The proposed methodology with small scale experiments can provide a useful tool for the prediction of the sloshing effects for the MYRRHA design and safety analysis.

Performing good-quality computational fluid dynamics (CFD) simulations is not an easy task as the CFD engineer has to judge whether the results of the simulation can be trusted. While reality is continuous in space and time, computers work in the discrete domain. This chapter refers in the basis to generic existing CFD best practice guidelines but elaborates on the specific guidelines that arise from the simulation of liquid metals. The guidelines are provided following the structure of simulation, that is, preprocessing, simulation (or processing), and postprocessing. Finally, the most relevant guidelines specific for nuclear liquid-metal CFD simulations are summarized.

Following the Fukushima-Daiichi accident, many countries decided to strengthen the safety systems of their nuclear power plants in order to increase the capabilities of managing severe nuclear accidents. To achieve this goal detailed analysis of postulated design and beyond-design-basis accidents is essential. The main tools available for the analysis of these complex scenarios are Advanced Lumped Parameters (LP) and Computational Fluid Dynamics (CFD) codes. Currently, the most limiting factor in the application of these tools is their validation. This activity is performed in support of the development and validation of a CFD model, built in the numerical environment of OpenFOAM, for hydrogen behavior in the containment of Light Water Reactors. For validation purpose, numerical simulations of the erosion of a light gas stratification by means of a hot air jet have been performed, and results of the simulations were analysed and compared against experimental results obtained by Ritterath (2012) in the MiniPanda facility. Two different scenarios have been considered, each characterized by a different value of the mass flow rate of the jet used to erode the stratification.

La modélisation physique du circuit primaire d'un nouveau type de réacteur nucléaire apporte des informations importantes au niveau du comportement thermo-hydraulique. Dans le cadre des études R&D pour le réacteur MYRRHA, développé par le centre belge de l'énergie nucléaire (SCK-CEN), un modèle à eau à échelle réduite a été construit à l'Institut von Karman. Des mesures PIV caractérisent l'écoulement dans différents plans du modèle à eau. Pour obtenir des images de qualité dans un environnement présentant aussi peu d'accès optique, une méthodologie a été développée, basée sur l'utilisation d'images de calibration. L'utilisation d'un ensemencement à base de particules fluorescentes a permis également d'augmenter considérablement le rapport signal/bruit. Les résultats sont présentés pour deux modes de fonctionnement du modèle à eau: condition nominale et convection naturelle. Des informations importantes au niveau du champ de vitesse à l'entrée du réacteur et plus particulièrement sur l'importance de la composante radiale de la vitesse ont pu être extraites. Ces données servent à la validation des codes de simulation numérique.

The development of Gen IV nuclear reactors entails research oh heavy liquid metals, which are considered as appropriate coolant agents due to their high thermal conductivity and favorable nuclear safety characteristics. These fluids are characterized by a very low Prandtl number (between 0.006 and 0.025). During the design of such reactors, the scenarios of maintenance or loss-of-flow accidents consider a pure natural convection loop inside the reactor. The CFD modeling of the momentum and heat transfer must contemplate natural convection in low Prandtl number fluids, however in literature these flow configurations are not usual. This is why in the present article, Large Eddy Simulations over a natural convection boundary layer at Pr = 0.025 are presented and described. Turbulence is triggered by adding a perturbation in the boundary layer. The adopted numerical approach is validated through a Large Eddy Simulation at Pr = 0.71 , where heat transfer, friction coefficient and mean and turbulent flow fields are compared with measurements in literature. Once turbulence is reached in the Pr = 0.025 boundary layer, flow characteristics as Nusselt number, skin friction coefficient and turbulent profiles are calculated and compared with the Pr = 0.71 simulation to assess the effect of Prandtl number on heat and momentum transfer.

The Ahmed body is used as a reference model for fundamental studies of car-type bluff body aerodynamics, in particular focused on the influence of the rear slant angle on the drag coefficient. The objectives of the present work are to obtain reliable drag coefficient comparable to the literature and to explain, based on the nature of the flow, its variation when changing the rear slant angle from 10° to 40°. The drag coefficients measured in both an open and a closed test sections differ by less than 0.5% which proves the reliability and reproducibility of the results. The sensitivity of the drag coefficient to some parameters such as the model roughness or the oncoming boundary layer and the lack of precise information on these parameters in the literature could explain the difference observed with the Ahmed drag coefficient data. The various types of measurement techniques used in the study underline their complementarity. The combination of particle image velocimetry and oil visualization provides a deeper understanding of the flow behaviour around the Ahmed body and a physical interpretation of the drag coefficient evolution.

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

An experimental investigation is carried out on a semi-industrial prototype of regenerative roll quench system. It allows the determination of the thermal contact conductance between a roll and a moving strip. The application of the quantitative infrared thermography leads to the angular evolution of the strip temperature while the mean roll temperature is obtained by using thermocouples connected to a wireless telemetry data acquisition system. The experimental facility and the thermal model associated to the thermal conductance determination are presented. The results emphasize the effect of operating parameters such as the velocity, initial temperature and tensile strength of the strip. Modeling of the interfacial thermal conductance is finally proposed.

The POD-X program (Pressure Oscillation Demonstrator – eXperimental), co-funded by CNES French Space Agency and ESA FLPP to study Pressure Oscillations both on segmented and nonsegmented SRMs, is now approaching the first ground firing test (DM) to be performed at DGA-EM test bench near Bordeaux (France) in the last quarter of 2013. This first test shall be performed on a configuration reproducing the current Ariane 5 MPS P230 booster but scaled of a factor equal to 1/4,5; then, this first demonstrator is composed of three segments, as the MPS. This choice is due to the fact that a wider pressure oscillation database is available for the A5 MPS than for any other SRM manufactured in Europe and that many studies (including numerical simulations and small scale tests) have been dedicated to this motor. Nevertheless, the POD-X is conceived to reproduce the pressure oscillation behavior of other SRMs including monolithic motors as Ariane 6 SRMs configurations and the VEGA 1 stage motor P80 and its evolutions. This versatility is assured by the modular concept of the POD-X and by the fact that the reusable metallic structure is conceived for internal pressure much higher than the MPS one. The first goal of this program is to validate the subscale MPS design by

Large rocket motors, especially those with segmented grains, often exhibit pressure and thrust oscillations caused by coupling between large scale vortices (vortex shedding) and the acoustic modes of the chamber. The pressure oscillation of most concern in large motors is the longitudinal modes roughly corresponding to the first acoustic modes of the motor. The resultant thrust oscillations are a small percentage of the overall thrust level, but this excitation source can have major impact on the designs of the motor components, flight vehicle structure, avionics, and payload.

This paper deals with the mapping of the convective heat transfer in a multijet anti-icing system by application of the quantitative infrared thermography technique. The experiments are conducted on a full scale mock-up of the leading edge section of a jet aircraft slat. A steady state infrared thermography technique combined with the heating foil method is developed and successfully applied to recover the complete distribution of the thermal exchange coefficient on the concave inner surface of the skin. The heating performance of such a multijet system depends on the jet Reynolds number, the distance of the supply duct and the skin and the spanwise and chordwise jet arrangement. Tridimensional numerical simulations are also performed with the commercial CFD code FLUENT. Despite the complexity of the flow phenomena and the stiff character of the boundary conditions involved, satisfactory agreement is found between the predictions and the IR data.

The Belgian Nuclear Research Centre (SCK•CEN) is at the forefront of heavy liquid-metal (HLM) nuclear technology worldwide with the development of the Multi-purpose hYbrid Research Reactor for High...

The current paper describes the loss of flow (LOF) transient investigated in the MYRRHA reactor by the means of Computational Fluid Dynamics. This scenario is starting from the nominal operation case then the two pumps stop simultaneously. An unsteady solution with resolved interface was considered with calculating conjugate heat transfer through the relevant structures. Due to a postulated event (e.g. loss of the electric grid) the pumps are not powered anymore stops. After the detection of the problem (temperature difference above the core rises with 20 degree) the reactor power is stopped by the safety rods (delay of 1 s). The fuel elements, however, continue to generate residual heat according to the decay heat curve. Due to the loss of the pumps, the pressure difference between the cold and the hot plenum is decreasing, which result in a gravitational flow equilibrating the two free surfaces to the same level. The objective of the work was to determine the flow through the core during the coast down of the pumps and eventual flow reversal into the pump/heat-exchanger box due to the gravitational flow. The simulation revealed that after losing power, the LBE flow reverses into the pumps in less than 0.1 s according to the simulations. In the core there is a brief moment of reverse flow, too, but only after the core is scrammed, therefore, the loss of cold LBE flow is not causing overheat. Once the core is scrammed, the position of the maximum temperature in the system shifts to the Above Core Structure, where the residual hot plume rising from the core impinges to the Above Core Upper Closure. The levels of the lower and upper plenum equilibrate roughly 20 s after the pump failure event.

This article reports the experimental and DNS database that has been generated, within the framework of the EU SESAME and MYRTE projects, for various low-Prandtl flow configurations in different flow regimes. This includes three experiments: confined and unconfined backward facing steps with low-Prandtl fluids, and a forced convection planar jet case with two different Prandtl fluids. In terms of numerical data, seven different flow configurations are considered: a wall-bounded mixed convection flow at low-Prandtl number with varying Richardson number (Ri) values; a wall-bounded mixed and forced convection flow in a bare rod bundle configuration for two different Reynolds numbers; a forced convection confined backward facing step (BFS) with conjugate heat transfer; a forced convection impinging jet for three different Prandtl fluids corresponding to two different Reynolds numbers of the fully developed planar turbulent jet; a mixed-convection cold-hot-cold triple jet configuration corresponding to Ri=0.25; an unconfined free shear layer for three different Prandtl fluids; and a forced convection infinite wire-wrapped fuel assembly. This wide range of reference data is used to evaluate, validate and/or further develop different turbulent heat flux modelling approaches, namely simple gradient diffusion hypothesis based on constant and variable turbulent Prandtl number; explicit and implicit algebraic heat flux models; and a second order turbulent heat flux model. Lastly, this article will highlight the current challenges and perspectives of the available turbulence models, in different codes, for the accurate prediction of flow and heat transfer in low-Prandtl fluids. © 2019 American Nuclear Society. All rights reserved.