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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...

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.

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.

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.

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.

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.

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 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 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.