The vibration of fuel rods induced by the fast flowing coolant is a known challenge to nuclear power plant designers and operators. If not dealt with adequately, the vibrations could lead to undesirable wear and tear of the cladding, which, in turn, could result in fuel rod failures and costly unplanned outages of the power plant. Hence, knowledge of these vibrations is important in the design phase. Due to the increase in computational power, numerical tools are increasingly often used to assess flow-induced vibrations. As these vibrations are a result of the local axial turbulent flow, scale-resolving methods are typically required for accurate predictions. Such methods, though, are usually computationally too expensive to use for industrial nuclear applications. Medium resolution turbulence models such as URANS generally cannot be used as they only resolve the average flow conditions and not turbulent fluctuations. To overcome this limitation of the URANS models, an Anisotropic Pressure Fluctuation Model (AniPFM) has been developed, which is an improved version of the earlier developed isotropic version. Compared to the isotropic model the generated synthetic turbulence is now anisotropic and is correlated in time based on the transport and decorrelation of turbulence. The current paper gives an overview of the new AniPFM, and presents results for a first fluid–structure interaction test case, demonstrating it can indeed induce and sustain vibrations of a rod, and that results are improved compared to the old model.@en

Liquid metal reactors typically employ wire wraps as spacers between the fuel pins. In the past, design and safety calculations were largely one-dimensional and based on experimental data. Nowadays, with modern state-of-the-art computer power and tools, three-dimensional Computational Fluid Dynamics (CFD) simulations allow designers and safety specialists to obtain much more detailed information on the flow and heat transport in liquid metal cooled fuel assemblies, obviously in close collaboration with experimental campaigns. This may lead to new insights possibly decreasing the safety margins. This paper intends to provide an overview on the activities in the frame of design and safety support for wire-wrapped fuel assemblies. It all starts with validation. Therefore, validation efforts will be shown for fuel assemblies as they are designed on the drawing board for ‘cold’ conditions. Such analyses will profit from the quantification of uncertainties and determination of most influencing parameters. Nevertheless, in reality a fuel assembly will not be employed as designed in ‘cold’ conditions. Therefore, they will probably deform. This requires an assessment of the effect of deformations. Another aspect possibly occurring during operational conditions is vibrations. State-of-the-art coupled CFD and finite element method fluid structure interaction techniques have been developed and applied to a wire wrapped fuel pins, providing insights in the vibration behavior of such assemblies. Apart from that, vibration experiments have been performed at complete fuel assembly scale providing important insights to safety analysts and designers. However, design and safety analysts will not only have to cope with operational conditions, but also have to show the heat transport behavior under accident conditions. Assessments of the effect and formation of blockages are necessary. In all above cases, it should be clear that experiments and numerical simulations go hand-in-hand. Numerical simulations are used to design the experiment, the experiment is used to validate the simulations, and the simulations are used to interpret the experimental results.@en

In nuclear fuel rod bundles, turbulence-induced pressure fluctuations caused by an axial flow can create small but significant vibrations in the fuel rods, which in turn can cause structural effects such as material fatigue and fretting wear. Fluid-structure interaction simulations can be used to model these vibrations, but for affordable simulations based on the URANS approach, a model for the pressure fluctuations must be utilised. Driven by the goal to improve the current state-of-the-art pressure fluctuation model, AniPFM (Anisotropic Pressure Fluctuation Model) was developed. AniPFM can model velocity fluctuations based on anisotropic Reynolds stress tensors, with temporal correlation through the convection and decorrelation of turbulence. From these velocity fluctuations and the mean flow properties, the pressure fluctuations are calculated. The model was applied to several test cases and shows promising results in terms of reproducing qualitatively similar flow structures, as well as predicting the root-mean-squared pressure fluctuations. While further validation is being performed, the AniPFM has already demonstrated its potential for affordable simulations of turbulence-induced vibrations in industrial nuclear applications. @en

Liquid metal fast reactors (LMFRs) are foreseen to play an important role in the future of nuclear energy, thanks to their increased fuel utilization and safety features profiting from the optimal heat transfer performance of the metallic coolants. Accurate thermal-hydraulic analysis of their fuel assemblies, typically employed with wire-wraps as spacers, is recognized as a crucial scientific and engineering contribution to support the deployment of such technology. This challenges the modeling and simulation community. To this aspect, various reference databases (both experimental and numerical) for different wire-wrapped fuel assembly configurations have been created recently and are being used for validation of engineering simulation approaches based on Reynolds Averaged Navier Stokes (RANS) modelling. These databases include: • 7-pin rod bundle: A detailed experiment with Particle Image Velocimetry (PIV) is performed. In order to allow accurate measurements of the flow topology, a matched-index-of-refraction technique was used employing water as working fluid. • 19-pin bundle: A series of experiments is performed covering a wide range of Reynolds and Peclet numbers as well as thermal powers. The experiments use liquid lead-bismuth eutectic as working fluid. The measurements include pressure drop and local temperatures. • 61-pin rod bundle: This large eddy simulation including conjugate heat transfer from the pin cladding to the coolant allows to bridge the gap from small bundles (less than 37 pins) to large bundles (more than 37 pins). In literature, a fundamental different behavior has been observed for small bundles compared to large bundles. • 127-pin bundle: Isothermal experiments using lead-bismuth eutectic characterizing pressure drop are performed on a full scale fuel assembly representative for the MYRRHA reactor. • Infinite pin bundle: This reference quasi-direct numerical simulation profits from periodicity in all directions. It provides a detailed view into the flow field and in addition reveals details of the heat transfer from the rod bundle into the flow.Reference databases aim to serve the nuclear scientific community to validate engineering simulation approaches. The paper will introduce these reference databases, and how they have been used to validate RANS based turbulence modelling approaches within a mainly European context.@en