The PALLAS-reactor is an advanced nuclear reactor designed for producing medical isotopes and for carrying out research on (medical) nuclear technology. RELAP5 is the primary tool for modeling the accident scenarios relevant to the licensing of this reactor. The Groeneveld Look-Up Tables (LUT) are the default model of RELAP5 for estimating Critical Heat Flux (CHF). Literature has shown that the LUT over-predict CHF inside narrow, rectangular channels such as the cooling channels of the core of the PALLAS-reactor. Empirical correlations such as that of Sudo-Kaminaga or its modified version by Kim et al. (SK-Kim) are better suited for these geometries. However, these correlations are typically implemented in one-dimensional (1D) System Thermal-Hydraulics codes whereby the quantities of interest are averaged-out over the control domains. Computational Fluid Dynamics (CFD) could capture three-dimensional effects that could help reducing unnecessary conservatism embedded in some 1D approaches (e.g. Sudo-Kaminaga or SK-Kim correlations). Therefore, CFD is to be employed to support the Deterministic Safety Analyses of the PALLAS-reactor if there are scenarios in which the SK-Kim correlation predicts CHF. Boiling flows inside round or annular vertical channels have been extensively modeled with CFD. However, attempts at modeling CHF in narrow, rectangular geometries are still scarce. Therefore, this work aims at developing and validating a CFD approach for predicting CHF in rectangular channels. Four combinations of turbulence interaction and turbulent dispersion models are investigated; the results are compared against experiments and the SK-Kim correlation; sensitivity analyses are conducted on the turbulence models as well as the mesh. Certain combinations of turbulence interaction and turbulent dispersion models promote the transport of steam towards the center of the channel, thus increasing CHF, while other cause the steam to remain confined at the wall, thereby lowering the CHF. This study provides several CFD options to estimate CHF depending on the required level of conservatism. Estimating CHF in a conservative manner is crucial for licensing the PALLAS-reactor and other reactors of similar design, thereby securing the production of medical isotopes in Europe and the rest of the world.

In recent years, there has been renewed interest in molten salt reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, SPECTRA and SAM are two system analysis codes that have been enhanced to include MSR-specific modeling capabilities, including delayed neutron precursor drift and modified point kinetics equations.This paper discusses the efforts taken to verify and validate these features. A standard MSR system test problem was developed to verify and demonstrate the capability of SPECTRA and SAM on the MSR transient simulation. Sixteen transients were simulated. The results obtained from SPECTRA and SAM show good agreement. The Molten Salt Reactor Experiment transient experiments were reviewed and selected to validate the SPECTRA and SAM codes. The experiments included pump startup and coastdown tests at zero power, reactivity insertion tests at different power levels, frequency tests, and a natural convection test. The simulation results from SPECTRA and SAM show good agreement with the experimental data.

The PALLAS-reactor is an advanced nuclear reactor dedicated to the production of medical isotopes and (medical) nuclear technological research currently under construction in The Netherlands RELAP5 is the primary tool used for modeling the accident scenarios relevant for licensing the PALLAS-reactor. The Groeneveld Look-Up Tables (LUT) are the main Critical Heat Flux (CHF) model of RELAP5. These have been extensively validated against experiments with round tubes. However, literature suggests that the LUT over-predict CHF inside the narrow, rectangular fuel channels of research reactors such as the PALLAS-reactor. As literature abounds with CHF correlations of limited applicability and low degrees of validation and confidence, correlations that can accurately predict CHF in rectangular channels are needed. The goal of this work is twofold. First, it aims at selecting a number of promising CHF correlations and to assess these against experiments representative of the PALLAS-reactor design. Secondly, this work ascertains whether the LUT are indeed optimistic with respect to CHF in rectangular channels. To this purpose, a literature review has short-listed the Sudo-Kaminaga (S-K), the Kim-modified S-K, and the Hall–Mudawar (H-M) correlations. These correlations are implemented in the SPECTRA code developed at NRG and assessed against two experiments in rectangular channels. The Kim-modified S-K correlation fits the experiments better than the original S-K scheme for high mass fluxes, while predicting higher CHF values at low flow rates. Interestingly, the H-M correlation predicts the experiments accurately, although the correlation was developed from round tubes data. At high mass fluxes, the LUT predictions converge towards the experiments, thereby suggesting that the effect of the channel geometry becomes less important at high flow rates. At low mass fluxes, on the other hand, the LUT markedly fail to match the experiments. This work contributes to accurately estimating CHF in narrow rectangular channels, which is crucial for licensing research reactors such as the PALLAS-reactor, thus assisting in the betterment of many people’s lives.

Fluid flows through rod bundles are observed in many nuclear applications, such as in the core of Gen IV liquid metal fast breeder nuclear reactors (LMFBR). One of the main features of this configuration is the appearance of flow fluctuations in the rod gaps due to the velocity difference in the sub-channels between the rods. On one side, these pulsations are beneficial as they enhance the heat exchange between the rods and the fluid. On the other side, the fluid pulsations might induce vibrations of the flexible fuel rods, a mechanism generally referred to as Flow Induced Vibrations (FIV). Over time, this might result in mechanical fatigue of the rods and rod fretting, which eventually can compromise their structural integrity. Within the SESAME framework, a joint work between Delft University of Technology (TU Delft), Ghent University (UGent), and NRG has been carried out with the aim of performing experimental measurements of FIV in a 7-rods bundle and validate numerical simulations against the obtained experimental data. The experiments performed by TU Delft consisted of a gravity-driven flow through a 7-rods, hexagonal bundle with a pitch-to-diameter ratio P/D=1.11. A section of 200 mm of the central rod was made out of silicone, of which 100 mm were flexible. Flow measurements have been carried out with Laser Doppler Anemometry (LDA) whereas a high-speed camera has measured the vibrations induced on the silicone rod. The numerical simulations made use of the Unsteady Reynolds-averaged Navier-Stokes equations (URANS) approach for the turbulence modelling, and of strongly coupled algorithms for the solution of the fluid-structure interaction (FSI) problems. The measured frequency of the flow pulsations, as well as the mean rod displacement and vibration frequency, have been used to carry out the benchmark.

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.

The core of a Liquid Metal Fast Breeder Reactor (LMFBR) consists of cylindrical fuel rods that are wrapped by a helicoidally-wound wire spacer to enhance mixing and to prevent damage by fretting. It is known that the liquid metal close to the rod is forced to follow the wires, and that liquid metal further away from the rod crosses the wires (called: migratory flow). This work aims at gaining more insight into the physics behind migratory flow and to provide a model for its bending angle. To this purpose, the flow field in a 7-rods, wire-wrapped, hexagonal bundle with water is studied within the Reynolds number range of 4990-16330 by using Particle Image Velocimetry (PIV). Refraction of the light is minimized by using Fluorinated Ethylene Propylene (FEP), which is a refractive index-matching (RIM) material. These measurements confirm that liquid near the rod follows the helicoid path and bends cross-wise with respect to the wire further away from the rod. A theoretical model for the bending angle of the flow is derived from the Euler equations and shows that the bending is primarily caused by the pressure gradient field induced by the wire. The model shows a very good correspondence with the experimentally obtained PIV data. These findings improve our understanding of the physics at play in rod bundle flows with wrapped wires and can be of assistance in developing practical correlations for frictional pressure losses and heat transfer in such bundles.

Being able to quantify mechanical vibrations is of key importance for the safety of nuclear power plants, as they are able to induce damage. In this work, numerical simulations are used to compute water flow and vibration in a densely packed bundle of 7 rods, mimicking an experimental setup. This flow configuration is chosen to resemble the coolant flow through a nuclear reactor core. Because of the wall proximity, a considerable velocity difference between the narrow gaps and the subchannels exists, with an inflection point in the velocity profile. This yields an unstable situation, and large vortices are continuously created through a mechanism similar to the Kelvin-Helmholtz instability. The vortex streets in between the rods are associated with a fluctuating pressure field, causing vibrations of the rods. The experimental setup contains 7 steel cylinders, encased in a hexagonal duct. The central rod contains a section where the steel is replaced by a water-filled silicone tube, clamped at both extremes to the steel rod, and the vibrations of this section are examined. The numerical approach consists of coupled fluid-structure interaction (FSI) simulations, with the flow being modelled using computational fluid dynamics (CFD) and the structure using computational solid mechanics (CSM). The available experimental data consist of Laser Doppler Anemometry (LDA) measurements and high-speed camera footage of the wall movement of the silicone rod. Equivalent data is collected from the numerical simulations. The simulations are repeated for different flow rates. The frequency spectrum of the coherent structures, and the frequency and amplitude of the wall movement are compared for each operating point, as well as their trend as a function of the flow rate. The dominant frequencies found in the simulation results were similar to the experimental results, although slightly higher. They also showed a linear trend, just like the experiments. A larger mismatch was present for the structural response, the frequencies found using the FSI model being more than twice as high.

Being able to quantify mechanical vibrations is of key importance for the safety of nuclear power plants, as they are able to induce damage. In this work, numerical simulations are used to simulate water flow through a densely packed bundle of 7 rods, mimicking an experimental setup used at Delft University of Technology. This flow configuration is chosen to resemble the coolant flow through a nuclear reactor core. Because of the wall proximity, a considerable velocity difference between the narrow gaps and the subchannels exists, with an inflection point in the velocity profile. This yields an unstable situation, and large vortices are continuously created through a mechanism similar to the Kelvin–Helmholtz instability. The vortex streets in between the rods are associated with a fluctuating pressure field, causing so-called vortex-induced vibrations of the rods. These vibrations are well-known to cause damage to structures in cross-flow. In this case they are of smaller amplitude, but could still cause fretting or fatigue damage. The experimental setup contains 7 steel cylinders, encased in a hexagonal duct. The central rod contains a section where the steel is replaced by a water-filled silicone tube, clamped at both extremes to the steel rod, and the vibrations of this section are examined. The numerical approach consists of coupled fluid-structure interaction (FSI) simulations, with the flow being modelled using computational fluid dynamics (CFD) and the structure using computational solid mechanics (CSM). For the CFD-part, the software package ANSYS Fluent is used, implementing the finite volume method. A deformable grid is applied with the Arbitrary Eulerian–Lagrangian (ALE) formulation. The k-ω SST model is employed for the turbulence. Streamwise periodic boundary conditions are used. As the flow is then inherently developed, this allows for a great reduction in domain length, keeping the simulation computationally manageable. Structural simulations are performed by the Abaqus software, implementing the finite element method. Quadratic elements are used for the silicone part. An in-house code manages the coupling, using the quasi-Newton algorithm IQN-ILS. The available experimental data consist of Laser Doppler Anemometry (LDA) measurements and high-speed camera footage of the wall movement of the silicone rod. Equivalent data is collected from the numerical simulations. The simulations are repeated for different flow rates. The frequency spectrum of the coherent structures, and the frequency and amplitude of the wall movement are compared for each operating point, as well as their trend as a function of the flow rate. The dominant frequencies found in the simulation results were similar to the experimental results, although slightly higher. They also showed a linear trend, just like the experiments. A larger mismatch was present for the structural response, the frequencies found using the FSI model being more than twice as high.

Gap vortex streets characterise many industrial applications involving rod bundle flows, such as heat exchangers and nuclear reactors. These structures, known as gap vortex streets, may excite the structural components of the bundle to resonance, leading to fretting and fatigue. This work aims to measure these coherent structures and the resulting displacement and oscillation frequency of the neighbouring rod, to provide unique data for fluid structure interaction studies and to develop a general correlation for estimating the coherent structure's wavelength. A water loop was built to host a hexagonal rod bundle. Fluorinated Ethylene Prophylene (FEP), a refractive index matching (RIM) material, was used to have undisturbed optical access in the area around the central rod. The flow was measured with Laser Doppler Anemometry (LDA) to detect coherent structures, while the vibrations were measured with a high speed camera. A new correlation for estimating the wavelength of the coherent structures is derived with dimensional analysis based on experimental evidence. The correlation is tested on different geometries: rectangular channels with single or half-rods, and two rod bundles, within the pitch-to-diameter ratio (P/D) range 1.02-1.2. Moreover fluctuations in the flow, given by the detected coherent structures, govern the structural response of the rod. The rod is excited to resonance if these fluctuations match twice the natural frequency of the rod.

Axial flow in tube bundles with small pitch-to-diameter ratio, a geometry encountered in nuclear reactor cores and heat exchangers, often displays periodic fluctuations. A significant velocity discrepancy between the inter-cylinder gap and subchannel center originates from the difference in through-flow area, feeding an instability. As it is associated with velocity-shear, it is similar to the Kelvin-Helmholtz type and the term 'gap instability' is adopted. A vortex street arises and structural vibration of the cylinders might develop due to the fluctuating pressure. Numerical simulations of this phenomenon were performed. The computational domain was constructed to match the most important geometrical features of an experimental setup. The bundle consists of 7 steel tubes in triangular array, placed in a hexagonal conduit. A flexible segment made of silicone is embedded in the central tube, with both extremes clamped to the steel parts of the cylinder. In the experiment, data of the fluctuating velocity was gathered using laser Doppler anemometry measurements. As first step, a completely rigid structure was considered. Unsteady Reynolds-averaged Navier-Stokes (URANS) simulations were used to test if this particular geometry also triggers the gap vortex street, which was the case. The phenomenon clearly appears as oscillations of the velocity components. Subsequently, fluid-structure interaction (FSI) simulations, taking into account the flexible part, allowed to assess the effect of the fluctuating flow field on the structure. A comparison between one-way and two-way coupled simulations was made.

The enhancement of heat transfer from fuel rods to coolant of a Liquid Metal Fast Reactor (LMFR) decreases the fuel temperature and, thus, improves the safety margin of the reactor. One of the mechanisms that increases heat transfer consists of large coherent structures that can occur across the gap between adjacent rods. This work investigates the flow between two curved surfaces, representing the gap between two adjacent fuel rods. The aim is to investigate the presence of the aforementioned structures and to provide, as partners in the EU SESAME project, an experimental benchmark for numerical validation to reproduce the thermal hydraulics of Gen-IV LMFRs. The work investigates also the applicability of Fluorinated Ethylene Propylene (FEP) as Refractive Index Matching (RIM) material for optical measurements. The experiments are conducted on two half-rods of 15 mm diameter opposing each other inside a Perspex box with Laser Doppler Anemometry (LDA). Different channel Reynolds numbers between Re = 600 and Re = 30,000 are considered for each P/D (pitch-to-diameter ratio). For high Re, the stream wise velocity root mean square vrms between the two half rods is higher near the walls, similar to common channel flow. As Re decreases, however, an additional central peak in vrms appears at the gap centre, away from the walls. The peak becomes clearer at lower P/D ratios and it also occurs at higher flow rates. Periodical behaviour of the span wise velocity across the gap is revealed by the frequency spectrum and the frequency varies with P/D and decreases with Re. The study of the stream wise velocity component reveals that the structures become longer with decreasing Re. As Re increases, these structures are carried along the flow closer to the gap centre, whereas at low flow rates they are spread over a wider region. This becomes even clearer with smaller gaps.