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.

The inter-wrapper flow (IWF) between fuel assemblies in an LMFR plays an important role in the reduction of the peak cladding temperature in abnormal scenarios, such as blockages or flow reductions. This paper focusses on the heat transfer through the IWF in a geometry which is representative for the MYRRHA reactor under design by SCK.CEN in Belgium. To this purpose an LBE experiment is designed in the Karlsruhe Lead LAboratory (KALLA) at KIT in Germany. Three 7-pin wire-wrapped rod bundles, creating a rotating flow representative for a MYRRHA wire-wrapped fuel assembly, and the adjacent inter-wrapper region form the experimental setup. The thickness of the wrapper and the region for the inter-wrapper flow have the same dimensions as selected for MYRRHA. The instrumentation is designed such that the temperatures and heat fluxes can be measured and used to validate Computational Fluid Dynamics (CFD) models for the inter-wrapper flow. This validation will allow CFD models to be applied in full core models for liquid metal cooled reactors. This paper concerns the experiment design and the supporting CFD calculations. The influence of the equipment upstream of the test section is assessed to determine the domain size and boundary conditions required for the code validation simulations. The selected domain and boundary conditions are subsequently applied in simulations to determine the most suitable locations for the instrumentation to measure the most interesting phenomena of the heat transfer through the inter-wrapper region and to be suitable for CFD validation.

Early studies of the flow in rod bundles with spacer grids suggest that the pressure drop can be decomposed in contributions due to flow area variations by spacer grids and frictional losses along the rods. For these shape and frictional losses simple correlations based on theoretical and experimental data have been proposed. In the OECD benchmark study LACANES it was observed that correlations could well describe the flow behavior of the heavy liquid metal loop including a rod bundle with the exception of the core region, where different experts chose different pressure-loss correlations for the losses due to spacer grids. Here, RANS–CFD simulations provided very good data compared to the experimental data. It was observed that the most commonly applied Rehme correlation underestimated the shape losses. The available correlations relate the pressure drop across a grid spacer to the relative plugging of the spacer i.e. solidity emax. More sophisticated correlations distinct between spacer grids with round or sharp leading edge shape. The purpose of this study is to (i) show that CFD is suitable to predict pressure drop across spacer grids and (ii) to access the generality of pressure drop correlations. By verification and validation of CFD results against experimental data obtained in KALLA we show (i). The generality (ii) is challenged by considering three cases which yield identical pressure drop in the correlations. First we test the effect of surface roughness, a parameter not present in the correlations. Here we compare a simulation assuming a typical surface roughness representing the experimental situation to a perfectly smooth spacer surface. Second we reverse the flow direction for the spacer grid employed in the experiments which is asymmetric. The flow direction reversal is chosen for convenience, since an asymmetric spacer grid with given blockage ratio, may result in different flow situations depending on flow direction. Obviously blockage develops differently along flow direction due to the asymmetry, so that the same numerical mesh can be used for two distinct simulations. The study shows that prediction of pressure drop across spacer grids by RANS–CFD using well resolved meshes in the spacer grid region yields more accurate results than any available correlation.

Heat transfer in supercritical water reactors (SCWRs) shows a complex behavior, especially when the temperatures of the water are near the pseudocritical value. For example, a significant deterioration of heat transfer may occur, resulting in unacceptably high cladding temperatures. The underlying physics and thermodynamics behind this behavior are not well understood yet. To assist the worldwide development in SCWRs, it is therefore of paramount importance to assess the limits and capabilities of currently available models, despite the fact that most of these models were not meant to describe supercritical heat transfer (SCHT). For this reason, the Gen-IV International Forum initiated the present blind, numerical benchmark, primarily aiming to show the predictive ability of currently available models when applied to a real-life application with flow conditions that resemble those of an SCWR. This paper describes the outcomes of ten independent numerical investigations and their comparison with wall temperatures measured at different positions in a 7-rod bundle with spacer grids in a supercritical water test facility at JAEA. The wall temperatures were not known beforehand to guarantee the blindness of the study. A number of models have been used, ranging from a one-dimensional (1-D) analytical approach with heat transfer correlations to a RANS simulation with the SST turbulence model on a mesh consisting of 62 million cells. None of the numerical simulations accurately predicted the wall temperature for the test case in which deterioration of heat transfer occurred. Furthermore, the predictive capabilities of the subchannel analysis were found to be comparable to those of more laborious approaches. It has been concluded that predictions of SCHT in rod bundles with the help of currently available numerical tools and models should be treated with caution.

Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are:Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena.Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena.Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems.This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue. (C) 2014 Elsevier B.V. All rights reserved.

Thermal-hydraulics is a key scientific subject to be investigated for the development of innovative reactor systems. For applications using liquid metals as coolants this task is particularly challenging due to their very low Prandtl number (Pr), preventing the application of common analogies between the turbulent transport of momentum and heat. Thus specific models and validation data with low-Pr fluids are required for representing safety-related thermal-hydraulic scenarios, such as heat transfer in fuel assemblies.Aiming to achieve a better understanding of these flow scenarios, in the European FP7 cooperation project THINS (2010-2014) this subject is investigated at three complementary levels. An experimental campaign consisting of an electrically heated 19-pin hexagonal rod bundle cooled by lead-bismuth eutectic (LBE) is carried out at typical reactor conditions in terms of operating temperature, power density and velocity. Both pre- and post-test analyses using existing numerical tools are performed for evaluating the differential pressure and heat transfer characteristics of the test section. Moreover, advanced turbulence models and numerical techniques are developed and applied to this geometry.Overall, the goals of this project are achieved. The experiments show good degree of repeatability and provide reliable validation data. For intermediate flow rates a good agreement is observed with the results of the heat transfer simulations, based on a constant turbulent Prandtl number. Two advanced approaches for representing the turbulent heat transport considering look-up tables and a four-equation model are successfully tested and overcome the limitations of using a constant turbulent Prandtl number. Using a coarse-grid CFD approach the turbulent momentum transport along two bundles is studied, yielding a good accuracy with a 1000-fold mesh reduction. (C) 2014 Elsevier B.V. All rights reserved.

A prototype of a heavy liquid metal free surface target as proposed for the multi-purpose hybrid research reactor for high-tech applications in Mol, Belgium, has been set up and experimentally investigated at the Karlsruhe Liquid Metal Laboratory. A stable operation was demonstrated in a wide range of operating conditions and the surface shape was detected and compared with numerical pre-calculations employing Star-CD. Results show a very good agreement of experiment and numerical predictions which is an essential input for other windowless target designs like the META:LIC target for the European Spallation Source.

Accelerator Driven Systems (ADS) are extensively investigated for the transmutation of high-level nuclear waste within many worldwide research programs. The first advanced design of an ADS system is currently developed in SCK.CEN, Mol, Belgium: the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA). Many European research programs support the design of MYRRHA. In the framework of the Euratom project 'Thermal Hydraulics of Innovative nuclear Systems (THINS)' a liquid-metal free-surface experiment is performed at the Karlsruhe Liquid Metal Laboratory (KALLA) of Karlsruhe Institute of Technology (KIT). The experiment investigates a full-scale model of the concentric free-surface spallation target of MYRRHA using Lead Bismuth Eutectic (LBE) as coolant. In parallel, numerical free surface models are developed and tested which are reviewed in the article. A volume-of-fluid method, a moving mesh model, a free surface model combining the Level-Set method with Large-Eddy Simulation model and a smoothed-particle hydrodynamics approach are investigated. Verification of the tested models is based on the experimental results obtained within the THINS project and on previous water experiments performed at the University Catholic de Louvain (UCL) within the Euratom project 'EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in Accelerator Driven System (EUROTRANS)'. The design of the target enables a high fluid velocity and a stable surface at the beam entry. The purpose of this paper is to present an overview of both experimental and numerical results obtained for free surface target characterization. Without entering in technical details, the status, the major achievements and lessons for the future with respect to model development are described as well as some applications, which were carried out within the work package 'multi-phase flow' of THINS. (C) 2014 Elsevier B.V. All rights reserved.

The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-E model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-E) does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ D-H =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons.However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10(-2) requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable.Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly. (C) 2011 Elsevier B.V. All rights reserved.

As highly promising coolant for new generation nuclear reactors, liquid Lead–Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations – ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI – contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal–hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

The objective of the European 6th framework Integrated Project (IP) EUROTRANS (EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System) is to demonstrate the feasibility of transmutation of high level nuclear waste using subcritical Accelerator Driven Systems (ADS). The spallation target represents the most challenging new component in an ADS since it is the component coupling the accelerator and the nuclear core and is subjected to very high thermal load in a high radiation field. In this document the thermal hydraulic activities which led to reliable design rules for a windowless target are presented and the status of the heavy liquid metal target mock-up experiment at the KArlsruhe Liquid metal LAboratory (KALLA) are reported.

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELLOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER(2)) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.