Liquid Metal Loop Research Articles

Design and construction of GaInSn experimental facility for studies of mixed-convection MHD flows

A multifunctional liquid metal loop named MaTHE-XJTU (Magneto-Thermo-Hydrodynamic Experiments-Xi'an Jiaotong University) that utilizes eutectic alloy GaInSn as a working fluid has been designed and constructed. The function of the MaTHE-XJTU facility is to study the magnetohydrodynamic (MHD) flow and mixed convection characteristics under the coupling effect. The main operating parameters of the loop are: the maximum magnetic field intensity is 3T, the effective magnetic field region is 300 mm × 800 mm × 1000 mm, the maximum flow rate of electromagnetic pump (EM pump) is 8 m3/h, the maximum pressure head of EM pump is 0.5 MPa. The paper describes the major components and basic operation procedures of the loop, the related flow diagnostics method, and near-future experiments. This loop could provide a high-parameter experimental platform (Ha∼104, Gr∼109, Re∼104) for investigations that improve the present understanding of magnetohydrodynamic and heat transfer performance in liquid metal blankets.

Network Graph Laplacian-Based Sensor Projection in System-Level Modeling of Liquid-Metal Loop

Accurate modeling of nuclear reactor transients is important for improving operation and maintenance planning strategies in advanced reactors. An adequate understanding of the connectivity between various components and sensor signals can provide the means to project plant information even where sensors are not operational. This work presents a method for projecting temperature signals using network graph Laplacian and support vector regression (SVR). A system model of a scaled liquid-metal-cooled reactor experiment facility is used to generate thermal-hydraulic data simulating cold-shock transients into the reactor plenum. Two different flow rates are investigated, 0.03 kg/s and 0.06 kg/s, describing the different degrees of nonlinear disturbances in two cases. The temperatures at 72 locations within the system are used to test the temperature prediction model. The inter-connectivity of these points is described using the network graph Laplacian, which describes a location in the loop as a node and then comprises a weighted matrix of connections between each of these nodes. Additional nodes, called ghost nodes, are used to model the heat transfer of the loop with the surroundings. The graph Laplacian and corresponding nodal sensing data are then used to train a kernel, which can inform the impact of various phenomena at one location in the plant on phenomena at other locations in the plant. The sensing data and graph network of these nodes are used to construct a surrogate model of the liquid-metal loop. This model is then used to predict the behavior at certain nodes where the training data are not provided to the model. These model predictions are compared against the test data for the two different inlet flow rates. When optimized, the average error of the simulated temperature data remained below 5% when 3 of the 72 nodes are predicted using the graph-based SVR model. As the number of unknown sensors is increased, the root-mean-square error increases slightly but still remains below 2% when 24 of the 72 nodes are unknown or tested. The test or unknown sensor location plays a larger role than the number of unknowns, with sensor locations near ghost nodes and near the outlet pipes in the plenum having the largest error, with a maximum recorded error of 7%. The objective of the graph-based SVR model is to not only capture the temperature or field variable accurately but also to capture the relative connectivity between the simulated sensors. The correlation coefficient matrix provides a scaled reference for the correlation between temperatures at two different node locations. The calculated correlation coefficient for the simulated temperature data and actual temperature data is within 5% for most of the system, with a maximum relative error of 15%.

Two-stage multichannel liquid–metal cooling system for thermal management of high-heat-flux-density chip array

A two-stage multichannel liquid–metal cooling system for a high-heat-flux-density chip array is investigated in this study. This system exhibits high cooling performance, low cost, excellent temperature uniformity, and high reliability. The first stage, which comprises a multichannel liquid metal convection subsystem, is implemented in the high-heat-flux-density chip area to achieve high cooling performance. The second stage, which comprises a water convection subsystem, is utilized in the low-heat-flux-density area to effectively reduce the system cost. Because the liquid–metal loops are equipped with independent electromagnetic pumps for each chip, the chip temperature can be conveniently adjusted, thereby guaranteeing good temperature uniformity. The experimental results show that the proposed liquid–metal cooling system can accommodate a heat flux of 50–200 W/cm2 with a convective heat transfer coefficient exceeding 20 000 W/(m2·K). Moreover, the flow of liquid metal can be continuously adjusted by varying the input current of the electromagnetic pump to ensure a chip array temperature uniformity of approximately 1 °C. The proposed liquid–metal cooling system can serve as a high-performance solution for the thermal management of high-heat-flux-density chip arrays in data centers.

Experiment and Codes to Support Safety Assessments for Sodium Fast Reactors (KASOLA, SOLTEC and KARIFA)

Abstract The experimental liquid metal loops hosted within the Karlsruhe Sodium laboratory (KASOLA) comprise a set of facilities to study liquid metal flows for various types of energy applications ranging from room temperature conditions used for education and training and fundamental research up to challenges posed by multiphysics problems such as material–fluid interactions at high temperatures. Extreme conditions, such as sodium boiling, relevant to thermo-electric conversion or fast reactor safety are covered in a dedicated small-scale facility (Karlsruhe receiver test facility (KARIFA)). The complete experimental range is complemented by system code support and computational fluid dynamics (CFD) simulation. The outcome of the experiments is used for code validation and development allowing application not only on component but also on system level. The paper presents main experimental facilities and supporting computational codes used at KASOLA.

Modeling the lithium loop in a liquid metal pool-type divertor

Considering that solutions for the steady-state power exhaust problem in future fusion reactors (e.g. DEMO) are not provided by present experiments and it is uncertain if they will be provided by ITER, because the expected heat fluxes, as well as the level of neutron irradiation, will be much higher, dedicated work packages are being devoted to this problem within EUROfusion and even a dedicated facility (the Divertor Tokamak Test − DTT) is being proposed in Italy. Among the possible options, a liquid metal (LM) divertor is being considered. The present work aims at developing a simple model of the LM loop in the case of a pool-type divertor, including the most important physical phenomena and allowing to roughly determine the operating range of the system, in terms of surface temperatures and vapor pressures. This work therefore sets a preliminary basis for the conceptual design of a LM divertor for the DTT facility. The model includes the incoming plasma heat load and a basic treatment of the interactions of the Li vapor with the plasma. The reduction of the Li vapor efflux due to ionization by the plasma is also taken into account. The model includes two chambers: a first divertor box, the evaporation chamber (EC), is open towards a second divertor box, the differential chamber (DC), which is in turn connected to the main plasma chamber (MC). The model is used to study the effectiveness of the LM vapor in radiating isotropically the parallel heat flux incoming in the divertor. The results indicate that the presence of the DC allows a significant reduction of the Li vapor efflux towards the MC, which in turn would imply a lower contamination of the core plasma.

Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8kW to 33.6kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

Determination of the Heat-Exchange Characteristics of Low-Pressure Heat-Exchangers for Lead and Lead-Bismuth Cooled Loops

Heat removal during transient processes in nuclear power installations and bench loops cooled by lead and lead-bismuth coolant is studied. It is shown that low-pressure heat-exchange equipment is effective. The studies were conducted in a non-isothermal loop with circulating of lead coolant at 450–500°C, average speed in the narrow section of the gap in longitudinal flow over the heat-exchange surface 0.1–1.05 m/sec, Peclet number 300–3300, and the oxygen thermodynamic activity in lead 1–10–3. The processes occurring in a non-isothermal liquid-metal loop with an experimental heat-exchange section model the dependence of the heat-exchange characteristics on impurity mass transfer in the loop.

CFD analysis of pressure drop across grid spacers in rod bundles compared to correlations and heavy liquid metal experimental data

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.

Natural circulation studies in a LBE loop for a wide range of temperature

Lead–Bismuth Eutectic (LBE) is increasingly getting more attention as a coolant for advanced reactor systems. It is also the primary coolant of the Compact High Temperature Reactor (CHTR) being designed at Bhabha Atomic Research Centre (BARC). A high temperature liquid metal loop named as Kilo Temperature Loop (KTL) has been installed at BARC for thermal hydraulics, instrument development and material related studies. Natural circulation experimental studies were carried out for the power range of 200–1200W in the loop. The corresponding LBE flow rate is calculated to be in the range of 0.075–0.12kg/s. Transient studies for start-up of natural circulation in the loop, loss of heat sink and step power change have also been carried out. The maximum temperature of the loop operated so far is 1100°C. A computer code named LeBENC has been developed at BARC to simulate the natural circulation characteristics in closed loops. The salient features of the code include ability to handle non-uniform diameter components, axial thermal conduction in fluid and heat losses from the piping to the environment. The code has been modified to take into account of two natural circulation loops in series so that the natural cooling by argon gas in the secondary side of the loop can be simulated. This paper deals with the description of the loop and its operation. The various experimental studies carried out in the loop are discussed. Further the theoretical analysis using LeBENC code and the comparison of the predicted results with the experimental data are discussed in the paper.

Experimental Investigations of the Thermohydraulic Characteristics of Transverse Lead-Coolant Flow over the Tubes of a Steam Generator

The results of experimental studies of lead-to-wall heat transfer during transverse flow over a straight-line tube bundle in a steam generator in the circulation loop during monitored and controlled mass-exchange and mass-transfer of oxygen impurities are presented. The investigations were performed in a nonisothermal loop with circulation of lead coolant at temperature 450–550°C, average coolant velocity in a narrow-gap section 0.2–1.5 m/sec, Peclet number 900–3000, and thermodynamic activity of oxygen in lead 1–10–4. The processes occurring in a nonisothermal liquid-metal loop with an experimental cooling section model the dependence of the characteristics of heat exchange on the mass transfer of impurities in the loop.

Overview of Indian activities on fusion reactor materials

This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils.Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V–4Cr–4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu–1wt%Cr–0.1wt%Zr – an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall – was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER).India has proposed to install and test a lead–lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb–Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box are briefly described.India has also initiated the development of fusion grade superconductors. Success achieved in the fabrication of Nb3Sn based multi-filamentary wires using the internal tin process and cable-in-conduit-conductors is also briefly presented.

A neutron production target for ESS based upon the Canned-rods concept

The neutron production targets operating within the present day spallation neutron sources in the MW power range are either based on water-cooled solid state devices such as that implemented at the SINQ source at PSI or liquid metal loops such as those installed at SNS and MLSF. Here we describe a water-cooled rotating solid target as an option for the 5MW ESS project as an alternative to the current design based upon a helium-cooled solid rotating target. Implementation of the proposed option would provide comparable neutronic performance to that of the gas-cooled concept and furthermore, it would involve a relatively straightforward adaptation of the current ESS baseline geometry.

IFMIF/EVEDA lithium test loop: design and fabrication technology of target assembly as a key component

The engineering validation and engineering design activity (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER broader approach activities. In the concept of the IFMIF, two 40 MeV deuteron beams are injected into a liquid Li stream (Li target) flowing at a velocity of 15 m s−1. The EVEDA Li test loop (ELTL) is aimed at validating the hydraulic stability of the Li target at a velocity up to 20 m s−1 under a vacuum condition of 10−3 Pa as the most important issue. Construction of the ELTL, which is the largest liquid metal loop possessing 5.0 m3 Li for the fusion research ever, was completed in the O-arai Research & Development Center in the Japan Atomic Energy Agency on 22 November 2010. This paper presents the design and fabrication technology of a target assembly called integrated target assembly, in which the Li target is produced by a contraction nozzle along a concave channel. There are two concepts regarding the target assembly: the integrated target assembly and the bayonet target assembly. Both target assemblies are outlined in this paper, and then the newly proposed design of the integrated target assembly for the ELTL and its fabrication technology are given. The integrated target assembly was processed by a five-axis milling machine and the processing accuracy was measured by 3D measurement tools. Finally, methods applied for the validation of the stability of the Li target are introduced in this paper.

Leakage Control of Tritium Through Heat Cycles of Conceptual-Design, Laser-Fusion Reactor KOYO-F

To reduce the tritium permeation from the primary liquid metal loop to the secondary water loop, a heat exchanger concept that incorporates small diameter tubes containing an oxidizer was proposed. An inert gas containing a small amount of oxidizer flows in the small tubes oxiding tritium that comes from the primary liquid metal coolant. The tritiated water is sent to a tritium recovery system minimizing leakage to the secondary water loop. Our evaluation results indicated that the tritium leakage through the heat exchanger was reduced to 1/105 with an acceptable increase in the size of the heat exchanger.

Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

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.

Integrated infrastructure initiatives for material testing reactor innovations

The key goal of the European FP6 project MTR+I3 was to build a durable cooperation between Material Testing Reactor (MTR) operators and relevant laboratories that can maintain European leadership with updated capabilities and competences regarding reactor performances and irradiation technology. The MTR+I3 consortium was composed of 18 partners with a high level of expertise in irradiation-related services for all types of nuclear plants. This project covered activities that foster integration of the MTR community involved in designing, fabricating and operating irradiation devices through information exchange, know-how cross-fertilization, exchanges of interdisciplinary personnel, structuring of key-technology suppliers and professional training. The network produced best practice guidelines for selected irradiation activities. This project allowed to launch or to improve technical studies in various domains dealing with irradiation test device technology, experimental loop designs and instrumentation. Major results are illustrated in this paper. These concern in particular: on-line fuel power determination, neutron screen optimization, simulation of transmutation process, power transient systems, water chemistry and stress corrosion cracking, fission gas measurement, irradiation behaviour of electronic modules, mechanical loading under irradiation, high temperature gas loop technology, heavy liquid metal loop development and safety test instrumentation. One of the major benefits of this project is that, starting from a situation of fragmented resources in a strongly competitive sector, it has created a framework opening the way to further collaboration between the involved partners regarding the development and utilization of irradiation devices. In addition, the courses developed in the different institutes within the framework of professional training can continue after the end of the project.

CFD ANALYSIS OF HEAVY LIQUID METAL FLOW IN THE CORE OF THE HELIOS LOOP

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER² ) 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.

Silicon production from phosphate industry waste (Na2SiF6)

Production of salable goods from industrial waste can become an additional innovative component of liquid-metal technology of atomic engineering. In this paper, we used sodium silicofluoride (Na2SiF6), i.e., phosphate Ca(HPO4)2 production waste. Fundamentals of a worldwide unique ecologically clean low-temperature technology of Si reduction from Na2SiF6 and Si deposition on substrates in a liquid-metal loop in a single process was developed, as well as a gas-phase technology of Si reduction by sodium from gaseous SiF4 released during thermal dissociation of Na2SiF6.

Calculation of dissolution/deposition rates in flowing eutectic Pb–17Li with the MATLIM code

It is generally accepted that corrosion of steel components in flowing liquid Pb–17Li is ruled by dissolution/deposition processes. A simple method has been devised to calculate the mass transfer in liquid metal loops. It is based on the use of mass transfer coefficients which determine the mass flux from the wall into the fluid and vice versa. These coefficients depend on a characteristic dimensionless thermo-hydraulic number, namely the Sherwood number, which itself depends under forced convection flow conditions on the Reynolds number and the Schmidt number. This is supplemented by the application of differential equations for dissolved steel components. The dissolution and deposition rates determine then the evolution of the flow channel geometry along the loop and the flow velocity. The newly developed code MATLIM can routinely deal with multi-modular liquid metal loops. Thus, calculations have been done for the PICOLO loop based on iron solubility and diffusivity correlations of Feuerstein et al.

Unusual wavy weld pool boundary from dimensional analysis

The geometry of a weld pool affects its solidification behavior. Here we show that the welding parameters and material properties responsible for the formation of unusual, wavy fusion boundaries can be identified from heat and fluid flow calculations and confirmed by independent experiments. These parameters and properties can be expressed by dimensionless numbers using the Buckingham π-theorem. The wavy boundary originates from the interaction of counter-rotating liquid metal loops at high Marangoni numbers. The Prandtl number affects the type of inflection.