Liquid Breeder Research Articles

Investigation on numerical simulation models for heat transfer characteristics of lead-bismuth eutectic

The liquid lead-bismuth eutectic (LBE) has distinct thermal and physical properties that challenge the effectiveness of traditional RANS turbulence models and turbulent Prandtl number (Prt) models for accurately simulating its flow and heat transfer characteristics. This study aims to investigate numerical simulation models of LBE based on existing experimental data. Simulations are conducted using different combinations of three turbulence models and four Prt models, and comparisons are made with experimental data. The local and overall heat transfer characteristics of LBE are analyzed, and the applicability of each model combination is evaluated. Results indicate that for simulating local heat transfer characteristics, the SST k-ω turbulence model combined with the Prt model proposed by Cheng et al. yields the highest accuracy. Additionally, the empirical correlation for heat transfer proposed by Kutateladze provides the best prediction of the local Nusselt number. For overall heat transfer characteristics, the combination of the SST k-ω turbulence model and the Prt model introduced by Reynolds et al. demonstrates the highest accuracy and applicability. This investigation might offer a pivotal benchmark for the discernment of appropriate computational fluid dynamics (CFD) models for liquid metal breeder reactor (LMBR) applications utilizing lead-bismuth eutectic (LBE) as coolant.

Growth of α-Al2O3 layer involving abnormal oxides in FeCrAl alloy tube fabricated by WEDM process and electrical insulating performance in fusion reactor blanket

The tubular specimens of FeCrAl alloy APMT (Fe-22Cr-5Al-3Mo) were fabricated by a wire electrical discharge machining (WEDM) process. The oxidation characteristics of the machined surface were investigated by means of the oxidation tests performed in air atmosphere at the temperatures of 1273 K and 1373 K up to 100 h. The abnormal oxides such as CuAl2O4 and ZnAl2O4 were grew on the recast layer which was locally remained on the machined surface even after the mechanical polishing procedure. The α-Al2O3 layer involving the abnormal oxides was formed on the machined surface in the oxidation treatment at 1273 K for 10 h. The apparent electrical conductivity of the layer involving the abnormal oxides was larger than that without the abnormal oxides. Thick Al-rich oxide layer was formed under the abnormal oxides when the oxidation treatment was performed at 1373 K. The apparent electrical conductivity of the oxide layer became lower due to the presence of the thick Al-rich oxide. This oxide layer formed at 1373 K could function as an electrical insulating layer to reduce the magnetohydrodynamic pressure drop in liquid breeder blanket systems of magnetic confinement fusion reactors.

Impact of Reverse Buoyancy on Heat and Mass Transfer in MHD Flow of PbLi Within a Poloidal Duct in a Typical DCLL Liquid Breeder Blanket

The typical dual-coolant lead-lithium (PbLi) design of a liquid breeder blanket in a magnetic confinement fusion reactor involves the utilization of PbLi as the working fluid to effectively remove neutron heat. However, the nonuniform heating of neutrons with a significant radial gradient induces a buoyancy effect, resulting in the formation of vortexes ices within the downward flow duct. These vortexes have an adverse impact on the heat and mass transfer characteristics of the magnetohydrodynamic (MHD) flow of PbLi. The simulations in this work employed a MHD buoyant mixed-convection solver to resolve the characteristics of PbLi flow and a one-way coupled Lagrangian method to analyze the qualitative characteristics of tritium transport in PbLi flow. The results indicate that buoyant reverse flow can create vortexes that contain hot spots in the PbLi fluid, which can significantly impede heat transport. Additionally, the vortex causes tritium recirculation in the flow field and retention, resulting in adverse effects on tritium transport.

Liquid lithium corrosion of SiC/SiC composites

Silicon carbide fibre reinforced silicon carbide matrix composite (SiC/SiC) is a key structural material for fusion reactors to allow high temperature operation of liquid lithium breeder blankets. This research investigated the corrosion behaviour of two SiC/SiC composites (using Tyranno SA3 and SA4 fibres) immersed in static liquid lithium at 600 °C for 100 h. Utilizing immersion tests and advanced microstructural analysis, the study reveals considerable corrosion in SiC/SiC composites, particularly in local areas enriched with residual carbon or oxygen. The interface carbon layer, while enhancing mechanical properties, induces preferential sites for corrosion cracking, diminishing the material's corrosion resistance. This study offers critical insights into the corrosive interaction between SiC/SiC composites and liquid lithium, highlighting the need for manufacturing process optimization and exploration of alternative interface layers, protective coatings, or avoiding contact between SiC/SiC and molten lithium in breeder blanket system.

Design and Development of Hydrogen Isotopes Extraction System at IPR

Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles Li2TiO3 (Lithium Titanate) as solid breeders and second is the lithium based alloy Pb-Li (Lead Lithium) as a liquid breeder. Tritium bred in solid breeders is extracted with purge helium gas, whereas tritium bred in liquid Pb-Li needs suitable gas-liquid contactor. It should be noted that, tritium recovery (i.e.,-extraction) from Pb-Li is very challenging due to low solubility of tritium in liquid Pb-Li and no such separation technology readily available commercially. Therefore, the necessity for developing a reliable and efficient tritium recovery technology (i.e.,- extractor) for liquid Pb-Li is of key interest.In present work, design and development of Hydrogen Isotopes Extraction System (HIES) for hydrogen recovery from liquid Pb-Li is discussed. As hydrogen and tritium are isotopes, hydrogen is used instead of tritium in present system due to several safety concerns with tritium. Recovery of hydrogen from liquid Pb-Li is carried out using structured packing as a Gas-Liquid Contactor (GLC). The design details of gas-liquid contactor (i.e.-extractor column) and various process tanks of HIES are discussed. In present design, configuration of vertically mounted extractor column makes overall HIES relatively tall (∼ 3 m height) and use of liquid Pb-Li as process fluid makes the system heavy weight (∼1500 kg) too. In addition to this, system needs to be operated above Pb-Li melting temperature (> 235 C) to keep Pb-Li in liquid form. Therefore, thermal analysis of piping as well as seismic analysis of whole system is carried out and discussed along with various operational stages of HIES.

Development of the high temperature PbLi experimental facility for CFETR

Chinese Fusion Engineering and Test Reactor (CFETR) aims to demonstrate the fusion energy production up to 200MW, and finally reach commercial electricity power level 1GW. As one key component, the blanket is in charge of tritium breeding, neutron shielding and energy conversion. The liquid breeder blanket has many important advantages of simpler structure, tritium extraction and fuel replenishment online, as well as the high thermoelectric conversion efficiency. Moreover, the supercritical carbon dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket has been proposed for CFETR. However, the key issue of MagnetoHydroDynamics (MHD) needs further investigation, especially for the experiments. Therefore, the high temperature PbLi experimental loop is under development, which mainly includes three systems of PbLi, oil and water. The main components include the mechanical pump, main heater, cold trap as well as the superconducting magnet. At the test section, the mass flow rate, temperature and magnetic intensity will be up to 32 kg/s, 700 °C and 3T, respectively. In addition, the ultrasonic doppler velocimeter and potential probe instruments will be adopted to obtain the flow field. The main goal is to establish flow and heat transfer models, which are the critical input for the liquid breeder blanket design and optimization.

Evaluation of the strength of FeCrAl alloy/surface Oxide film interface by micro double-notch shear test

The shear strength of the film formed by pre-oxidation treatment of FeCrAl(-ODS) alloys, likely to be used as pipework material for liquid LiPb breeder blankets, was evaluated by micro double-notch shear testing. Fracture generally occurred at the interface between the oxide film and the alloy substrate, but the pop- in the stress-strain curves was difficult to interpret, so all the test results in this study were compared to the full rupture strength. The rupture strength appears to be a function of the nature of the interface between the film and the alloy rather than the thickness of the film; for APMT and FeCrAl-ODS, the rupture strength was greater for FeCrAl-ODS.

Overview of tritium management in WCLL test blanket system of ITER

A key aspect for the achievement of the fuel self-sufficiency of future fusion power plants is the management of tritium in the different stages of the reactor fuel cycle. In particular, one of the greatest challenges for the success of liquid metal breeder concepts such as the Water-Cooled Lithium-Lead (WCLL) is the extraction of tritium from the liquid LiPb eutectic alloy (15.7 at.%Li), which fulfils the main functions of tritium breeder, tritium carrier and neutron multiplier, and the processing in a form suitable for an accurate tritium accountancy. Within this frame, significant advancements have been carried out in the last years as regards the WCLL Test Blanket System (TBS) of ITER. Tritium is mainly produced inside the LiPb and extracted through the Tritium Extraction Unit (TEU). The reference technology for the TEU of the WCLL-TBS is the Gas-Liquid Contactor (GLC) in the packed-column configuration. Tritium is then concentrated in ZrCo getter beds before being sent to the accountancy system where different technologies for both static and dynamic accountancy are adopted in order to provide precise and reliable tritium measurements, which are an essential need in view of an application to the European DEMO reactor. In this paper, the present status and the design solutions foreseen for tritium management in WCLL-TBS are presented and discussed. The implementation and optimization of the conceptual design of the lithium-lead loop (LLP), the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS) have been performed in order to manage tritium concentration in LiPb and in the ancillary systems.

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

Magnetohydraulic flow in a rectangular channel filled with stream-wise obstacles

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two leading candidates for implementation as driver blanket in the European DEMO reactor, which is expected to start operation in the late 2050s and serve as a BB test facility. One of the main risks associated with the current WCLL concept is the large number of welds necessary to realize its horizontal (toroidal-radial) breeding zone cooling system and stiffening structure, which negatively affects its reliability. Hence, an alternative variant concept has been proposed which relies on vertical (radial-poloidal) “double-bundle” cooling tubes and stiffening plates: the WCLL double-bundle (WCLL-db). In this paper, we report the results of numerical analyses performed to characterize the Magnetohydrodynamics (MHD) regime in the WCLL-db where the liquid metal breeder flows vertically and it is obstructed by a large number of electrically conductive obstacles aligned with the stream-wise direction.Direct numerical simulations are performed with the aid of computational MHD tools to characterize the flow features and pressure losses in the WCLL-db breeding zone. The presence of many stream-wise obstacles affects the velocity distribution since the cooling tubes provide additional paths for the electric current closure. The most striking feature is the breaking down of the classic MHD slug core flow into many separate smaller regions, separated by internal layers. These occur tangential to the pipe walls, featuring velocity overshoots, and tend to propagate along the imposed magnetic field. A moderate (3.83–9.73%) pressure penalty is found compared with the analytical prediction for a channel devoid of obstacles.

Design of a scaled mockup of the WCLL TBM for MHD experiments in liquid metal manifolds and breeder units

Design of a scaled mockup of the WCLL TBM for MHD experiments in liquid metal manifolds and breeder units

Reduction of MHD pressure drop by electrical insulating oxide layers in liquid breeder blanket of fusion reactors

Reduction of MHD pressure drop by electrical insulating oxide layers in liquid breeder blanket of fusion reactors

Initial Neutronics Investigation of a Chlorine Salt–Based Breeder Blanket

Tritium breeding blankets within D-T-fueled fusion reactors contain lithium compounds and typically require neutron multiplier materials to achieve a tritium breeding ratio (TBR) consistent with self-sustaining operation. Liquid breeder blankets have some advantages over solid blankets, and previous blanket studies have investigated liquid metal as well as liquid salt–based blankets. Liquid salts have reduced magnetohydrodynamic effects as compared to liquid metals, but typically have a lower TBR. Recently, advanced fission reactor concepts have considered chloride-based salts in their design, and there is a significant amount of research work occurring to study these salts. Chloride salts have previously been considered for fusion reactors, but studies have typically found lower breeding ratios than for fluoride salts, such as 2(LiF)-BeF2 (flibe) so they have not been further developed. In this work, we use a one-dimensional cylindrical radiation transport model of a conceptual tokamak fusion reactor to investigate the neutronics feasibility of using a chloride salt–based blanket that uses chlorine enriched in 37Cl, which has both a low capture cross section and a substantial (n,2n) cross section. It is found that chloride salts (LiCl mixed with BeCl2 and/or PbCl2) can potentially achieve a ~3% to 5% higher TBR than fluoride molten salts, notably flibe, in the absence of a solid multiplier. Including a solid multiplier, however, does narrow this advantage, with TBRs estimated within ~1% of flibe with a 2-cm Be multiplier. Chloride salts can also reach lower melting points than flibe, potentially improving the scope for the use of reduced activation ferritic-martensitic steel as a structural material. There is substantial uncertainty in the calculations driven by limited thermochemical data for the Cl salts, plus cross-section uncertainties. The production of 36Cl through 35Cl(n,g) and 37Cl(n,2n) has the potential to challenge the waste disposal rating of the blanket. Calculations indicate that, while this is not an immediate showstopper, this case depends upon the exact waste disposal rating criteria used for 36Cl. Further work could reduce these uncertainties with improved thermochemical data, higher-fidelity modeling for downselected salts, and more refined waste disposal calculations and regulatory guidance. Finally, it must be recognized that, as for all molten salts, corrosion and chemistry can present appreciable technical challenges that require further assessment in developing a practical blanket concept, and also that the enrichment of chlorine presents an additional technical and supply chain challenge.

An Innovative Approach to Prepare Liquid-Solid Dual-Phase Flowable Tritium Breeder with Low MHD Effect

In present paper, a novel flowable tritium breeder is prepared by mixing the Li2TiO3 micro-powders and liquid GaInSn alloy, where GaInSn alloy is used to simulate the fluid behaviors of lithium-based liquid tritium breeder, forming a type of composite characterized by liquid-solid dual phase. In detail, the effects of the volume fraction of ceramic micro-powders on viscosity and conductivity of the composite in magnetic field are the focus. The XRD results prove that the obtained Li2TiO3 micro-powders contained Li2TiO3 phase without impurities. The results shows that once the magnetic field intensity exceeds the critical value, the viscosity of liquid GaInSn metal becomes significantly greater than that of liquid-solid dual-phase composites. Furthermore, the addition of Li2TiO3 micro-powders could effectively reduce the magneto hydro dynamic (MHD) fluid effect, and the dual-phase composites exhibit comparatively lower flow resistance under the strong magnetic field. Moreover, the conductivity of the tritium breeder composites decreases rapidly with the addition of Li2TiO3 micro-powders. The MHD pressure-drop-increasing rate decreases with the increase of viscosity, which indicates that the addition of Li2TiO3 micro-powders effectively reduces the MHD effect. The conductivity of the composites increased slightly and then remained stable after static placing for several tens of minutes. The present investigation provides a novel insight into the fabrication strategy of tritium breeder materials with low MHD effect.

A duct-dimension-regulation approach for eliminating MHD PbLi flow reversal in buoyancy-opposed poloidal duct relevant for typical liquid breeder blanket designs

Lithium-lead (PbLi) flow plays the major roles: removing nuclear heating, breeding tritium and multiplying neutrons, in the appeared typical liquid breeder blanket designs for magnetically confined fusion reactors. Under the buoyancy effect caused by the volumetric nuclear heating with large gradient in the radial direction, flow reversals can be generated in the magnetohydrodynamic (MHD) PbLi flow in the poloidal downward-flow duct(s) of such blanket designs, which may import negative consequences to the functionality of the breeder blanket. Based on a theory regarding mixed convection in fully developed MHD duct flows developed by Smolentsev et al. (Smolentsev, Moreau & Abdou, Characterization of key magnetohydrodynamic phenomena in PbLi flows for the US DCLL blanket, Fusion Engineering and Design 83, 771-783, 2008), this study attempts to further quantitatively analyze the above-mentioned flow-reversal issues, and in the meantime to develop an approach to eliminate the reverse flow regime by regulation of the duct dimensions. Numerical simulations are performed for buoyancy-opposed MHD duct flows of PbLi in order to demonstrate the effectiveness of the proposed duct-dimension-regulation approach on eliminating the occurrence of reverse flows. It is found that, in the reverse flow regime, the reversal structures travel to the downstream from their onset locations at a speed essentially equal to the mean flow velocity in the bulk; the reverse flow regime can be effectively eliminated by regulating the duct dimensions. Comprehensive discussions about the reverse flow related issues in the DCLL type blanket are then stressed.

Status of maturation of critical technologies and systems design: Breeding blanket

The scope of the EUFOfusion Work Package Breeding Blanket is to develop a blanket concept for the EU DEMO reactor; this includes the blanket segments inside the Vacuum Vessel and the related Tritium Extraction/Removal Systems. In the Pre-Concept Design (PCD) Phase, two concepts have been selected as candidates; a solid and a liquid breeder blanket cooled with helium and water, respectively. The design of these two blanket systems has been adapted to the DEMO plant design developed in the PCD Phase and performances assessed. A large R&D programme has been implemented with the scope to evaluate different technologies for these blankets; including the development of breeders, tritium extraction and cooling technologies, and the manufacturing of the blanket system. A major milestone in the subsequent Concept Design Phase is the final selection of the blanket concept for DEMO.

Design of the CIEMAT Corrosion Loop for Liquid Metal Experiments

The main components of a liquid breeder blanket in a fusion power reactor are lead lithium alloy (PbLi) and the steel structure in which the liquid is enclosed (EUROFER). Several compatibility tests have shown that structural materials always suffer from corrosion attacks. The governing mechanism can be attributed to the dissolution of the steel by the liquid breeder and is strongly related to the PbLi chemistry, velocity profile, and temperature. A new facility, CiCLo-C (CIEMAT Corrosion Loop, Internally Coated), is dedicated to the study of corrosion in materials under the severe breeding blanket condition. An effort was made to design an experimental facility with a specific test section able to work at quite ambitious operation parameters: up to 550 °C and a 1 m/s flow of PbLi. Furthermore, an innovative tantalum coating was introduced in the whole loop to avoid impurities coming from the pipeline, which can disturb the measurements, and to better preserve the installation.

High temperature zirconium alloys for fusion energy

This review considers current Zr alloys and opportunities for advanced zirconium alloys to meet the demands of a structural material in fusion reactors. Zr based materials in the breeder blanket offer the potential to increase the tritium breeding ratio above that of Fe, Si and V based materials. Current commercial Zr alloys might be considered as a material in water-cooled breeder blanket designs, due to the similar operating temperature to fission power plants. For breeder blankets designed to operate at higher temperatures, current commercial Zr alloys will not meet the high temperature strength and thermal creep requirements. Hence, Zr alloys with an operational temperature capability beyond that of current commercial fission alloys have been reviewed, specifically: binary Zr alloy systems Zr-Al, Zr-Be, Zr-Cr, Zr-Nb Zr-Ti, Zr-Si, Zr-Sn, Zr-V and Zr-W; as well as higher order Zr alloys Zr-Mo-Ti, Zr-Nb-Ti, Zr-Ti-Al-V and Zr-Mo-Sn. It is concluded that, with further work, higher order Zr alloys could achieve the required high temperature strength, alongside ductility, while maintaining a low thermal neutron cross-section. However, there is limited data and uncertainty regarding the structural performance and microstructural stability of the majority of advanced Zr alloys for temperatures 500–700 °C, at which they would be expected to operate for helium- and liquid metal-cooled breeder blanket designs.

Study of Weldability for Aluminide Coated Steels through A-TIG Welding Process

Abstract To achieve higher efficiency in high-temperature environments, 9Cr-1Mo steels (also known as P91 steels) are promising candidates in power plants, petrochemical industries, heat exchangers, and nuclear applications. To enhance the performance against harsh environments, such as high-temperature oxidation and corrosion during operations, adding aluminum oxide (Al2O3)/iron aluminide (FeAl) coating through the aluminizing process has been found promising. Such coatings have been considered reliable for resistance to liquid-metal corrosion against flowing lead-lithium liquid breeders and against tritium permeation in blanket modules of fusion reactors. However, the welding process for aluminized coated steels is a challenge and is scarcely reported. The aluminum-rich coating at the top may alter the metallurgy of weld metal and subsequently the mechanical properties after welding. To investigate the same, preliminary bead-on-plate trials with a novel activated-tungsten inert gas welding process has been attempted for the first time. Bead-on-plate welding samples were prepared (with the same welding parameters) with five different fluxes, such as iron (III) oxide (Fe2O3), nickel (II) oxide (NiO), copper (II) oxide (CuO), cobalt tetroxide (Co3O4), and Al2O3, for aluminized coated 9Cr-1Mo steel. The resultant weld metal microstructure was evaluated using optical and electron microscopy, elemental analysis using energy-dispersive X-ray spectroscopy, phase analysis using X-ray diffraction, and Vickers hardness testing. It was observed that the presence of oxide-type fluxes results in arc constriction, which thereby affects the depth of penetration. The weld metal prepared with Fe2O3 flux demonstrates a superior depth-to-width ratio compared to other fluxes. An analysis for delta ferrite was also carried out and was observed to be present in Fe2O3 flux welds only. A detailed analysis of the weld metal under different activated fluxes is presented in the paper. The outcome of this work presents the feasibility of the welds of aluminized 9Cr steels as relevant for fusion reactor applications.

Structural Assessment of the European DEMO Water-Cooled Lithium Lead Breeding Blanket Central Outboard Segment Under Remote Maintenance Loading Conditions

Within the framework of the European research activities devoted to design the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead (WCLL) BB concept is one of the candidates currently assessed in EU. One of the main issues connected to its design is represented by the remote maintenance (RM) operations. In particular, previous thermo-mechanical analysis highlighted a Ductile to Brittle Transition Temperature (DBTT) shift above the room temperature within the WCLL central outboard (COB) segment during normal operation. This may pose problems during the RM operations foreseen in the BB replacement phase. Indeed, within those WCLL COB segment’s regions where the DBTT shifted above the room temperature, the stress level arising during RM operations may be able to promote crack growth and migration resulting in a fragile fracture event. Then, in this work, the structural assessment of the DEMO WCLL COB segment under RM scenarios is performed, in order to provide the fracture mechanics analysis with the necessary input in terms of stress spatial distribution and to understand if the deformation occurring during RM can be accommodated by the available space reservation. Several potential RM scenarios have been considered, taking into account also the potential occurrence of a seismic event or the necessity to remove the WCLL COB segment with the liquid breeder not drained. The study has been performed adopting a numerical approach based on the Finite Element Method and adopting the quoted Abaqus code. The obtained results are quite encouraging and will allow, in the follow up, the refinement of the RM plan foreseen for the WCLL BB.