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Liquid metals and magnetic fields are used in many technical applications such as metallurgy, crystal growth and nuclear fusion reactors. When an electrically conducting fluid moves in a magnetic environment, electric currents and electromagnetic forces are generated that affect velocity and pressure losses in the flow. These magnetohydrodynamic (MHD) interactions have to be investigated to optimize the engineering processes. The characteristics of MHD flows depend on the geometrical configuration, the strength of the applied magnetic field, the electrical properties of fluid and structural materials and the thermal conditions. In the so-called blankets for fusion reactors, where liquid metals are used to breed the plasma fuel component tritium and to extract the generated heat, magneto-convective flows play a crucial role in determining heat and mass transfer. Therefore, the availability of numerical codes to simulate this type of flow is mandatory and their validation is a necessary step to guarantee the reliability of the results. For that reason, a benchmark problem has been defined to simulate liquid metal flows in a horizontal rectangular duct heated from below and exposed to a non-uniform magnetic field. Results obtained by five research groups using different codes are compared.

Tritium self-sufficiency is a critical prerequisite for future fusion reactors. The tritium breeding blanket, as the component responsible for in-vessel tritium generation, requires coordinated neutronic and engineering optimization in order to maximize its achievable tritium breeding ratio (TBR). In this work, a high-fidelity 22.5° toroidal sector neutronics model of the China Fusion Engineering Test Reactor (CFETR) equipped with a Helium-Cooled Ceramic Breeder (HCCB) blanket was established. On this basis, we developed a Multi-physics Coupling Intelligent Neutronic Optimization code (MCINO), a two-stage neutronics optimization framework that combines global exploration by simulated annealing with subsequent local refinement. The objective was to maximize the global TBR by optimizing the radial distribution of breeder (Li4SiO4) and neutron multiplier (Be) zones. The optimized design increased the global TBR to approximately 1.193, corresponding to an 8.36% improvement over the initial configuration. The improvement is associated with a more effective radial allocation of breeding and multiplying materials, which enhances neutron moderation, multiplication, and use for tritium production. The optimization workflow was designed to reduce the number of expensive high-fidelity transport recalculations, thereby improving computational efficiency relative to direct brute-force search. Finally, the engineering feasibility of the optimized design was checked through three-dimensional thermal-hydraulic verification, which confirmed that the representative modules remained within their prescribed operating limits. The present work provides an efficient and physically transparent framework for integrated blanket neutronics design and optimization.

Efficient tritium extraction from liquid lead-lithium (PbLi) breeding blankets is critical for fusion reactor technology development. In this study, numerical simulation methods based on a fundamental gas-liquid contactor (GLC) model were employed to optimize boundary conditions and external structures, ensuring stability of liquid PbLi level during operation. Computational results show that the improved GLC achieves a tritium extraction efficiency of 19.18%. Further enhancement was realized by increasing GLC mesh density and installing a gas-guiding device at the internal helium inlet, significantly improving gas-liquid contact within the GLC. Final results indicate that GLC’s tritium extraction efficiency increased to 25.73%, suggesting that the gas guiding device introduction effectively enhanced gas-liquid contact and tritium extraction efficiency.

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α ′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

Fusion reactor materials for the first wall and blanket must have high strength, be radiation tolerant and be reduced activation (low post-use radioactivity), which has resulted in reduced activation ferritic/martensitic (RAFM) steels. The current steels suffer irradiation-induced hardening and embrittlement and are not adequate for planned commercial fusion reactors. Producing high strength, ductility and toughness is difficult, because inhibiting deformation to produce strength also reduces the amount of work hardening available, and thereby ductility. Here we solve this dichotomy to introduce a high strength and high ductility RAFM steel, produced by a modified thermomechanical process route. A unique multiscale microstructure is developed, comprising nanoscale and microscale ferrite, tempered martensite containing fine subgrains and a high density of nanoscale precipitates. High strength is attributed to the fine grain and subgrain and a higher proportion of metal carbides, while the high ductility results from a high mobile dislocation density in the ferrite, subgrain formation in the tempered martensite, and the bimodal microstructure, which improves ductility without impairing strength. Reduced-activation ferritic-martensitic (RAFM) steels are promising fusion reactor materials having high radiation tolerance and low post-use radioactivity, but achieving high strength, ductility and toughness is difficult. The authors demonstrate a modified thermomechanical process to produce RAFM steel with a multiscale microstructure which endows both high strength and high ductility.

The IFMIF-DONES Irradiation Modules consist of a set of systems designed to be tested in the IFMIF-DONES facility and which are thought to produce, in the future, an extensive, qualified and unique set of fusion-like irradiated experimental data associated to the materials and the structures to be used in specific breeding blankets (BBs) that are currently under consideration for the EU DEMO fusion reactor. The irradiation conditions (neutron fluence and spectra) at the position of the IFMIF-DONES Irradiation Modules inside the test cell, will be comparable to that expected as reference in the most exposed zones of the EU DEMO reactor. Nevertheless, medium and low flux regions will be also available. In this work, an introduction to the description of the main IFMIF-DONES Irradiation Modules is given. Particularly, the following 5 modules are described: the (IFMIF-DONES) High Flux Test Module, HFTM, tailored to powerful irradiate structural materials such as small specimens of Eurofer, the (IFMIF-DONES) Start-up Monitoring Module, STUMM, designed to monitor the radiation fields produced just behind the neutron source, under transient and steady state conditions, the (IFMIF-DONES) Blanket Functional Material Module, BLUME, which is a representative section of the Helium Cooled Pebble Bed (HCPB) blanket breeder zone, the (IFMIF-DONES) Liquid Breeder Validation Module, LBVM, for the irradiation of functional materials such as PbLi and anti-corrosion/anti-permeation barriers, and the (IFMIF-DONES) Tritium Release Test Module, TRTM, which is defined for the irradiation of lithium ceramic pebble beds associated to the HCPB BB.

The Supercritical Carbon Dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket is under development for Chinese Fusion Engineering and Test Reactor. The thermal hydraulic assessment plays an important role for the comprehensive performance evaluation on the fusion blanket among the multi-physics fields. As the fusion reactor will enter into the engineering construction stage, it is important to study the thermal hydraulics performance on basis of the full model. Because it can accurately check the heat removal capability and thermoelectricity conversion efficiency, as well as provide essential input for the other physical fields. In this demand-driven, the analyses and optimization on the cooling system are put into priority on basis of the full banana model, including the manifold design and inlet/outlet pipes locations. Finally, the coolant pressure drop is highly reduced and the mass flow distribution becomes much more uniformly. For the S-CO2, 82.3% of the total mass flow rate is distributed into the key component first wall, and this is beneficial to face the high radiation heat flux. Besides, under different level of heat flux, the required total mass flow rate and pressure drop are obtained on premise that the coolant has enough ability to safely remove the heat away. For the Lead–Lithium (PbLi), the distribution of mass flow rate is designed as ‘ladder’ shape to adapt the unevenly spatial distributed nuclear heat along the radial direction, and the ratio is 8:2:1. Furthermore, the first law of thermodynamics is adopted for the trade-off analysis, which converts the total pressure drop of the two coolants into the pumping power, and it occupies only 1.3% of the total thermal power. This provides accurate and valuable data for the primary and secondary loop design, as well as the economic assessment on the fusion reactor. Finally, the Two Dimensional thermal hydraulic model containing the detailed layouts of different materials is used to study the coupling heat transfer effects between PbLi and S-CO2, as well as the MagnetoHydroDynamics (MHD) effects. The boundary conditions are derived from the results of full banana model, and the results show that the temperature of all materials is not exceeding the upper limits.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM’s corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic–martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

The tritium breeding capacity of fusion blankets is a critical factor in achieving tritium self-sufficiency in fusion reactors. Current designs for tritium production in blankets are based on neutronics simulations, whose accuracy requires experimental validation. This study focuses on the experimental validation of the neutronic design reliability of the supercritical CO2-cooled lithium-lead (COOL) blanket developed for the China fusion engineering test reactor (CFETR). For this purpose, an experimental mock-up was designed and fabricated to replicate the key neutronics characteristics of COOL blanket. The mock-up was irradiated using D–T neutron generator and validated using multiple techniques: tritium production rate (TPR) online monitoring with a miniature back-to-back lithium glass scintillator detector, TPR integral measurement using Li2CO3 pellets analyzed by liquid scintillation counting (LSC), and neutron flux measurement with activation foils (Au, Zr). To enhance accuracy, corrections were applied for lithium-lead (PbLi) segregation based on element analysis and for neutron source intensity based on in-situ depth profiling of tritium in the tritide target. The results show excellent agreement between experiments and simulations, with calculation-to-experimental (C/E) value ranging from 0.96 to 1.11 for TPR measured by lithium-glass scintillator detectors, 0.90–1.09 for TPR by Li2CO3 pellets, and 0.84–1.13 for reaction rates measured by activation foils.