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Alpha particles with energies on the order of megaelectronvolts will be the main source of plasma heating in future magnetic confinement fusion reactors. Instead of heating fuel ions, most of the energy of alpha particles is transferred to electrons in the plasma. Furthermore, alpha particles can also excite Alfvénic instabilities, which were previously considered to be detrimental to the performance of the fusion device. Here we report improved thermal ion confinement in the presence of megaelectronvolts ions and strong fast ion-driven Alfvénic instabilities in recent experiments on the Joint European Torus. Detailed transport analysis of these experiments reveals turbulence suppression through a complex multi-scale mechanism that generates large-scale zonal flows. This holds promise for more economical operation of fusion reactors with dominant alpha particle heating and ultimately cheaper fusion electricity. Experiments at the Joint European Torus tokamak show improved thermal ion confinement in the presence of highly energetic ions and Alfvénic instabilities in the plasma.

High-entropy alloy (HEA) superconductors—a new class of functional materials—can be utilized stably under extreme conditions, such as in space environments, owing to their high mechanical hardness and excellent irradiation tolerance. However, the feasibility of practical applications of HEA superconductors has not yet been demonstrated because the critical current density ( J c ) for HEA superconductors has not yet been adequately characterized. Here, we report the fabrication of high-quality superconducting (SC) thin films of Ta–Nb–Hf–Zr–Ti HEAs via a pulsed laser deposition. The thin films exhibit a large J c of >1 MA cm −2 at 4.2 K and are therefore favorable for SC devices as well as large-scale applications. In addition, they show extremely robust superconductivity to irradiation-induced disorder controlled by the dose of Kr-ion irradiation. The superconductivity of the HEA films is more than 1000 times more resistant to displacement damage than that of other promising superconductors with technological applications, such as MgB 2 , Nb 3 Sn, Fe-based superconductors, and high- T c cuprate superconductors. These results demonstrate that HEA superconductors have considerable potential for use under extreme conditions, such as in aerospace applications, nuclear fusion reactors, and high-field SC magnets. Thin-film high-entropy alloy (HEA) superconductors have recently attracted a lot of attention, but their critical current density and potential usefulness in engineering applications has remained unclear. Here, the authors fabricate HEA films with remarkably high critical current density and resistance to radiation damage.

In nuclear fusion reactors, plasmas are heated to very high temperatures of more than 100 million kelvin and, in so-called tokamaks, they are confined by magnetic fields in the shape of a torus. Light nuclei, such as deuterium and tritium, undergo a fusion reaction that releases energy, making fusion a promising option for a sustainable and clean energy source. Tokamak plasmas, however, are prone to disruptions as a result of a sudden collapse of the system terminating the fusion reactions. As disruptions lead to an abrupt loss of confinement, they can cause irreversible damage to present-day fusion devices and are expected to have a more devastating effect in future devices. Disruptions expected in the next-generation tokamak, ITER, for example, could cause electromagnetic forces larger than the weight of an Airbus A380. Furthermore, the thermal loads in such an event could exceed the melting threshold of the most resistant state-of-the-art materials by more than an order of magnitude. To prevent disruptions or at least mitigate their detrimental effects, empirical models obtained with artificial intelligence methods, of which an overview is given here, are commonly employed to predict their occurrence—and ideally give enough time to introduce counteracting measures. Tokamak plasmas are prone to sudden collapses that terminate the nuclear fusion reactions. This perspective discusses the prediction of these so-called disruptions with artificial intelligence techniques.

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.

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.

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.

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.

Refractory high entropy alloys have broad application prospects due to their excellent comprehensive properties in high temperature environments, and they have been widely implemented in many complex working conditions. According to the latest research reports, the preparation technology of bulk and coating refractory high entropy alloys are summarized, and the advantages and disadvantages of each preparation technology are analyzed. In addition, the properties of refractory high entropy alloys, such as mechanical properties, wear resistance, corrosion resistance, oxidation resistance, and radiation resistance are reviewed. The existing scientific problems of refractory high entropy alloys, at present, are put forward, which provide reference for the development and application of refractory high entropy alloys in the future, especially for plasma-facing materials in nuclear fusion reactors.

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.