Search Results (110)

Tungsten is a prime candidate material for the plasma-facing components of fusion reactors, thanks to its high melting point, high temperature strength, good thermal conductivity, high erosion resistance, etc. Yet, it has some limitations, mainly its brittle nature, difficulty of machining, and propensity to recrystallize at elevated temperatures. Among the approaches to the improvement of particular properties are alloying, dispersion strengthening, thermomechanical processing, and modifications to the sintering process. This study explores the possibility of combining fine powder size with ultra-high pressure to achieve significant densification at moderate temperatures during hot pressing. Two powder sizes and a range of temperatures from 1000 to 2000 °C were used, and their effects were observed. The resulting tungsten compacts were characterized for their microstructure, density, and mechanical and thermal properties. The high pressure enabled substantial densification already at relatively low temperatures, thanks to the plastic deformation of the powder particles. A significant degree of sintering, as manifested by the microstructural and property evolution, occurred however only at higher temperatures. The compacts exhibited brittleness, calling for further optimization of the method. Full article

Magnetic confinement nuclear fusion offers a promising solution to the world’s growing energy demands. The DEMO reactor presented here aims to bridge the gap between laboratory fusion experiments and practical electricity generation, posing unique challenges for magnetic plasma diagnostics due to limited space for diagnostic equipment. This study employs Bayesian inference and Gaussian process modeling to integrate data from pick-up coils, flux loops, and saddle coils, enabling a qualitative estimation of the plasma current density distribution relying on only external magnetic measurements. The methodology successfully infers total plasma current, plasma centroid position, and six plasma–wall gap positions, while adhering to DEMO’s stringent accuracy standards. Additionally, the interchangeability between normal pick-up coils and saddle coils was assessed, revealing a clear preference for saddle coils. Initial steps were taken to utilize Bayesian experimental design for optimizing the orientation (normal or tangential) of pick-up coils within DEMO’s design constraints to improve the diagnostic setup’s inference precision. Our approach indicates the feasibility of Bayesian integrated data analysis in achieving precise and accurate probability distributions of plasma parameter crucial for the successful operation of DEMO. Full article

Tungsten (W) is regarded as the most promising plasma-facing material in thermonuclear fusion reactors due to its excellent properties, such as high strength, a high melting point, and a low sputtering rate. However, its low-temperature brittleness, recrystallization embrittlement, and irradiation embrittlement seriously limit the practical application of W. In this research, the properties of tungsten-based materials were improved by introducing second phases into W. Core–shell composite powders with W particles as core and Sm(OH)₃ thin films as shell were prepared by electroless plating, and sintered by spark plasma sintering (SPS) to obtain bulk. After sintering, the Sm(OH)₃ shell transformed into the Sm₂O₃ phase with a different size, mainly distributed at W grain boundaries. The average size of W grains in the composite material was smaller than that of pure W sintered bulk due to the pinning of W grain boundaries by Sm₂O₃, while the porosity of the composite is reduced. Compared with pure W sintered bulk, the composites exhibited better mechanical properties and radiation resistance; although the thermal conductivity decreased somewhat, it still maintained a high level. With the increase in sintering temperature and pressure, the evolution of core–shell powders during the sintering process could be simplified into six stages, which occurred approximately in sequence. Full article

The pursuit of compact and efficient fusion energy systems necessitates innovative structural concepts capable of withstanding extreme operational conditions. This study presents a preliminary structural evaluation and stress assessment of a bucked-and-wedged configuration for the EU DEMO tokamak, targeting a peak magnetic field of 16.5 T. The proposed concept leverages mutual wedging of the Toroidal Field (TF) coils and their interaction with the Central Solenoid (CS) to optimize stress distribution in the inner legs, a critical region in high-field fusion reactors. To address the significant tangential forces arising during plasma operation, the design integrates outer inter-coil structures and shear pins to enhance mechanical stability. A hybrid simulation approach—coupling 3D electromagnetic and structural finite element analyses—is employed to assess stress behavior and structural integrity under both in-plane and out-of-plane loading conditions. The results contribute to the optimization study of high-field fusion reactor components and offer insights into viable mechanical design strategies for next-generation nuclear energy systems. Full article

This study assesses how different anode materials influence neutron production rates (NPRs) in multi-state fusion (MSF) reactors, with a particular focus on the effects of deuterium (D) pre-loading on the anode surface. Three types of mesh anodes were assessed: stainless steel (SS), zirconium (Zr), and D pre-loaded zirconium (ZrD). MSF operates using two electrodes to confine ions to various fusion reactions, including D-D and D-T. The reactor features a negatively biased central cathode and a grounded anode within a vacuum vessel. Neutrons and protons are produced through the application of high voltage (tens of kV) and current (tens of mA) on the system to spark the plasma and start the fusion. Assessments at voltages up to 50 kV and currents up to 30 mA showed that Zr mesh anodes produced higher NPRs than SS ones, reaching 1.912 at 30 kV. This increased performance is attributed to surface fusion processes occurring in the anode. These processes were further modified by the deuterium pre-loading in the ZrD anode, as compared to SS and Zr with 1.832 at 30 kV. The findings suggest that material properties and deuterium pre-loading play significant roles in optimizing the efficiency of MSF reactors and the NPR. Future research may explore the long-term stability and durability of these anode materials under continuous operation conditions to fully harness their potential in fusion energy applications. Full article

Helium (He) accumulation in tungsten—widely used as a plasma-facing material in fusion reactors—can lead to clustering, trap mutation, and eventual formation of helium bubbles, critically impacting material performance. To clarify the atomic-scale mechanisms governing this process, we conducted systematic molecular statics and molecular dynamics simulations across a wide range of vacancy cluster sizes (n = 1–27) and temperatures (500–2000 K). We identified the onset of trap mutation through abrupt increases in tungsten atomic displacement. At 0 K, the critical helium-to-vacancy (He/V) ratio required to trigger mutation was found to scale inversely with cluster size, converging to ~5.6 for large clusters. At elevated temperatures, thermal activation lowered the mutation threshold and introduced a distinct He/V stability window. Below this window, clusters tend to dissociate; above it, trap mutation occurs with near certainty. This critical He/V ratio exhibits a linear dependence on temperature and can be described by a size- and temperature-dependent empirical relation. Our results provide a quantitative framework for predicting trap mutation behavior in tungsten, offering key input for multiscale models and informing the design of radiation-resistant materials for fusion applications. Full article

Refractory high-entropy alloys (HEAs) are promising candidates for next-generation nuclear applications, particularly fusion reactors, due to their excellent high-temperature mechanical properties and irradiation resistance. Here, the microstructure and mechanical behavior were investigated for an equimolar WTaTiVZr HEA, designed from a palette of low-activation elements. The as-cast alloy exhibited a dendritic microstructure composed of W-Ta rich dendrites and Zr-Ti-V rich inter-dendritic regions, both possessing a body-centered cubic (BCC) crystal structure. Room temperature bulk compression tests showed ultra-high strength of around 1.6 GPa and plastic strain ~6%, with fracture surfaces showing cleavage facets. The alloy also demonstrated excellent high-temperature strength of ~650 MPa at 500 °C. Scratch-based fracture toughness was ~38 MPa√m for the as-cast WTaTiVZr HEA compared to ~25 MPa√m for commercially used pure tungsten. This higher value of fracture toughness indicates superior damage tolerance relative to commercially used pure tungsten. These results highlight the alloy’s potential as a low-activation structural material for high-temperature plasma-facing components (PFCs) in fusion reactors. Full article

Powder injection molding (PIM) has been used to produce nearly net-shaped samples of tungsten-based alloys. These alloys have been previously shown to have favorable characteristics when compared with standard ITER-grade tungsten. Six different alloys were produced with this method: W-1TiC, W-2Y₂O₃, W-3Re-1TiC, W-3Re-2Y₂O₃, W-1HfC and W-1La₂O₃-1TiC. These were tested alongside ITER-grade tungsten in the PSI-2 linear plasma device under ITER-relevant plasma and heat loads to assess their suitability for use in a fusion reactor. All materials showed good behavior when exposed to the lower pulse number tests (≤1000 ELM-like pulses), although standard tungsten performed slightly better, with no observable difference in surface roughness. High-power shots, namely one laser pulse of 1.6 GWm⁻², revealed that samples containing yttria are more prone to melting and droplet ejection. After high pulse number tests (10,000 and 100,000 pulses), with and without plasma, the reference tungsten showed the most cracking and highest surface roughness of all materials, while the PIM samples seemed to have a higher resistance to cracking. This can be attributed to the higher ductility of these alloys, particularly those containing rhenium. This means that tungsten-based alloys, whether produced via PIM or other methods, could potentially be used in certain areas of a fusion reactor. Full article

Compact toroid (CT) injection, with its characteristics of high plasma density and extremely high injection velocity, is considered a highly promising method for core fueling in fusion reactors. Previous studies have lacked investigation into the transport process of CT within drift tubes. To investigate the dynamic processes of CT in drift tubes, this study developed a compressible magnetohydrodynamics (MHD) solver and a magnetic diffusion solver based on the OpenFOAM platform. They were integrated into a multi-region coupling framework to create a multi-region coupled MHD solver, mhdMRF, for simulating the dynamic behavior of CT in drift tubes and its interaction with finite-resistivity walls. The solver demonstrated excellent performance in simulations of the Orszag–Tang MHD vortex problem, the Brio–Wu shock tube problem, analytical verification of magnetic diffusion, and validation of internal coupling boundary conditions. Additionally, this work innovatively explored the effects of the geometric structure at the end of the inner electrode and finite-resistivity walls on the transport processes of CT. The results indicate that optimizing the geometric structure at the end of the inner electrode can significantly enhance the confinement performance and stability of CT transport. The resistivity of the wall profoundly influences the magnetic field structure and density distribution of CT. Full article

In this study, particle-in-cell (PIC) simulation was used to validate a conceptual design for a quasi-spherical, net power, hydrogen-plus-boron-11-fueled fusion reactor incorporating high-temperature superconducting (HTS) magnets. By burning a fully thermalized plasma, our proposed MET6 reactor uses the principles of the 1980 magneto-electrostatic trap design of Yushmanov to improve the classic Polywell design. Because the input power consumed by the reactor will barely balance the waste bremsstrahlung radiation, future research must focus on reducing the bremsstrahlung losses to reach practical net power levels. The first step to reducing bremsstrahlung, explored in this paper, is to tune the reactor parameters to reduce the energies of trapped electrons. We assume the quality factor Q can be approximated as the ratio of fusion power output divided by bremsstrahlung power loss. Thus, assuming the particles’ power loss is negligible compared to bremsstrahlung power loss, the resulting quality factor is estimated to be Q ≈ 1.3. Full article

Beryllium-based intermetallic compounds, such as Be₁₂Nb, are attracting growing interest for their high thermal stability and potential to replace pure beryllium as neutron reflectors and multipliers in both fission and future fusion reactors, with additional applications in metallurgy, aerospace, and hydrogen technology. The paper presents the results of an investigation of the thermal treatment and phase formation of the intermetallic compound Be₁₂Nb from a mixture of niobium and beryllium powders in the temperature range of 800–1300 °C. The phase evolution was assessed as a function of sintering temperature and time. A nearly single-phase Be₁₂Nb composition was achieved at 1100 °C, while decomposition into lower-order beryllides such as Be₁₇Nb₂ occurred at temperatures ≥1200 °C, indicating thermal instability of Be₁₂Nb under vacuum. Careful handling of sintering in low vacuum minimized oxidation, though signs of possible BeO formation were noted. The findings complement and extend earlier reports on Be₁₂Nb synthesis via plasma sintering, mechanical alloying, and other powder metallurgy routes, providing broader insight into phase formation and synthesis. These results provide a foundation for optimizing the manufacturing parameters required to produce homogeneous Be₁₂Nb-based components and billets at an industrial scale. Additionally, they help define the operational temperature limits necessary to preserve the material’s phase integrity during application. Full article

Overload detection is crucial in nuclear fusion experiments to prevent damage to plasma-facing components (PFCs) and ensure the safe operation of the reactor. At Wendelstein 7-X (W7-X), real-time monitoring and prediction of thermal events are essential for maintaining the integrity of PFCs. This paper proposes a machine learning approach for developing a real-time overload detector, trained and tested on OP1.2a experimental data. The analysis showed that Self-Organizing Maps (SOMs) are efficient in detecting the overload risk starting from a set of plasma parameters that describe the magnetic configuration, the energy behavior, and the power balance. This study aims to thoroughly evaluate the capabilities of the SOM in recognizing overload risk levels, defined by quantizing the maximum criticality across different IR cameras. The goal is to enable detailed monitoring for overload prevention while maintaining high-performance plasmas and sustaining long pulse operations. The SOM proves to be a highly effective overload risk detector. It correctly identifies the assigned overload risk level in 87.52% of the samples. The most frequent error in the test set, occurring in 10.46% of cases, involves assigning a risk level to each sample adjacent to the target one. The analysis of the results highlights the advantages and drawbacks of criticality discretization and opens new solutions to improve the SOM potential in this field. Full article

This study investigates the synergistic effects of pulsed heat load and helium plasma irradiation on the surface damage evolution of high-purity tungsten, a candidate plasma-facing material (PFM) for future fusion reactors. Using a self-developed linear plasma device, tungsten samples were exposed to controlled single-pulse heat loads (32–124 MW·m⁻²) and helium plasma fluxes (7.76 × 10²²–2.40 × 10²³ ions·m⁻²·s⁻¹). SEM and XRD analyses revealed a progressive damage mechanism involving helium bubble formation, pit collapse, coral-like nanostructure evolution, and melting-induced restructuring. These surface changes were accompanied by grain refinement, lattice contraction, and peak shifts in the (110) diffraction plane. Mechanical testing showed a flux-dependent variation in hardness, with initial hardening followed by softening due to crack propagation. Surface reflectivity significantly declined with increasing load, indicating severe optical degradation. This work demonstrates the nonlinear coupling between thermal and irradiation effects in tungsten, offering new insights into damage accumulation under realistic reactor conditions. The findings highlight the dominant role of transient heat loads in driving structural and property changes and emphasize the importance of accounting for synergistic effects in material design. These results provide essential experimental data for optimizing PFMs in divertor and first-wall applications and suggest directions for future research into cyclic loading, long-term exposure, and microstructural recovery mechanisms. Full article

The exploitation of fusion energy in tokamak reactors relies on efficient and reliable tritium management. The tritium needed to sustain the deuterium–tritium fusion reaction is produced in the Li-based blanket surrounding the plasma chamber, and, therefore, the effective extraction and purification of the tritium bred in the Li-blankets is needed to guarantee the tritium self-sufficiency of future fusion plants. This work introduces a new technology for the extraction of tritium from the Pb–Li eutectic alloy used in liquid blankets. Process units based on the concept of Membrane Gas–Liquid Contactor (MGLC) have been studied for the extraction of tritium from the Pb–Li in the Water Cooled Lithium Lead blankets of the DEMO reactor. MGLC units have been preliminarily designed and then compared in terms of the permeation areas and sizes with the tritium extraction technologies presently under study, namely the Permeator Against Vacuum (PAV) and the Gas–Liquid Contactors (GLCs). The results of this study show that the DEMO WCLL tritium extraction systems using MGLC require smaller permeation areas and quicker permeation kinetics than those based on PAV (Permeator Against Vacuum) devices. Accordingly, the MGLC extraction unit exhibits volumes smaller than those of both PAV and GLC. Full article

The selection of a suitable plasma-facing material (PFM) that must protect the divertor, cooling systems, and structural components is an important challenge in the design of advanced fusion reactors and requires careful consideration. Material degradation due to melting and evaporation may lead to plasma contamination, which must be strictly avoided. Among the candidate materials, tungsten (W) is the most promising because of its thermo-mechanical and physical properties, which allow it to endure repeated exposure to extremely harsh conditions within the reactor. The plasma spraying (PS) technique is gaining increasing interest for the deposition of tungsten (W) coatings to protect heat sink materials, due to its relatively low cost, high deposition rates, and capability to coat complex-shaped surfaces and fix damaged coatings in situ. This review aims to provide a systematic assessment of tungsten (W) coatings produced by PS techniques, evaluating their suitability as PFMs. It discusses W-based materials, plasma spraying technologies, the role of the interface in joining W coating and metallic substrates such as copper alloys and steels, and the main issues related to coating surface erosion under steady-state and transient heat loads associated with advanced fusion reactor operation modes and off-normal events. Quantitative data available in the literature, such as porosity, oxygen content, thermal conductivity of the coatings, residual stresses accumulated in the coating–substrate interface, surface temperature, and material loss following heat load events, were summarized and compared to bulk W ones. The results demonstrate that, following optimization of the fabrication process, PS-W coatings exhibit excellent performance. In addition, previously mentioned advantages of PS technology make PS-W coatings an attractive alternative for PFM fabrication. Full article