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Tungsten is foreseen presently as the plasma-facing material for divertors in fusion power plants. In order to achieve durable operation of divertors of current fusion reactors, an efficient way of maintaining the divertor functionality is needed. A system capable of in situ tungsten coating of the divertor via low-pressure plasma spraying was proposed to maintain the divertor integrity. In this work, tungsten was deposited on NB31 carbon fibre composite substrates using the low-pressure plasma spraying technology to evaluate the feasibility of this technique. The thickness, porosity, composition, adhesion, and microstructure of the coatings were investigated by scanning electron microscopy image analysis and energy dispersive spectroscopy. Based on the initial results, the spray parameters were iteratively improved in a campaign-based study. The coatings exhibited improving properties through an adjusting of the carrier gas flow, the scanning speed, and the spray distance. By lowering the carrier gas flow, the porosity of the coatings was reduced, resulting in coatings of 98% bulk density. Adjusting the carrier gas flow reduced the amount of semi-molten particles in the coatings significantly. A decrease in both scanning speed and spray distance increased the substrate’s temperature, which led to better adhesion and porosity.

In this paper, for the first time, we present a new model of current distribution in a tungsten sample and of substance evaporation when the surface is heated by an electron beam. The model is based on solving electrodynamics equations in a cylindrical coordinate system using a model temperature distribution in the sample and a thin layer of evaporated tungsten. The model is analyzed in a simplified formulation at constant values of electrical resistance and thermoelectric power in gas and metal. The dependence of the amplitude and isolines of thermal currents on the distribution of temperature at the sample surface is shown. The model parameters are taken from experiments at the Beam of Electrons for materials Test Applications (BETA) facility, created at the Budker Institute of Nuclear Physics of the Siberian Branch of the Russian Academy of Sciences.

The influence of extreme heat loads, as produced by a multiple pulses of non-homogeneous fl ow of slow plasma (0.1-1 keV) and fast ions (100 keV), on double-forged tungsten (DFW) was investigated. For generation of deuterium plasma and fast deuterons, plasma-focus devices PF-12 and PF-1000 are used. Depending on devices and conditions, the power flux density of plasma varied in a range of 10 -10 W/cm with pulse duration of 50-100 ns. Power flux density of fast ions was 10 -10 W/cm at the pulse duration of 10-50 ns. To achieve the combined effect of different kind of plasmas, the samples were later irradiated with hydrogen plasma (10 W/cm , 0.25 ms) by a QSPA Kh-50 plasma generator. Surface modification was analysed by scanning electron microscopy (SEM) and microroughness measurements. For estimation of damages in the bulk of material, an electrical conductivity method was used. Investigations showed that irradiation of DFW with multiple plasma pulses generated a mesh of micro- and macrocracks due to high heat load. A comparison with single forged tungsten (W) and tungsten doped with 1% lanthanum-oxide (WL10) reveals the better crack-resistance of DFW. Also, sizes of cells formed between the cracks on the DFW’s surface were larger than in cases of W or WL10. Measurements of electrical conductivity indicated a layer of decreased conductivity, which reached up to 500 μm. It depended mainly on values of power flux density of fast ions, but not on the number of pulses. Thus, it may be concluded that bulk defects (weakening bonds between grains and crystals, dislocations, point-defects) were generated due to mechanical shock wave, which was generated by the fast ions flux. Damages and erosion of materials under different combined radiation conditions have also been discussed.

Tungsten (W) has a high melting point, excellent thermal conductivity, and irradiation resistance, making it the most promising plasma facing material for divertors in fusion reactors, which are currently under development. However, since the divertor is exposed to an extremely harsh environment, it is considered necessary to develop suitable and cost-effective repair techniques. In this study, the applicability of the atmospheric plasma spraying (APS) method using a gas shroud as a repair technique for W components was investigated, in particular the possibility of strengthening the repaired part by applying friction stir processing (FSP) as a post-treatment. It was found that the application of a gas shroud can suppress in-flight oxidation to some extent, even when the W is deposited in air. In addition, the FSP treatment reduced grain size and porosity, resulting in an increase in microhardness of approximately 37.5% compared to the base material (W substrate) and 203.5% compared to the as-sprayed material. The gas shroud APS and FSP post-treatments have been shown to have potential as repair techniques for tungsten components in future fusion reactors.

Copper alloys are critical heat sink materials for fusion reactor divertors due to their high thermal conductivity (TC) and strength, yet their performance under extreme particle bombardment and heat fluxes in future tokamaks requires enhancement. While neutron-induced transmutation helium affects the properties of copper, the atomistic mechanisms linking helium bubble size to thermal transport remain unclear. This study employs non-equilibrium molecular dynamics (NEMD) simulations to isolate the effect of bubble diameter (10, 20, 30, 40 Å) on TC in copper, maintaining a constant He-to-vacancy ratio of 2.5. Results demonstrate that larger bubbles significantly impair TC. This reduction correlates with increased Kapitza thermal resistance and pronounced lattice distortion from outward helium diffusion, intensifying phonon scattering. Phonon density of states (PDOS) analysis reveals diminished low-frequency peaks and an elevated high-frequency peak for bubbles >30 Å, confirming phonon confinement and localized vibrational modes. The PDOS overlap factor decreases with bubble size, directly linking microstructural evolution to thermal resistance. These findings elucidate the size-dependent mechanisms of helium bubble impacts on thermal transport in copper divertor materials.

National Institute for Fusion Science (NIFS) launched in 2014 a research program for developing Dispersion Strengthened (DS) Cu alloys for application to the heat sink materials of divertors of fusion reactors, using newly installed ball-milling, encapsulation, and Hot Isostatic Pressing (HIP) facilities. A unique feature of these facilities is that the entire process can be performed without exposing the materials to air, enabling precise impurity control. Cu-Al, Cu-Zr and Cu-Y alloys have been produced in this program. Various technological advancement has been made for the fabrication, such as suppression of powder adhesion to the wall of containers during the ball milling, and encapsulation technology including development of small volume tubular capsules.

Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, λ iz , and the width of the magnetic sheath rather than the ratio of λ iz and the W + gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to W 5 + may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

This paper presents the results of experimental studies of the interaction of helium plasma with a near-surface tungsten carbide layer. The experiments were implemented at the plasma-beam installation. The helium plasma loading conditions were close to those expected in the ITER divertor. The technology of the plasma irradiation was applied in a stationary type linear accelerator. The impact of the helium plasma was realized in the course of the experiment with the temperatures of ~905 °C and ~1750 °C, which were calculated by simulating heat loading on a tungsten monoblock of the ITER divertor under the plasma irradiation at the load of 10 MW/m2 and 20 MW/m2, respectively. The structure was investigated with scanning microscopy, transmitting electron microscopy and X-ray analysis. The data were obtained showing that the surface morphology changed due to the erosion. It was found that the carbidization extremely impacted the plasma–tungsten interaction, as the plasma–tungsten interaction with the carbide layer led to the carbon sputtering and partial diffusion towards to the depth of the sample. According to these results, WC-based tungsten carbide is less protected against fracture by helium than W and W2C. An increase in temperature leads to much more extensive surface damage accompanied by the formation of molten and recrystallized flanges.

Micro-tensile testing has been used to study the response of pure tungsten and two tungsten alloys to helium ion irradiation. Commercially supplied plates of W, W-5Ta and W-5Re were irradiated using 6 MeV helium ions at room temperature. The ion energy was attenuated with an energy spreading device such that a uniform level of damage at 0.6 dpa (and 11,000 appm He) was deposited at the 3–9 µm depth. Focused ion beam milling was used to fabricate dog-bone shaped, micro-tensile samples 5 × 5 µm in cross-sectional area and 17 µm in length from the unirradiated and irradiated samples. All micro-tensile samples were tested at a quasi-static strain rate and the stress–strain curves were analysed to determine the mechanical properties. A close correlation was found between micro-tensile results and the bulk mechanical properties reported in the literature. Comparison between the unirradiated micro-tensile properties of W-5Re and W-5Ta with W showed that, as expected, W-5Re was softer than W whilst W-5Ta had only minor differences in micro-tensile properties compared with W. The micro-tensile results of the irradiated W, W-5Ta and W-5Re showed an increase in strength and an almost complete loss of ductility compared to the unirradiated samples. In comparing micro-tensile results to nanoindentation measurements, it was found that micro-tensile offers comparable level of precision in measurement of irradiation hardening amongst W, W-5Ta and W-5Re. The implications of the results with respect to the future performance of tungsten-based materials in the divertors in fusion reactors are discussed in detail. Graphical abstract

This article describes the study of low pressure helium plasma with magnetic confinement at the experimental test setup at the Moscow Power Engineering Institute: plasma linear multicusp (PLM). This facility is intended for testing refractory materials and prototypes of elements of the first wall within the framework of development of the national fusion reactor (DEMO–FNS) and the International Thermonuclear Experimental Reactor (ITER). The facility provides the conditions of plasma impact on the surface of tested sample close to the parameters and regime of operation of tokamak divertor plates. The facility is a magnetic trap with minimum magnetic field on the axis, where the plasma is created by the flow of electrons moving from the directly heated tantalum cathode toward the anode. It is possible to create stationary helium plasma in the facility and to maintain it for several hours under constant discharge parameters: helium pressure in the chamber of 10 –3 –10 –1 Torr, discharge current of 4–30 A, plasma column diameter of 35–40 mm, voltage drop across the discharge gap of 100–200 V. The thermal load on the surface of target introduced into the axial region of plasma column has reached 5 MW/m 2 . Optical emission spectroscopy is the main diagnostic tool in this work. The procedure for determining atomic concentrations from the data on the relative intensities of atomic spectral lines of metallic impurities is proposed in this work.