The properties of boron carbide and its coating, which seem important for the plasma-facing material in thermonuclear facilities, are presented. The basic steps for boron carbide (B4C) coating using carborane (C2B10H12) as the initial material are reported. It has been shown that is not susceptible to “chemical sputtering.” The sputtering rate of B4C is 3–4 times and the emission of carbon atoms is 15–20 times less than that of graphite and changes slightly up to 1500 K. Trapping of hydrogen isotope ions into the B4C coating tends to saturation at irradiation doses of about 7 × 1023 at/m2. A method for low-temperature “detritization” of the coating is proposed. Conditioning the coating by irradiating it with hydrogen ions with an energy of 50 eV left ≈8

An overview of the MEPhIST-0 educational and research small-scale spherical tokamak project is presented including the vacuum vessel, magnetic field systems, and diagnostics. In contrast to other small machines, it is an advanced tokamak with D-shaped plasma equipped with an electron-cyclotron resonance pre-ionization system for plasma startup and an ion-cyclotron resonance system for plasma heating and wall conditioning. The design choices taken are discussed from the perspective of a primarily educational installation. The machine design is simplified while remaining relevant to larger devices. First plasma results obtained in 2021 are presented.

The characteristics of fine-grained graphite, which will be used as the material for plasma-contacting elements of the T-15MD tokamak, are studied in this work. The density and porosity, thermal diffusivity and thermal conductivity, sizes of the crystalline grains and the amount of impurities in graphite are measured. The measurement results are compared with the corresponding characteristics of MPG-6, MPG-7, and MPG-8 graphite samples. The nature of the retention of hydrogen isotopes and methane in graphite and the conditions of desorption are determined depending on the temperature of preliminary annealing, its duration, and the exposure time of annealed samples in atmospheric gas under normal conditions. The effect of irradiation with deuterium ions of various energies on the regularities of hydrogen trapping and desorption is also considered. In all cases, attention is paid to the effect of experimental conditions on the retention and desorption of hydrogen, which remained in graphite from the time of its production and was trapped during exposure to air. Based on the obtained data and taking into account the expected conditions in the chamber of the T-15MD tokamak, the optimal conditions for annealing of graphite supplied from the manufacturer are identified, and the temperatures of the tokamak elements in contact with the plasma, which contribute to the removal of hydrogen from the graphite lining, are determined.

The diagnostic methods and devices used to study the interaction of plasma with materials in modern fusion facilities are considered. Based on the analysis of methods and devices, it is concluded that the measurements performed by them, separated in space and time, do not provide a sufficiently complete amount of information necessary to identify the mechanisms, regularities, and parameters of the development of processes on the surface of the first wall under plasma exposure. A multifunctional diagnostic probe is proposed, which includes materials science and spectrometric parts, as well as single and double Langmuir and Mach probes. The multifunctional probe is designed to carry out a set of mass spectrometric and materials science measurements necessary to identify the mechanisms and regularities of phenomena on the surface of the first wall of a thermonuclear facility under intense plasma irradiation, as well as to measure the parameters of impurity flows into the plasma.

In this work we investigated the possibility of increasing the efficiency of aluminum oxide protective coating on EP-823 steel by applying a sublayer consisted of aluminum and steel components before the coating deposition. It is shown that Fe-Cr-Al sublayer is able to prevent the steel corrosion in a molten lead flow in the case of damage of the oxide coating. In addition, the sublayer reduces the risk of delamination of aluminum oxide and slows down the transport of iron atoms from the steel towards the surface.

In this paper, features of the modification of near-surface layer of R6510P100D01 grade graphite under irradiation by deuterium and helium ions at a temperature of 2050 °C is investigated. It is shown that irradiation by such ions at a flux density of 1,4×10 18 ion/cm 2 s initiates vacancy transport from the irradiated surface to the bulk of the sample, leading to the formation of a porous layer inside the sample with a thickness of over 2 mm. High sputtering yields, exceeding those for irradiation by the same ions at room temperature with a moderate intensity ion fluxes, lead to the conclusion that in these conditions, sublimation is the main mechanism of the removal of carbon atoms from the surface of graphite.

The results of the measurement of physical properties, sputtering yield during irradiation by hydrogen ions and a study of surface morphology modification of graphite to be used for the plasma-facing materials of the T-15MD tokamak are presented in this work.

Tungsten is going to be implemented as a material for plasma-facing divertor tiles in an ITER tokamak. Accumulation of dust particles due to exfoliation of redeposited tungsten layers can severely affect the operation of a fusion devices. Behaviour of tungsten films deposited on tungsten, including layers with added impurities of iron (0,2 at.% and 2 at.% concentration), as well as dust particles with Al impurity, has been studied. It is shown that a film with 2 at.% impurity of iron exhibits exfoliation and blistering when exposed to thermal and radiation loads, while tungsten dust particles with aluminium impurity undergo various structural changes depending on their Al content when heated.

In this work, the results of the irradiation of untreated, preliminarily Ar+-bombarded and Cr-coated samples of zirconium alloy E110 (Zr-1%Nb) with deuterium atoms and ions of deuterium plasma were compared with the results of their exposure to superheated water steam (673 K, 11 MPa; 673 K, 0.1 MPa; 633 K, 0.1 MPa). It was concluded that, despite of the difference in the rates of the oxide layer formation, the features of hydrogenation are similar under the two kinds of impact. However, the hydrogen uptake under irradiation occurs hundreds of times faster than during the steam test. The most effective hydrogenation was achieved under irradiation of the E110 sample with 300-500 eV/at ions of deuterium plasma without oxygen. These circumstances allow suggesting the irradiation with ions of hydrogen plasma as an accelerated hydrogenation test of zirconium alloys and protective coatings under conditions simulating the environment of light-water reactor's core. The proposed method could be realized using a simple plasma device and it does not require high power consumption or special qualification of staff. (C) 2019 Hydrogen Energy Publications LLC. Published by Elsevier Ltd. All rights reserved.

It was shown (Buzhinsky, 2003) that in situ renewable coating of boron carbide can protect the tiles of the divertors of thermonuclear facilities from destruction and also to prevent accumulation of remarkable amounts of tritium in the plasma facing materials. In the paper presented a plasma method for deposition of boron carbide coating with a high adhesion to tungsten was developed. In the laboratory installation boron carbide coating on tungsten was subjected to cycling irradiation by the deuterium ion flux with power density up to 5.0 MW/m(2) in the temperature range up to 1500 K. The results of the tests showed that the composition, integrity and adhesion of the coating were not violated in the laboratory tests. In the T-10 tokamak the behavior of the coating was investigated in the temperature range up to 3600 K when irradiated with plasma power in the range of 20-100 MW/m(2) during plasma disruption. Being irradiated in T-10 tokamak, the coating retained its continuity, adhesion and protected tungsten from the effect of the even at temperatures of 2500-3600 K, when the coating melted under irradiation and its composition changed to B:C approximate to 1:1.

Abstract In this paper, a dependence of the erosion of graphite under a high energy ion flux at temperature of 2050°C on the ion dose is investigated. It is shown that, at ion flux densitiy of 1.42×1022 ion/m2s, irradiation stimulates diffusion processes that lead to the removal of carbon atoms from the bulk of the sample, leading to the formation of a porous layer, whereas, for ion flux density of 1.4×1020 ion/m2s, no such layer is formed.

Abstract Formation of dust particles and clusters is observed in almost all modern fusion devices. Accumulation of dust in next-generation thermonuclear installations can significantly affect plasma parameters and lead to accumulation of unacceptably large amounts of tritium. The use of a specially developed electrostatic probe is planned in the international thermonuclear experimental reactor ITER to collect dust for further analysis. The article describes a numerical model of dust particles movement in an electrostatic probe. Dust particles trajectories inside the probe were analyzed. Several electrostatic probe design modifications were proposed on the basis of the analysis in order to increase the efficiency of dust collection.

In this paper, a dependence of the erosion of graphite under a high energy ion flux at temperature of 2050 °C on the ion dose is investigated. It is shown that high doses of irradiation stimulate diffusion processes that lead to the removal of carbon atoms from the bulk of the sample, significantly altering morphology of graphite at depths exceeding penetration depth of irradiating ions for large doses.

In the paper, a simple and easy to operate stand is presented that allows irradiating the surface of solids by gas discharge plasma ions in the energies ranging from 5 to 40 keV. The results of first experiments are shown. Ion current parameters on the irradiated sample using gas discharge are presented for different gases.

The results of study on hydrogen and deuterium trapping and retention in MPG-8 grade graphite and boron carbide coating under consecutive irradiation by deuterium and hydrogen plasma in varying hydrogen irradiation conditions are presented in this paper. It is shown that deuterium content decreases both in graphite and boron carbide under irradiation by hydrogen plasma. It is also shown that the main mechanism of deuterium removal is sputtering for graphite, and isotope exchange for boron carbide.

The paper has investigated conditions and parameters of argon trapping in molybdenum, tantalum and tungsten layers during their deposition on tungsten substrate by the atoms sputtered from the respective targets in argon plasma. The substrate temperature during deposition was 1273 K. The rate of deposition was 1 mu m/h. It was shown that electron irradiation of the deposited layer with the beam intensity of 4 mA/cm(2) initiated argon trapping in tungsten and tantalum coating with approx. 2 x 10(27) at/m(3) and 8 x 10(26) at/cm(3), respectively, but did not stimulate argon trapping in the molybdenum layers. Features of argon trapping in the tungsten coating and its release are investigated in detail.

A brief overview of the high heat flux material testing facilities is presented which forms background for a new small-scale laboratory device capable of high heat flux material testing. The scheme of the device is presented and its working principles are described. High heat and particle loads are achieved by focusing of either electron or ion beam on the sample being tested. The sample is fixed to an actively cooled copper table. Experiments with hydrogen ion beam irradiation of an MPG-8 graphite sample are presented, in which heat loads of 270 MW/m(2) were achieved. SEM images of the sample surface after testing are presented.

The paper considers an influence of T-10 tokamak plasma disruption on boron carbide (B4C) coating on tungsten. The power density of coating irradiation reached 100 MW/m2. The relief and composition of the boron carbide coating sample areas heated up to different temperature due to influence of disruption is determined. Conclusion is made that B4C does not change integrity, when heated to temperatures of up to 2000 K. Local melting was observed in areas heated up to 2500 K. In the range of 2500–3600 K most of the coating was melted and collected into droplets. Composition rate is reduced to B:C = 1:1. In all temperature ranges the coating remained continuous and provided protection of tungsten from direct plasma irradiation.

Tungsten layers with iron impurity were deposited on tungsten substrates modeling re-deposited layers in a fusion device. The samples were tested by thermocycling and hydrogen ion beam tests. Thermocycling revealed globule formation on the surface. The size of the globules depended on iron impurity content in the coating deposited. Pore formation was observed which in some cases lead to exfoliation of the coatings. Hydrogen ion irradiation lead to formation of blisters on the coating and finally its exfoliation.