The influence of pulsed plasma irradiation on a boron carbide (B4C) coating using the QSPA-T setup was investigated. The duration of the rectangular plasma pulses was 0.5 ms, with intervals between pulses of 5 to 10 min. The maximum power density in the middle of the plasma flow reached 1 GW/m2. The thickness of the coating at various surface locations ranged from 20 to 40 μm. The modification of the surface layers and the transformation of the coating at elevated temperatures during the pulsed plasma irradiation process over four consecutive series of pulses are described. It was shown that the boron carbide coating withstood a complete cycle of tests under irradiation with 100 plasma pulses at a maximum power density of 1 GW/m2. No significant damage was detected on the surface of the coating, and the layer remaining after the tests maintained the crystalline structure of B4C throughout the irradiation zone to a depth of at least 2 μm.

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

The features of hydrogen trapping under the irradiation of zirconium alloys E110 and E635 with an electron beam (400 eV, 0.8 mA/cm 2 ) in various gaseous ambient containing oxygen and hydrogen (Ar + O 2 + H 2 ; Ar + H 2 O) are studied. The effect of surface temperature on the hydrogenation of zirconium alloys under electron irradiation is studied. It is shown that neither exposure nor electron irradiation in a gaseous ambient containing a mixture of H 2 and O 2 at 700 K for 20 h leads to a change in the hydrogen level in the E110 alloy, while hydrogen partially escapes from the E635 alloy during the experiment. The amount of hydrogen in zirconium alloys after exposure in a gaseous ambient containing H 2 O in a similar temperature–time regime also changes insignificantly; however, if the alloys are irradiated with electrons under these conditions, then they contain 1.5–2 times more hydrogen. When temperature of the samples is raised to 900 K, electron irradiation in the presence of water vapor, on the contrary, reduces the trapping of hydrogen in the E110 and E635 alloys, compared with exposure in the same gaseous ambient without irradiation. It is concluded that electron irradiation affects the balance of surface reactions of formation of hydroxyl groups from adsorbed water molecules and reverse reactions of formation of water molecules from surface hydroxyls. The direction of shifting the balance of these reactions depends on the surface temperature.

The choice of plasma contact materials and configurations for power fusion reactors is still not obvious. The tokamak with reactor technologies (TRT) under construction should help resolve this issue. Therefore, the most complete study of the effect of plasma on the tokamak divertor for various types of discharges and an analysis of their consequences are of great importance. Divertor probes are devices that measure thermal and corpuscular flows toward the surface of a divertor and/or record the results of their impact on the divertor. They have found wide application in fusion facilities. The features, advantages, and disadvantages of the divertor probes presented in literature, as well as cases of using samples of materials or devices to solve individual problems to which divertor probes are applicable are discussed in this work. A preliminary design and thermal calculations of the divertor probe for the TRT tokamak are presented, which, according to the authors, is capable of performing a set of measurements that make it possible to draw a conclusion about the mechanisms and regularities of processes on the divertor surface depending on the parameters and conditions of plasma irradiation, as well as to speed up the determination of optimal materials and modes of plasma irradiation of divertor tiles.

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.

In the article, a hydrogen exchange between the Zr-1%Nb alloy (E110) and the gas ambient was experimentally studied when the samples were irradiated with deuterium and argon plasma ions. It has been established that, upon irradiation with deuterium plasma ions with an energy of E = 650 eV/at, the enhanced absorption of deuterium exceeds the release of hydrogen initially contained in the samples, which leads to their loading with hydrogen isotopes. Adding 30 at.% oxygen to the plasma-forming gas or raising the sample temperature from T = 450 K to T = 600 K, significantly reduces the content of hydrogen isotopes in the sample. Based on the aggregate data obtained by atomic and ion irradiation, a mechanism of hydrogen exchange between zirconium alloy and gas ambient is proposed. The process includes three stages: reactions on the oxidized surface of the zirconium alloy (surface hydroxylation and formation of water molecules); reactions at the metal-oxide interface; transfer of hydrogen isotopes through the surface oxide layer in both directions due to hopping between neighboring oxygen ions. Surface reactions caused by irradiation of atoms and ions trigger the hydrogen exchange. The proposed model agrees with the experimental data on the irradiation of the E110 alloy with atoms and ions of hydrogen isotopes.

In this work, thermal-desorption spectrometry is used to study Be–W and Al–W intermetallic compounds formed on the surface of tungsten during the deposition of Be and Al atoms evaporated in the temperature range of 400–500 K, as well as Al atoms sputtered in argon plasma. Due to safety limitations imposed on work with beryllium in the university laboratory, most of the experiments are carried out with aluminum which is one of the proxy materials to beryllium. Regularities for the formation of intermetallics in the surface layer of tungsten during the deposition of Al and Be atoms, and in the layer formed during the joint deposition of Al and W atoms on tungsten are described. The features of the trapping of argon and hydrogen into the intermetallics are analyzed for different fluences of Al atoms deposited in plasma on the surface of tungsten, as well as for different concentrations of components in the deposition flux during the simultaneous deposition of Al and W atoms on tungsten. The features of the Al–W intermetallic formation and the nature of the retention and desorption of gases during the deposition of evaporated and plasma-sputtered aluminum atoms on tungsten are revealed. The release of a significant part of hydrogen from the Al–W intermetallic under irradiation with 50 eV ions of deuterium plasma at a temperature of 500 K is found. Most of the gases trapped by Be–W and Al–W intermetallics is desorbed in the interval of 1200–1600 K during intermetallics decomposition.

The paper investigates the trapping and desorption of hydrogen isotopes under irradiation of zirconium alloy E110 with ions of deuterium plasma. It is established that, under the irradiation, hydrogen exchange between the alloy and the gaseous ambient occurs: part of deuterium from the irradiation flux is trapped by the alloy, and at the same time, hydrogen initially contained in the alloy is partially desorbed. In general, hydrogen exchange results in an increase in the content of hydrogen isotopes in the sample. Addition of 30 at

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.

The features of oxidation and hydrogenation of zirconium alloy E110 under electron and plasma irradiation in various conditions are studied. It is revealed that irradiation intensifies both the oxide layer growth and the hydrogen absorption by zirconium. Applying the anode potential on the E110 sample in plasma allowed to achieve the maximum rates of these processes and at the same time to reproduce their characteristic features exhibited in traditional steam tests. This circumstance determines the choice of plasma anodizing as a perspective basis for the method of accelerated testing of Zr alloys.

Кроме 20 докладов, представленных на 4-й Международной конференции по подкритическим гибридным системам синтеза-деления, одобренных рецензентами и публикуемых в настоящем выпуске

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.

High-quality tungsten coating deposition on sintered aluminum nitride ceramic substrates (both of thin flat chips and structural boxes) was realized using an adapted plasma-aided coating deposition rig. The tungsten coating produced using this technique and the accompanying apparatus setup are of high-purity, strong adhesion, and controlled three-dimensional uniformity (<20% thickness variations). The coating also exhibits well-structured and smooth (Ra < 1.0 mu m) microscopic surface landscape with densely clustered tungsten granulations. The coated samples were tested under load conditions expected during ITER operation, including thermal cycling and superheated (up to 500 degrees C) steam. Exposure to thermal cycles and hot steam made no apparent changes to the coating's microscopic structure with no sign of cracks, blistering, or exfoliation seen under electron microscopy. These successes validated the microwave shield design for the ITER high-frequency magnetic sensor, which is based on this concept, and laid a solid foundation for the production of this component in the forthcoming procurement phase. Besides, a failure test was conducted for the tungsten coating in the temperature range of 500 degrees C to 1500 degrees C. Surface smoothing, pores, delamination, and mass loss in substrate were observed when temperature exceeded 1000 degrees C, possibly due to the evaporation of aluminum atoms. These findings unveiled the changes of tungsten coating properties under extreme conditions that are of both academic and practical values.

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.

A technique based on a flexible endoscope was selected as a tool for the diagnostic of dust in ITER (“The Way” in Latin). The diagnostic will consist of two tools-one for fine viewing of dust with a resolution down to a few tens of microns in a few millimeter spot and another one for dust collection. The endoscope will have to go up to 15-m deep inside the tokamak to the inspection region. Due to the specific design features of ITER, the endoscope will have to go upward on an inclined surface for inspection about 18 m away from the insertion point. In order to ensure that the endoscope gets to the desired region of inspection, it will be pushed through guide tubes having a number of bends along their length. Initial estimations of endoscope jacket materials, endoscope stiffness, and push/pull forces were defined experimentally. This article will give a brief reminder of the overall strategy for dust/erosion/tritium monitoring in ITER that is a Protection Important Activity (PIA) and the role of the dust monitor in this context. It focuses on experimental results of real-size tests inside guide tubes of the behavior of different endoscope emulators under various conditions.

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.