Персона: Писарев, Александр Александрович
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INTERACTION OF LITHUM-DEUTERIUM FILMS WITH ATMOSPHERIC GASES
A choice of plasma-facing materials is one of the key issues in thermonuclear fusion reactor design. Lithium as an element with the low atomic number is a promising material for plasma-facing components (PFC) in fusion installations and a number of experiments at tokamaks already demonstrated many positive effects on plasma operation [1,2]. Lithium can be used for conditioning, PFC on the base of of capillary porous system with liquid lithium are also considered. In any way, one can expect lithium co-deposition with hydrogen isotopes at the surface of PFC and at remote areas. Lithium as a good getter can accumulate high amount of hydrogen isotopes (deuterium, tritium) that can be a problem from the safety reason. Deuterium retention and thermal desorption from lithium films formed in plasma discharge were investigated in this work.
PLASMA NITRIDING OF TITANIUM ALLOY TI5A14V2MO
However, low hardness and low wear resistance of these materials is one of the reasons limiting their wider use. Nitriding of titanium alloys with the purpose of improvement of their wear resistance is an important task. Many works were devoted to nitriding of titanium and low alloyed titanium [1-4], but not much is known about nitriding of highly alloyed titanium, which is a promising material for many applications The superalloy Ti5Al4V2Mo has high strength and corrosion resistance, and it is widely used in industry at medium and high temperatures. Nevertheless, it was not studied in terms of plasma nitriding. This work is devoted to investigation of nitriding of Ti5Al4V2Mo in argon-nitrogen plasma of abnormal glow discharge.
HELIUM THERMAL DESORPTION FROM TUNGSTEN AFTER ION BEAM IRRADIATION AT ELEVATED TEMPERATURES
Helium (He) is a product of deuterium-tritium reaction, so appearance of helium impurities will be unavoidable. In addition to He implantation from fusion plasma, He can be introduced into material by both neutron irradiation and tritium radioactive decay. Presence of He in plasma-facing materials may significantly influence their mechanical properties and surface morphology [1, 2], as well as hydrogen isotope recycling [3, 4]. Tungsten (W) will be used as a plasma-facing material in ITER divertor [5], and it is considered also for application in future fusion devices. Therefore, investigation of He interaction with W is of great interest.
The dependence of lithium emissivity from temperature in vacuum ЗАВИСИМОСТЬ СТЕПЕНИ ЧЕРНОТЫ ЛИТИЯ ОТ ТЕМПЕРАТУРЫ В ВАКУУМЕ
© 2019 National Research Center Kurchatov Institute. All rights reserved.Correct interpretation of IR video observation data of the surfaces of plasma-facing elements in fusion devices requires detailed knowledge about the emissivity factor of these surfaces in different conditions. In this work, results of emissivity measurements for free metallic lithium surface and a lithium surface supported by the capillary-porous system (CPS) are measured as a function of temperature in the range from 400 to 800 K. Emissivity of solid lithium changed from ~0.04 at 400 K to ~0.09 at 453 K. During melting a sudden drop of emissivity down to ~0.04 was observed. Emissivity increased linearly from 0.04 to ~0.15 with temperature increasing from 455 to 800 K. For fully wetted CPS, emissivity was close to that of free lithium surface for temperature up to ~570 K, while at higher temperature it was lower, probably due to changes in microrelief at high temperatures.
PULSED ABNORMAL GLOW DISCHARGE WITH HOLLOW CATHODE FOR NITRIDING OF INTERNAL CYLINDRICAL SURFACES
Plasma nitriding is a method of thermochemical treatment, which is widely used to enhance surface hardness, fatigue strength, and corrosion resistance of steels [1,2]. In industry, glow discharge is commonly used for this purpose. A specific task of plasma technologies is treating of cavities, which are usually inaccessible for plasma. Still, the need to modify cavities and tubes of small diameters and large aspect ratios exists. Abnormal glow discharge with hollow cathode in the pulse-periodic mode was proposed for this purpose in [3]. This work is devoted to investigation of ignition and properties of the abnormal glow discharge in tubes and demonstration of nitriding of internal surfaces of the tubes.
ALUMINIUM OXIDATION IN PLASMA OF ABNORMAL GLOW DISCHARGE
Aluminum oxide layers in all its various forms have become widespread due to the unique combination of properties that can be modified by varying the conditions of their growth. The alumina layer can either be applied to the substrate by chemical and physical methods or grown by oxidation. For practical purposes, layers of various thicknesses and structures are needed. The standard method for obtaining thick (of the order of 100-1000 nm) porous layer is electric arc anodization in weak electrolytes. The standard method for obtaining thin (about 10 nm) layers is thermal oxidation. Dense layers of intermediate thickness are difficult to obtain by such methods. There are few papers in the literature that investigate the possibility of obtaining oxide layers with a thickness of tens of nanometers by oxidizing aluminum in oxygen plasma [1-5]. The description of the kinetics of the oxidation process in those works was given on the basis of the assumption of the diffusion character of oxygen transfer from the surface into the interior of the metal. This approach is absolutely unsuitable for description the transport of oxygen and aluminum through the oxide layer, since both oxygen and aluminum in the oxide are in the form of ions, and an electric charge is formed on the surface of the dielectric facing the plasma, so that the transport of oxygen coming from the plasma must occur under by the action of an electric field in the oxide dielectric layer. In this work, experiments on plasma enhanced oxidation (PEO) were carried out on the oxidation of aluminum in the anomalous glow discharge oxygen plasma, which provides uniform treatment over the entire surface of samples of arbitrary shape. Also, a simple model was proposed for description of the oxidation kinetics, taking into account oxygen transport in the electric field.
HYDROGEN CO-DEPOSITION WITH METALS IN PLASMA DISCHARGE
Deposition of a single element film is always accompanied by co-deposition of a certain amount of other elements. This can be done properly to improve properties of the coating or due to contamination by impurities. In the field of thermonuclear fusion research, where hydrogen isotopes are used as a fuel, co-deposition with sputtered material from the wall is one of major mechanisms of hydrogen isotopes accumulation in the installation. Since D-T fuel will be used in ITER and future fusion reactors, accumulation of radioactive tritium will limit the lifespan of the installations due to safety concerns. For example, tritium accumulation in ITER is limited by 1 kg. This is why carbon materials were not accepted for the use in ITER. Basing on experiments, it was predicted that the safety limit could be reached after 100 of shots with tritium. Recent experiments in JET [1] demonstrated in the case of “ITER-like” wall (first wall – Be, divertor area - tungsten) accumulation of deuterium fuel in the co-deposits was 20 times lower than in the full-carbon wall campaign. This is both due to smaller amount of co-deposits and smaller concentration of deuterium in them.
Substitution of heavy hydrogen isotopes in tungsten layers during gas exposure ЗАМЕЩЕНИЕ ТЯЖЕЛЫХ ИЗОТОПОВ ВОДОРОДА В ВОЛЬФРАМОВЫХ СЛОЯХ ПРИ ВЫДЕРЖКЕ В ГАЗЕ
© 2020 National Research Center Kurchatov Institute. All rights reserved.Efficiency of deuterium removal from tungsten co-deposited layers was studied by means of thermal desorption spectroscopy in the 30 to 200 °С temperature range in vacuum and hydrogen atmosphere (8 104 Pa). Tungsten layers 100 nm and 500 nm thick were deposited in magnetron discharge in argon-deuterium environment and contained ~2% at. of deuterium. Presence of hydrogen (protium) during sample degassing increases the rate of deuterium removal. Simultaneously, an additional amount of protium is captured in the co-deposited layers. The rate of deuterium removal increases with temperature. Exposure at a temperature of 473 K in a hydrogen atmosphere for 18 hours allowed 99% of the captured deuterium to be removed.
Deuterium trapping in co-deposited layers of ITER-relevant materials
Hydrogen isotope accumulation in fusion devices is a serious problem. Because deuterium-tritium mixture will be a working gas in future fusion devices, including ITER tokamak, tritium accumulation is an issue from the perspective of radiation safety. In total, only 700 grams of tritium are allowed to be present in ITER vessel at any time, with additional 120 in the cryopumps, and 180 grams allocated to measurement error, to the total of 1000 grams.
DEUTERIUM RE-EMISSION AND THERMAL DESORPTION FROM IRON AND EUROFER
Reduced-activation ferritic-marthensitic (RAFM) steels, such as Eurofer, are considered as candidates for structural materials in fusion reactors due to the high thermal conductivity, the low thermal expansion coefficient and good resistance to radiation swelling. There are also some concepts of fusion reactors, where RAFM steels also considered as material for plasma-facing components. In this regard, the key aspects of hydrogen (H) isotopes interaction with RAFM steels, such as tritium (T) retention and migration in these materials are particularly important as a point of safety concern.