V.M. Gureev

Microstructure and deuterium content of tokamak T-10 carbon erosion products

Abstract An analysis of erosion products deposited inside the T-10 tokamak vacuum chamber revealed the presence of two groups of substances: carbon films and dust microparticles. Specimens of the carbon films were investigated using scanning and transmission electron microscopy. In terms of microstructure, the films appeared to be either smooth (homogeneous) or globular, that is, formed by globe-shaped microparticles. Dust specimens were also collected and analysed. Particles with count median diameter of 10–15 nm were detected. The deuterium concentration in the sampled erosion products was measured using the SIMS/RGA technique.

Selfsputtering of beryllium and sputtering and erosion of CC composite in the experiments on plasma disruption simulation

Abstract The energy dependence of the Be-selfsputtering yield in the energy range of 1.5–10 keV was measured. The yield of Be-sputtering by Be + -ions attains its maximal value at the ion energy of 1.5 keV, being equal to 0.31 ± 0.02 atoms/ion; at the further increase in the energy the yield monotonously decreases. The temperature dependence of the sputtering yield of CC composite with H + ions are given. Results from disruption simulaiton experiments are described in which CC composite specimens were exposed to 0.5–1.0 MJ m −2 energy deposition during 50 μs from hydrogen plasma with a density of ∼ 10 15 cm −3 . The microstructure of redeposited layers, deposits morphology and the deposits size distribution were studied by the methods of transmission and scanning electron microscopy.

Peculiarity of deuterium ions interaction with tungsten surface in the condition Imitating combination of normal operation with plasma disruption in ITER

Tungsten is a candidate material for the ITER divertor. For the simulation of ITER normal operation conditions in combination with plasma disruptions samples of various types of tungsten were exposed to both steady-state and high power pulsed deuterium plasmas. Tungsten samples were first exposed in a steady-state plasma with an ion current density 10 21 m -2 s -1 up to a dose of 10 25 m - 2 at a temperature of 770 K. The energy of deuterium ions was 150 eV. The additional exposure of the samples to 10 pulses of deuterium plasma was performed in the electrodynamical plasma accelerator with an energy flux 0.45 MJ/m 2 per pulse. Samples of four types of tungsten (W-1%La 2 O 3 , W-131. monocrystalline W(1 1 1) and W-10%Re) were investigated. The least destruction of the surface was observed for W(1 1 1). The concentration of retained deuterium in tungsten decreased from 2.5 x 10 19 m - 2 to 1.07 × 10 19 m -2 (for W(1 1 1)) as a result of the additional pulsed plasma irradiation. Investigation of the tungsten erosion products after the high power pulsed plasma shots was also carried out.

Plasma impact on materials damaged by high-energy ions

We present a short review of experimental research carried out at the NRC Kurchatov Institute over recent years on the behavior of plasma-facing materials (PFMs) when a high level of radiation damage in plasma. Neutron-induced damage was modeled with accelerated ions (in the MeV range) and covered a 1–80 dpa interval. Irradiated carbon materials and tungsten were exposed to deuterium steady-state plasma at deuterium ion energies of 100–250 eV. The work focused on the damaging effect on erosion and on deuterium retention in irradiated materials. The influence of displacement damage was found on the erosion of carbon materials after their bombardment with C+ ions. Changes in deuterium retention were observed on tungsten damaged by 3–4 MeV helium ions. The experiments and results show the efficiency of the method for investigating plasma influence on PFMs for fusion applications taking into account the effect of accumulated radiation damage.

Studies of tritium desorption from beryllium and characterization of erosion products under plasma–beryllium interaction

Abstract The paper describes results of two studies related to the use of beryllium as plasma facing material in fusion reactors. The first study is devoted to the burst release of tritium and helium from neutron irradiated beryllium under high heating rates (up to 100 K s−1), and to reduction of temperatures corresponding to the gas release peaks with growth of the heating rates as well as to mechanisms of the above mentioned phenomena. Experiments were accompanied with theoretical estimates. Beryllium sputtering under hydrogen ion bombardment was under the second study. The microstructure of beryllium surfaces and redeposited films were investigated. Thickness of the redeposited film, sizes of erosion products, and hydrogen accumulation in it were measured.

Surface erosion of tungsten and the morphology of erosion products in experiments simulating plasma disruption

The surface erosion of different sorts of tungsten subjected to high-power pulsed plasma streams simulating plasma disruption is studied. With W-13I polycrystalline and (111) single-crystal tungsten samples used as examples, the size distributions for the erosion products collected at different angles to the target are compared. The typical drop erosion of the surface is observed. Fine drops either return to the surface or fly away in a direction parallel to the surface. Coarse drops leave the surface nearly at right angles to the surface. The single-crystal surface displays the absence of fine (<0.125 µm) drops typical of a polycrystalline tungsten surface. The erosion of the single-crystal samples is least among the tungsten sorts considered.

Investigation of the beryllium ion-surface interaction

Abstract The energy and temperature dependence of self-sputtering yields of beryllium were measured. The energy dependence of the beryllium self-sputtering yield agrees well with that calculated by Eckstein et al. Below 770 K the self-sputtering yields are temperature independent; at T irr. > 870 K the yield increases steeply. Beryllium samples were implanted at 370 K with monoenergetic 5 keV hydrogen ions and with a stationary hydrogen plasma power flux of about 5 MW/m 2 . In the fluence range of 5 × 10 22 -1.5 × 10 25 m −2 the depth profile is shifted towards the surface with increasing fluence and the concentration of trapped hydrogen atoms is reduced from 3.3 × 10 21 to 7.4 × 10 20 m −2 . About 95% of the trapped hydrogen is located within bubbles and only ∼ 5% is trapped as atoms. With increasing implantation fluence the bubbles coalesce, producing channels through which hydrogen escapes.

Experimental modelling of plasma–graphite surface interaction in ITER

Abstract The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10 22 cm −2 and subsequent irradiation by power (250 MW/cm 2 ) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770°C and 1150°C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength.

Erosion products of ITER divertor materials under plasma disruption simulation

Abstract Candidate ITER divertor armor materials: carbon–fiber-composite and four tungsten grades/alloys as well as mixed re-deposited W+Be and W+C layers were exposed in electrodynamic plasma accelerator MKT which provided a pulsed deuterium plasma flux simulating plasma disruptions with maximum ion energy of 1–2 keV, an energy density of 300 kJ/m2 per shot and a pulse duration of ∼60 μs. The number of pulses was from 2 to 10. The resultant erosion products were collected on a basalt filter and Si-collectors and studied in terms of morphology and size distribution using both scanning and transmission electron microscopy. Metal erosion products usually occurred in the form of spherical droplets, sometimes flakes. Their size distribution depended on the positioning of the collector. Simultaneously irradiated W, CFC and mixed W+Be targets appeared to have undergone a greater erosion than the same targets irradiated individually. Particles sized from 0.01 to 30 μm were found on collectors and on a molten W-surface. A model of droplet emission and behavior in shielding plasma is provided.

Subthreshold sputtering at high temperatures

The sputtering of tungsten from a target at a temperature of 1470 K during irradiation by 5-eV deuterium ions in a steady-state dense plasma is discovered. The literature values of the threshold for the sputtering of tungsten by deuterium ions are 160–200 eV. The tungsten sputtering coefficient measured by the loss of weight is found to be 1.5×10−4 atom/ion at a deuterium ion energy of 5 eV. Previously, such a sputtering coefficient was usually observed at energies of 250 eV. The sputtering is accompanied by a change in the target surface relief, i.e., by the etching of the grain boundaries and the formation of a wavy structure on the tungsten surface. The subthreshold sputtering at a high temperature is explained by the possible sputtering of adsorbed tungsten atoms that are released from the traps around the interstitial atoms and come to the target surface from the space between the grains. The wavy structure on the surface results from the merging of adsorbed atoms into ordered clusters.