V.G. Stolyarova

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.

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.

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.

Tungsten erosion under simulation of iter divertor operation

Under simulation of ITER gaseous divertor operation in a dense stationary plasma of the LENTA-facility, the tungsten sputtering by 5-eV deuterons at 1470 K has been detected. The sputtering yield was equal to 1.5 × 10−4 atom/ion. Below 1250 K the sputtering was not observed. The mechanism of subthreshold W-sputtering, based on a potential adatomic sputtering by slow deuterons, is proposed. Under plasma disruption simulation the W-droplet diameter distribution in respect to a collector position has been studied. Large droplets, up to 6 µm in diameter, fly away nearly perpendicular to the surface. Small droplets fly off in parallel to the surface or return to it. The minimal erosion is found for tungsten W (1 1 1) and the maximal one for W-10% Re.

Study of the impact of steady-state plasma on radiation-damaged tungsten

Experimental studies of tungsten (as a candidate plasma-facing material for a fusion reactor) whose properties will degrade as a result of its contact with near-wall plasma and irradiation with neutrons are performed. The effect of a high level of radiation damages (1–100 displacements per atom) on deuterium accumulation and erosion caused by tungsten irradiation with deuterium plasma was studied. Radiation damages are obtained as a result of the irradiation of tungsten samples with high-energy ions in an accelerator (He+2, C+3, 4–10 MeV). Then the samples are exposed to steady-state deuterium plasma at the LENTA facility (National Research Centre Kurchatov Institute). The effects of the erosion of tungsten and accumulation of hydrogen isotopes in it are studied. Modification of the surface microstructure and radiation swelling is observed. The helium and deuterium concentrations were measured using the methods of nuclear elastic backscattering and elastic recoil detection analysis. An increased accumulation of deuterium in the damaged layer to a depth of about 5 μm is revealed.

Effect of mechanical deformation on development of helium porosity

The formation of helium bubbles in 18–10 steel and 20–45 nickel alloy implanted by He ions during tension is studied, and helium release from them during high-temperature deformation is analyzed. During helium implantation, an applied tensile stress favors bubble formation and material swelling. Annealing and deformation of the irradiated materials increase the bubble size. Helium bubble migration and accumulation at grain boundaries cause cracking. Bubble migration is caused by a stress gradient. The deformation of the irradiated materials leads to an increase in the release rate of accumulated helium. A model is proposed for the development of helium porosity in a material under stress. A brittle fracture criterion is formulated for such a material.

Effect of carbon on accumulation of deuterium in beryllium irradiated with stationary plasma

Beryllium targets placed in the PLAST beam-plasma discharge facility were irradiated with a flux of stationary deuterium (D) plasma with a deuterium ion energy of 200 eV and plasma flux density of 3 × 1020 m−2 s−1 at temperatures of 370 and 670 K. The irradiation doses varied from 5 × 1021 to 1024 m−2. To heat the target and to ionize impurities near its surface the target was irradiated with an electron beam. The deuterium concentration at the target center exceeds its concentration at the periphery by a factor of more than two under all irradiation conditions. The target center was enriched with carbon up to 16–24 at %, as compared to 4–6 at % at the target periphery. The [D]: [Be] atomic concentration ratios at the target center were equal to 0.054 and 0.036 against 0.024 and 0.016 at the periphery at temperatures of 370 and 670 K, respectively. It has been found that these ratios depend on the concentration of carbon atoms which trapped deuterium atoms.