The article presents the results of the first experiments on the T-15MD tokamak. Plasma discharges with current up to 250 kA, discharge duration up to 2 s, and plasma electron temperature up to 3 keV have been obtained in the experiments.

One of the main objectives of the materials science studies on the T-15MD tokamak is to insert samples, including large-scale ones, both for their long-term life cycle tests in the Scrape-Off Layer (SOL) plasma areas and for studying the effect of pulse loads in the divertor section. A conceptual design of a complex for materials science tests on the T-15MD tokamak has been developed at the Department of General Physics and Nuclear Fusion of the National Research University Moscow Power Engineering Institute. The complex will enable unique experiments to be conducted for investigating the state and composition of the divertor and first wall surfaces without the need to disturb the tokamak vessel vacuum or dismantle the components. It will be possible to perform such tests in the time intervals between tokamak pulses, which will reveal changes on the divertor surface after each pulse and allow to estimate correlations between the plasma parameters and changes on the first wall surface. Means are provided in the attachment assembly for cooling mockups with both water flow and dispersed water-air flow. In addition, various dignostics for monitoring the first wall and divertor surfaces can be installed on the attachment assembly.

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.

Geodesic acoustic modes, Alfvén eigenmodes, stationary and oscillating components of the electric field, and their relations to plasma confinement have been studied. Theoretical concepts of the interaction of geodesic acoustic modes with other types of plasma turbulence have been tested. Concomitant problems such as the effect of heating methods on the plasma confinement and turbulence, as well as the properties of impurities and scrape-off layer have also been discussed. Key diagnostics such as heavy ion beam probe, correlation reflectometry, and multipurpose optical diagnostic complex have been described. The main experiments have been performed at the T-10 tokamak, which has a high-power electron cyclotron resonance heating system. The supporting experiments have been performed at the TJ-II stellarator in Spain, the COMPASS tokamak in Czech Republic, and the STOR-M tokamak in Canada. The results of the experiments have been compared with the analytical and numerical calculations.

Studies of samples of graphite heat-shielding facing of the T-15MD tokamak chamber with stationary plasma loads in the PLM plasma device and high-power electron beams simulating loads during ELMs and transient processes were carried out. When the graphite facing samples were tested with plasma loads up to 2 MW/m2, their surface temperature reached more than 1200°C, the surface heating did not lead to cracking, a change in the relief, or significant erosion of the graphite surface. The plasma action led to the growth of a layer of highly porous carbon structures on the surface in the zone of contact with the plasma column. Irradiation of graphite with electron beams with a load of less than 12 MW/m2 did not lead to changes in the surface structure; at a load exceeding 24 MW/m2, the processes of erosion and cracking of the surface along the grain boundaries began. At thermal cyclic loads exceeding 380 MW/m2, effects of significant erosion were observed with a rate of material removal from the graphite surface up to 175 μm/s. The conducted experimental studies and tests are considered as the basis for using the investigated type of graphite as the facing of heat-shielding components of the divertor and the first wall for operation in the T-15MD tokamak.

Plasma regime with a positive core electric potential (positive electric field) was obtained for the first time in a tokamak. This low-collisionality regime was reached by strong electron-cyclotron resonance heating (PEC = 2.2 MW) of the low-density (nе ~1.010 19 m) plasmas in the T-10 circular tokamak (B0=2.2 T, R=1.5 m, a=0.3 m, Ipl = 230 kA). The obtained positive electric field is not consistent with NC approach and suggests turbulence origins. The coupling of plasma potential and collisionality was extended towards the ‘banana’ collisional regime predicted for ITER, so the positive plasma potential is expected for ITER plasmas.

Investigations of samples made of a graphite tile of the lining of the chamber of the T-15MD tokamak with plasma loads simulating the conditions in the peripheral and divertor plasma of a large T-15MD tokamak. When testing samples in a plasma installation PLM with plasma loads of 2 MW/m 2 for more than 200 minutes, the surface temperature of the samples reached 1288 ° C. Plasma loading and heating of the surface within the specified limits does not lead to cracking change in the relief and graphite sputtering. The analysis of the morphology and structure of the surface was carried out using the methods of scanning electron microscopy of samples after thermal loads. Surface erosion is insignificant. Long-term plasma loads lead to the growth of a layer of porous carbon structures. The growth rate of structures does not exceed 1–2 μm/h at the maximum heating temperature of the samples and decreases 10 times at 800 ° C. Investigations of graphite samples in an electron-beam setup were carried out with a beam of accelerated 60 keV electrons, which simulates the effect on the graphite lining of the chamber of transient processes expected in the T-15MD tokamak, including ELMs. It is shown that the energy flux of 24 MW/m 2 for electron irradiation should be recognized as a threshold for the destruction of the surface of graphite tiles.

Surfaces facing the gap between W tiles of the ring limiter of tokamak T-10 were analyzed after T-10 decommissioning using LIBS, SEM/EDA, XRD, TDS, and NRA techniques. Gaps with the width of 5 mm and 0.1 mm were nearly completely covered to their full depths of 22 and 15 mm, respectively, by a deposited film. The film was formed mainly by deposition of lithium that came from Li limiter and transformed in air to Li2CO3 and Li2O. Carbon was deposited from volatile hydrocarbons sputtered from the tokamak walls. Besides, carbon appeared due to chemical reaction with lithium in air. Chemical interactions of W with C, O, and Li led to formation W2C, WC, WO2, and Li2WO4. Carbides formed in W over the entire surface to the full depth of the gaps. Trapping of deuterium and helium in tiles was demonstrated. Possible influence of auto-oscillating discharges on ionization and ion trapping of C,D, and He in gaps is discussed.

An investigation of lithium influx to plasma, radiation, transport, and its influence on plasma parameters is carried out using spectroscopy and transport modeling methods. The lithium source in the discharge of the T-10 tokamak is a moving Li-limiter based on a tungsten capillary-pore system (CPS). A strong dependence is found between the Li influx to plasma on the distance between the Li-limiter and plasma current boundary in the scrape-off layer (SOL) region. It is shown that if the Li-limiter is located on the radius of the chamber wall (r approximate to 40 cm), lithium influx to plasma is at a low level with a value similar to 2 center dot 10(18) s(-1) and the Li nuclei concentration in the plasma does not exceed similar to 0.5% of electron density ne with the effective charge value Z(eff)approximate to 1.2. If the Li-limiter is moved to the current radius (a(L)=30 cm), lithium influx to plasma reaches similar to 3.10(20) s(-1), which leads to a formation of lithium plasma with the density limit of approximate to 0.5 center dot n(e)(Gw) (the Greenwald density limit) with a Z(eff) value of approximate to 2.9 and approximate to 10% of the remaining deuteron density. In this case, the lithium radiation losses power estimated in Li-plasma is extremely low and does not exceed approximate to 15% of ohmic power. All obtained results are interpreted using transport models. Changing the lithization level of the chamber walls can significantly reduce the level of C, O, and W impurities and greatly affect the Z(eff) values. The dependencies of energy confinement times in the electron and ion components on the value of the averaged electron density are investigated in regimes with significantly varying Z(eff) values. The paper presents the computational analysis of the lithium radiation power at the plasma periphery and in the SOL as a function of boundary conditions, transport characteristics, and confinement times of lithium ions.

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.

The paper presents the recent results from the quasi-coherent (QC) mode studies performed in the ohmic plasma of the T-10 tokamak (R = 1.5 m. a = 0.3 m, B0 = 2.2 T, Ipl = 230 kA, and ne ≈ 1⋅1019 m−3) using Heavy Ion Beam Probe diagnostics (HIBP). The radial distribution of the amplitude of the density perturbations caused by the QC-mode has two pronounced peaks (approximately 5 and 1.5 % at radii of 25–30 and 8–14 cm, respectively) with an intermediate zone of lower amplitude in between them (<1% at radii of 16–21 cm). The mean QC-mode frequency remains nearly constant within the studied region (75–80 kHz). For the edge plasma, the HIBP observations were confirmed by the Langmuir probe measurements.

ВЛИЯНИЕ АККУМУЛЯЦИИ ПРИМЕСИ ВОЛЬФРАМА НА УСТОЙЧИВОСТЬ ОМИЧЕСКИХ РАЗРЯДОВ Т-10В.А.Крупин 1 , И

The structure of lithium deposited specimens at the T-10 tokamak vessel after examination of lithium capillary porous system were analysed. Scanning electron microscopy have been used to analyse the deposits. Composites of lithium carbonate LixCOy have been found by analysis. Plasma irradiation of these lithium specimens have been carried out in the PLM plasma device with plasma parameters similar to the tokamak divertor plasma. Stationary plasma load up to 1 MW/m2 during 200 minutes in the PLM provided the change of surface morphology revealed by post-mortem analysis.

The tungsten limiters were tested in the T-10 tokamak during the 2015-2017 experimental campaign. The limiters were made from the ITER-grade WMP "POLEMA" tungsten. Inspection of the tungsten limiters after experimental campaign has revealed their strong destruction due to the combination of the high heat and particle fluxes from the edge plasma toward the limiters surface. The influence of the edge tokamak plasma on tungsten limiters leads to significant cracking of the tungsten. Heat load up to 2 MW m(-2) leads to micro-cracks at the grain boundaries. Heat loads exceeding 5 MW-2 lead to macro-cracks formation. In the areas of highest heat loads, tungsten plates are melted and close to full destruction. In these zones, intensive sparking and arcing were observed. Arc craters have been scattered across the surface, and along the cracks. It is supposed that the nonambipolar flow due to the arcs and sparks can explain the observed overheating and melting of the tungsten surface. Disruptions and runaway electrons beams have driven extreme heat loads of more than 1 GW/m(2) causing strong melting of the tungsten on the outer board of the ring limiter.

Эффекты дугообразования при действии плазмы на вольфрамовые компоненты первой стенки в токамаке

Arc erosion of tungsten plates, which occurred after they were tested at the T-10 tokamak in shots with high-power ECR heating, was observed and analyzed. In the T-10 tokamak, the regimes with nonambipolar energy flow onto the tungsten limiter plates were obtained. In such shots, the tungsten plates at the inner contour of the limiter were heated up to a temperature above 2000 °C. Apparently, the nonambipolar energy flow onto the metal surface, which is caused by the effect of sparks and arcs, can be responsible for the surface overheating. The nonambipolar energy flow appears as a result of the explosive electron emission occurring in the course of spark formation which increases the electron flux from the surface into plasma by an order of magnitude.

The paper introduce the experimental dates of nonambipolar plasma flow, due to arcs and sparks, as mechanism of power exhaust, leading to a very high heat load. The ecton mechanism of electron emission results in the pulsed-periodic ignition of the explosive electron emission (EEE) events providing high enough electron current from the metal surface. Unlike standard thermionic emission, such mechanism can dramatically increase electron emission and, as a result, sparks and arcs activity, leads to a surface overheating and melting. Such phenomenon have been observed in experiments on the T-10 tokamak with ITER-grade tungsten (W) poloidal limiter under a powerful plasma electron cyclotron resonance heating (ECRH) and plasma ring shifted inside. The interior W tiles were heated up to temperature exceeded 2000 0C, estimated local thermal load were 40 MW/m2, leading to surface melting. Analysis shows, that the W surface overheating can explained by nonambipolar flow due to the arcs and especially sparks effects only. The reason of such phenomenon can be plasma-turbulence-driven fluctuations of particle and energy flux to the plasma-modified surface. The report analyzes consequences for ITER the EEE appearance on the divertor W surface the sharpening of SOL power width distribution, parallel to the magnetic field – q; the melting of the W leading edges of divertor targets and the recrystallization of the W surface as a result of the superheated liquid metal droplets appearance. Melt tungsten can be subject to J × B force. EEE can lead to the erosion enhancement of the divertor plates. Micro-explosions lead to droplets, which, like dust particles, can effectively deliver impurities to the central region of the plasma. The paper, on the basis of experimental data, demonstrate that taking into account the nonambipolarity of the particles flow onto the ITER divertor plates can lead to a significant increase in the calculated heat flux to these plates. As examples, experiments [1] and simulation of ITER divertor plate erosion [2] shows that repetitive ELM events increase the probability of arcing ignition with the injection tungsten impurity (neutral vapor, molten and solid droplets). Moreover, arcs can grow and eventually create hot spot. These have dominant thermal or burst-type electron emission with substantial amount of tungsten impurities into the plasma. When modeling the energy and particles balance in ITER, an ambipolar plasma flow usually used on the divertor plates and the first wall [3]. In fact, the parameters of plasma-wall interaction, laid into the models, are not obvious, and the calculated heat flux to divertor plates surfaces is minimal. Taking into account the nonambipolarity of plasma flow will lead to an increase in the heat flux to the divertor plates. The nonambipolar nature of the particle flux in tokamaks occurs at large electric fields cross the SOL, in regimes with the accelerated electrons in SOL, in thermal emission of electrons, as well as in the ignition of arcs. As shown by experiments [4], arcs formation can drastically change the conditions of heat flowing in tokamak. Experimental conditions Strong heating of the inner part of the poloidal graphite limiter was discovered back in 2000 on the T-10 tokamak in discharges with powerful ECR heating and shifting inside the plasma column. This regime led to improved plasma confinement [5]. In this regimes, the inner part of the circular graphite limiter was heated to temperatures exceeding 2000 0C. Estimation of heat flux to the inner part gives value near 50 MW / m2. A feature of this mode was a strong arc erosion of the end surface of the limiter and the formation of deep cracks on the lateral surface. A similar phenomenon was discovered on T-10 many years later, in 2016, after the replacement of the graphite limiter by tungsten limiter [4]. Russian tungsten supplied for ITER used, Again, as then, with the shift of the plasma column inside and the powerful ECRH, big part of energy with conduction and convection flowed to inner part of limiter. The heating of the tiles initially took place slowly, and then the temperature increased rapidly, even “explosively”. Tungsten tiles on the inner part of the circular limiter were heated to a temperature much higher than 2000 0C (Fig. 1). Estimates of the energy flux to this tiles by different methods gave a value no less than 20 40 MW/m2. As shown by post-mortem analysis, this resulted in melting of the limiter tiles edge.

This work is dedicated to simultaneous measurements of plasma potential oscillations at GAM frequencies in different locations on the T-10 tokamak and studying their correlation properties. Recent experiments with Heavy Ion Beam Probing and Langmuir probes have shown high coherency between signals of two diagnostics (up to 0.8) despite a large distance between the observation points: half of torus in toroidal and about pi in poloidal direction, up to 12 cm in radial direction. The phase shift between potentials measured with two diagnostics has been obtained in two plasma scenarios. It was found the most likely that potential oscillations at the GAM frequency propagate outward, but influence of 2 pi phase shift cannot be excluded.

Geodesic Acoustic Modes (GAMs) and broadband (f <400 kHz) turbulence of the plasma electrostatic potential and electron density have been directly studied in OH and ECRH plasmas in the T-10 tokamak. Potential and density fluctuations measurements were carried out with Heavy Ion Beam Probe (HIBP) from the plasma core to the edge. Four different types of modes were observed in T-10 plasmas: GAM (fGAM=20-25 kHz) with higher frequency satellite (fsat=22-28 kHz), Stochastic Low-Frequency Modes (SLF) with fSLF=0-30 kHz, Low-Frequency Quasicoherent Modes (LFQC) with fLFQC=50150 kHz and High Frequency Quasi-Coherent modes (HFQC) with fHFQC=200-400 kHz). Both GAM and satellite have homogeneous radial distribution of the frequency and amplitude of electric potential fluctuations in OH and ECRH plasmas. Cross-correlation analysis supports modes identification, while bicoherence analysis have shown three-wave interaction for both GAM and satellite.