The results of experiments at the T-10 tokamak using lithium capillary-porous structures are presented. It is shown that lithium sputtering under conditions of graphite diaphragms can significantly reduce deuterium recycling and the level of impurities in the plasma. At the same time, recycling increases significantly five discharges after the start of the day of the experiment, and the effect of reducing the level of impurities persists for 150–300 discharges. The results of using a capillary-porous structure with lithium filling as a movable rail diaphragm in the T-10 configuration with tungsten main diaphragms are presented. The introduction of a lithium diaphragm into the SOL region makes it possible to reduce recycling and obtain discharges with an effective plasma charge approaching unity. In this case, the effect increases as the lithium sputtered in the chamber is accumulated. It is shown experimentally that a capillary-porous structure with lithium filling can be used as a main diaphragm with longitudinal plasma heat fluxes up to 3.6 MW/m2. However, a necessary condition is the complete impregnation of the porous structure with lithium and the prevention of extrusion of lithium into the discharge as a result of the interaction of the current flowing to the diaphragm with the toroidal magnetic field. Experiments have shown that to obtain discharges with a small lithium admixture, a strong gas injection of deuterium or impurity is required to reduce the temperature of the plasma periphery and effective cooling of the diaphragm below 450°C. Otherwise, the diaphragm transfers into a strong evaporation mode with high lithium flows, which lead to a significant increase in the lithium concentration in the plasma. Strong evaporation reduces the heat inflow and stabilizes the diaphragm temperature.

The KINX and VENUS codes were used for simulation of the baseline inductive and steady-state scenarios of the ITER tokamak operation. The perturbations of plasma electron density and magnetic field caused by the Alfvén modes were calculated in the flux coordinates for these scenarios. The perturbation fields obtained were converted into the engineering coordinates in order to calculate the propagation of probe electromagnetic radiation of the reflectometer using the two-dimensional full-wave TAMIC RτX code in the expected geometry of the experiment. The calculations performed show that for the baseline inductive scenario, in the case of reflection of the extraordinary wave at the lower cutoff frequency from the high magnetic field side, the electric field relative perturbations of the reflected reflectometer signal correspond to the margin of linear range of the diagnostics operation or even go out of this range. It was found that in a number of scenarios, not only the electron density perturbations, but also the magnetic field perturbations significantly contribute to the total signal perturbations that makes even more difficult the further data interpretation. Another possible problem is the narrow frequency range of probing frequencies where the Alfvén mode can be observed. In addition to simulating the reflection of electromagnetic waves from plasma, it was analyzed also the possibility of measuring the Alfvén modes parameters when the extraordinary wave pass through the plasma in the transparency window between the upper and lower cutoff frequencies of the extraordinary wave (refractometry). It is shown that at the fundamental frequency, the phase perturbations range from 3 to 60 degrees, which makes it impossible to use the amplitude-modulated refractometer for analyzing signals. The “synthetic diagnostics” approach was used, which showed itself well for simulating the operation of reflectometers at plasma facilities.

Applications of the two-dimensional full-wave electromagnetic code Tamic Analyzer, developed by B.V. Sestroretskii to study the limiting possibilities of diagnostics of reflectometry in tokamak plasma. It is shown that by comparing experimental reflectometry data with calculations using this code, it is possible to determine the structure of density fluctuations in a tokamak, but the experimental radial correlation lengths can be significantly overestimated. Simulation using the Taimic Anylyser code made it possible to determine the limiting level of density fluctuations at which it is possible to measure the plasma density profile and observe Alfven fluctuations in the International Experimental Thermonuclear Reactor of the tokamak type (ITER) under construction. Possible ways to increase the speed and counting field of the code are considered.

Schemes of the application of refractometry for measuring the average plasma density at TRT (Tokamak with Reactor Technologies) facility, which is currently being designed, are considered and analyzed. Various schemes for the implementation of refractometry at TRT are proposed, taking into account the currently known plasma parameters in various scenarios of the facility operation, and their operation under TRT conditions is analyzed. The effect of different effects on the accuracy of determining the integral plasma density using refractometry in TRT is studied. On the basis of the analysis performed, the number of frequency channels used in refractometry is optimized, a variant of placement of refractometry diagnostic elements on the facility is proposed, the parameters of the diagnostic system are estimated: the accuracy of determining the integral plasma density, time resolution, minimum and maximum measurable plasma densities, and the possibility of using the diagnostics system online is studied.

Reflectometry is considered one of the promising diagnostics for measuring plasma parameters at fusion devices now. This diagnostic has a good compatibility with high neutron fluxes, it is tolerant to mechanical and thermal loads during the plasma regime and has a low sensitivity to dust. Reflectometry is widely used at existing installations to measure the electron density profile and the parameters of electron density fluctuations. In this paper, an analysis is made of the capabilities of this diagnostics at the recently put into operation T-15MD tokamak from the viewpoint of its physical ability to perform measurements, the required parameters of reflectometers, and its technical implementability, taking into account the today state of microwave technology and the experience of operating similar diagnostics in Russia and worldwide. In the development of the diagnostics, it is proposed to widely use the accrued experience collected during the development of the “ITER Reflectometry from the High Magnetic Field Side” diagnostics. The enumerated approaches used can also be applied to the development of reflectometers at other plasma installations.

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.

Reflectometry is currently considered as one of the promising diagnostics for measuring parameters of plasma in thermonuclear devices. This diagnostics is characterized by good compatibility with high neutron fluxes. It withstands mechanical, thermal, and electromagnetic loads during the plasma regime. In addition, it has low sensitivity to dust. Reflectometry is widely used at existing installations for measuring electron density profile and parameters of density fluctuations. Here, we analyze the possibilities of reflectometry at the TRT device that is currently under construction from the point of view of physical possibility of performing the measurements, parameters of required reflectometers, and technical feasibility of the diagnostics taking into account the modern level of development of microwave technology and reflectometric diagnostics in Russia and abroad. It is suggested to incorporate of experience gained upon design of the similar diagnostics for the ITER tokamak during the development of the reflectometric diagnostic for the TRT facility.

The dependence was studied of the level of density fluctuations recorded at different devices on the minor and major radii of the tokamak. It is known that a high level of density fluctuations can negatively affect the operation of diagnostics, in particular, the plasma reflectometry diagnostic. The increase of density fluctuations decreases the quality of raw data by increasing the error of measuring the density profile and, when density fluctuations exceed the threshold value, it renders such measurements impossible. Based on experimental data obtained on devices with substantially different sizes, a dependence of the density fluctuation level on the major and minor tokamak radii was proposed. Since the main experiments were carried out in round limiter tokamaks in ohmic (OH) heating regimes, the extrapolation result is applicable, generally speaking, to installations of larger size with the same configuration and regimes. However, experiments with electron cyclotron heating at the T-10 tokamak also allow one to extend the obtained dependence to regimes with auxiliary heating. It was shown that the obtained dependence is applicable to limiter tokamaks Tore Supra and TFTR of larger size. The applicability of the dependence to installations with diverter configuration is discussed and the possible level of density fluctuations is extrapolated to the international reactor tokamak ITER that is being constructed in France.

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.

The concept of plasma facing elements (PFE) of a stationary fusion tokamak reactor based on the use of stagnant or slow-flowing liquid lithium enclosed in a capillary-porous structure which allows integrating the advantages of liquid metal with the possibility of its uniform distribution over the surface and ensuring a high stability under the influence of magnetohydrodynamic forces is considered in this study. Heat removal is provided by thermal conductivity through the PFE structure to the coolant. Development of the PFE design includes the creation and testing of a series of models of experimental devices such as a limiter and a divertor target with external systems of lithium refilling/change and heat removal designed for research in modeling conditions of bench installations and the T-11M tokamak. The paper presents the design and main parameters of such devices.

The introduction of lithium as a material for production of in-vessel plasma-facing elements of the tokamak necessitates the development of the corresponding diagnostic instruments. A series of diagnostic devices have been developed and fabricated for tokamaks T-10 and T-11 M which make it possible to investigate the processes of lithium transport in the tokamak plasma scrape-off layer (SOL), real-time dynamics of lithium deposition at various temperatures of a collecting surface by means of the microbalance technique, adsorption/desorption process of hydrogen isotopes on the lithium surface, and influence of an electric field on lithium trapping. Scanning of plasma parameters is provided by Langmuir probes. Such devices can be used to extract the lithium deposited on the inner walls of the tokamak vacuum chamber without opening it. For these purposes, a lock chamber and bellows-free vacuum input allowing movement and rotation is provided. It is planned to perform the study of the tokamak plasma interaction with in-vessel lithium-based elements by means of infrared (IR) thermometry. It is planned to try out this technique on the Т-11М tokamak using a special device based on a IR camera.

The specific power deposited in plasma under on-axis electron-cyclotron resonance heating (ECRH) is characterized by strong peaking, resulting in large-amplitude sawtooth oscillations. To analyze the global properties of plasma, instantaneous experimental profiles of the electron temperature should be averaged over time and space. In the present work, a modified canonical profile transport model for predictive calculation of such averaged profiles is proposed. As an example, the model is used to determine the parameters of plasma with ECRH in the T-15MD tokamak currently under construction.

The paper is focused on the new systematic measurements of the 3D spatial distributions of the amplitudes, the radial correlation lengths and the long-range correlations along the magnetic field lines for the different turbulence types. The density fluctuations were measured by the heterodyne correlation reflectometry (CR) using the plasma probing with ordinary mode. CR data was supported by recent experiments with the measurements of the perturbation properties using an heavy ion beam probe (HIBP). The new reflectometer antenna array in T-10 tokamak consists of sets of horns distributed at four places toroidally and poloidally over the torus. The experiments confirmed previously found strong poloidal asymmetry of the amplitude for the broadband (BB) and quasi-coherent (QC) fluctuations. It was found that amplitude of density fluctuations is uniform poloidally for the stochastic low frequency (SLF) fluctuations. The radial correlation were measured at four poloidal angles to reveal the poloidal dependence of the radial correlation length for the different fluctuation types. The significant decrease of the radial correlation lengths towards the high magnetic field side was observed for the QC and the SLF fluctuations. The long-range correlations along the field lines were measured by the reflectometers in two cross-sections separated by 1/4 of the torus. The reflectometers had the same probing frequency thus provide reflection from the same magnetic surface. The measurements were carried out at the low and the high field sides with two currents and two magnetic configurations with simultaneous reversal of the toroidal field and plasma current. The positions of the resonance radius were calculated also using 3D tracing of the magnetic field line and demonstrated good agreement with experiment ones. These results allow to propose the new approach for the current profile measurements in tokamaks.

A new, previously unused combination of the high magnetic field side probing and the lower extraordinary mode cutoff was used in the T-10 tokamak for the electron density profile measurements by using a frequency modulation continuous wave reflectometer. This scheme has a significant advantage for large fusion machines where flat electron density profiles and high electron temperatures are expected. The reflectometer design is based on a common scheme with voltage controlled oscillators, active multipliers, and in-phase/quadrature signal detection. Iterative procedures were applied to calibrate a microwave source frequency and compensate the dispersion of probing signals in the conventional rectangular waveguides used as the transmission lines. About 0.2% stability of the beating frequency was reached after the calibration procedure had been finished. An advanced finite-difference scheme was proposed for the reconstruction of the electron density profile to reduce errors and improve the overall performance of the algorithm. Reconstructed electron density profiles demonstrate good agreement with conventional T-10 diagnostic data. Some specific technical and physical aspects of the diagnostic were also discussed on the basis of T-10 operation experience.

The use of lithium as a material of the tokamak in-vessel plasma-facing components made it necessary to develop appropriate diagnostic instruments. For the T-10 and T-11M tokamaks, devices have been developed that allow one to investigate the processes of lithium transport in the tokamak scrape-off layer, the dynamics of lithium deposition at different temperatures of the collecting surface in real time by using a piezoelectric quartz detector, adsorption and desorption of the plasma-forming gas by lithium, and the influence of the electric field on the process of lithium collection. The plasma parameters are monitored using Langmuir probes. The developed devices can be used to extract lithium deposited on the tokamak vessel wall without breaking vacuum conditions. For these purposes, a gateway and a vacuum input without bellows have been designed on the basis of an innovative liquid-metal coupling.

In ohmically heated (OH) plasma with low recycling, an improved particle confinement (IPC) mode is established during gas puffing. However, after gas puffing is switched off, this mode is retained only for about 100 ms, after which an abrupt phase transition into the low particle confinement (LPC) mode occurs in the entire plasma cross section. During such a transition, energy transport due to heat conduction does not change. The phase transition in OH plasma is similar to the effect of density pump-out from the plasma core, which occurs after electron cyclotron heating (ECH) is switched on. Analysis of the measured plasma pressure profiles in the T-10 tokamak shows that, after gas puffing in the OH mode is switched off, the plasma pressure profile in the IPC stage becomes more peaked and, after the peakedness exceeds a certain critical value, the IPC-LPC transition occurs. Similar processes are also observed during ECH. If the pressure profile is insufficiently peaked during ECH, then the density pump-out effect comes into play only after the critical peakedness of the pressure profile is reached. In the plasma core, the density and pressure profiles are close to the corresponding canonical profiles. This allows one to derive an expression for the particle flux within the canonical profile model and formulate a criterion for the IPC-LPC transition. The time evolution of the plasma density profile during phase transitions was simulated for a number of T-10 shots with ECH and high recycling. The particle transport coefficients in the IPC and LPC phases, as well as the dependences of these coefficients on the ECH power, are determined.