Electron cyclotron (EC) resonance heating in the T-15MD tokamak is calculated using the GENRAY code. The injection schemes without current generation for heating plasma using an extraordinary wave with a frequency of 140 GHz at the third harmonic of the fundamental EC frequency are considered. The cases of the wave injected through vertical, upper inclined, and equatorial port-plugs for three scenarios of the T-15MD operation—the baseline configuration and two configurations with different moderate values of plasma elongation and triangularity—are analyzed. A comparison of calculations performed using the GENRAY code with the published calculations based on the OGRAY code shows that the efficiency of EC heating in single-pass absorption and the spatial position of the absorbed power maximum are in good agreement, while the shape of the spatial profile of the absorbed power can differ significantly.

One of the conditions of safe operation for the experimental tokamak reactor ITER is the possibility of mitigating disruption instability by massive injection of inert gases, in particular, of argon and neon. Here we present the results of assessing the influence of multiplet splitting and line radiation imprisonment during the discharge quenching by intense argon injection in ITER. In this paper, the fine structure of energy levels and the noncoronal collisional-radiative kinetics for the radiating excited state are used. For the radiation of two argon ions, Ar+15 and Ar+3, which have spectral lines of high intensity and could be used for plasma diagnostics, it is shown that the optical thickness for the ionic strongest lines has no significant effect on the total power losses of plasma radiation in the considered quenching scenario (massive argon injection in the 15 MA, Q 10 basic scenario in ITER, carried out at the quasi-stationary stage of the discharge, flat-top of the current). The most significant effect appears to be the multiplet splitting of atomic levels, which provides an increase in the radiative losses, e.g., by a factor of 2 for low-ionized atoms at low temperatures, because the resolution of the fine structure of atomic levels for Δn = 0 transitions leads to a contribution of lower excitation energy than that in the model of multiplet-average radiative transitions.

A simple qualitative model of the influence of grouping of atomic levels on the radiation losses of impurity ions in a near-wall low temperature plasma is proposed using the example of line radiation losses when excited states are populated by impact excitation by plasma electrons. A comparison of the relevant data simulated by the ADAS codes for low-charge ions of argon, neon, and nitrogen ions confirms the predicted effects of possible underestimation or overestimation of radiation losses when the number of atomic levels taken into account in the simulation in the collisional-radiative model decreases.

A new method for estimating the global erosion of beryllium (Be) in ITER is proposed. The method uses ray tracing-aided tomography to reconstruct the three-dimensional (3D) profile of beryllium visible-light emissivity in boundary plasma from images captured with filtered cameras of VIS/IR wide angle viewing system, H-alpha (and Visible) Spectroscopy diagnostics and signals collected with divertor impurity monitor. The light reflected into the detectors from metallic plasma-facing components (PFCs) is filtered out in the process. The reconstructed Be emissivity is then used to assess the Be influx density distribution along all Be PFCs by integrating the product of the emissivity and the S/XB coefficient along the normal to the PFC surface. The accuracy of this method is evaluated by a comparison with synthetic emissivity data produced by recent simulation of global Be erosion and migration in ITER using the ERO2.0 code. The impact of the uncertainty of PFC light reflection properties on the error in reconstructing the 3D Be emissivity profile and Be influx density is analyzed. The method allows to recover with good accuracy the Be influx density in plasma-wetted areas under the conditions of H-mode fusion power operation with high plasma density in far scrape-off layer (SOL). Under the conditions of lower far-SOL plasma density and L-mode operation, only the total Be influx integrated over the area of the first wall panels with relatively high Be erosion can be reconstructed with a high accuracy. It is shown that neglecting the effects of light reflection may lead to a twofold overestimation of the total Be influx.

The effect of thermodynamic nonequilibrium of hydrogen isotope recycling, which consists in significant deviation of the velocity distribution function (VDF) of neutral hydrogen isotope atoms from the Maxwellian VDF, on the passive signal of the charge-exchange recombination spectroscopy (CXRS diagnostics) of edge plasma in tokamaks is analyzed using the ITER tokamak reactor as an example. The calculations of the VDF for atomic and molecular deuterium are carried out using the three-dimensional Monte Carlo code EIRENE as a part of the two-dimensional transport code SOLPS and the semi-analytical Ballistic Model BM1D2V, which is one-dimensional in coordinate and two-dimensional in projections of atomic velocity. The source of level population of the Be IV beryllium ion, the line radiation of which is used in the CXRS diagnostics, is calculated for the above two methods of calculating the VDF and the average atomic kinetic energy, as well as in the case of neglecting the thermodynamic nonequilibrium of recycling, when the temperature of atoms is equal to the temperature of the plasma ions. The calculations of the cross section for the charge-exchange reaction of beryllium ions Be V on deuterium atoms using the ARSENY code and other codes and the erosion kinetics of the beryllium first wall using the three-dimensional code ERO2.0 are used. The calculations of the level populations of the Be IV beryllium ion using the nl-KinRyd code, the luminosity of the ions, and the passive charge-exchange radiation intensity on the observation chord are carried out for the three indicated cases of the VDF of deuterium atoms. A significant dependence of the results of the conducted predictive modeling on the thermodynamic nonequilibrium of recycling is shown.

A simulator of the kinetics of hydrogen isotopes recycling in plasma for the H-alpha diagnostics of the fuel ratio in a tokamak-reactor is proposed. The simulator represents a generalization of the well-known SXB method developed for determining the density of impurity flux from the vacuum vessel first wall into plasma from intensity of the spectral line of an atom or ion integrated with respect to wavelength within the spectral line width, to the case of an isotope mixture. The simulator allows on to determine the parameters of the fuel ratio for a mixture of deuterium and tritium hydrogen isotopes in real time (e.g., within 100 ms, according to the requirements specified for controlling the parameters of hydrogen isotopes in the ITER demonstration tokamak-reactor). The developed approach allows one to determined the flux density of hydrogen isotopes from the first wall into the plasma based on the results obtained by means of the high-resolution spectroscopy of the Balmer series lines without using the molecular spectra of hydrogen that are difficult to interpret. Calculations carried out for typical conditions of the edge plasma in the tokamak-reactors revealed that the flux density and fuel ratio in a certain part of the operational space of the reactor can be reconstructed with an acceptable accuracy. The role of the simulator for more accurate but more time-consuming interpretation of the measurements using the H-alpha diagnostics is discussed.

The development of synthetic diagnostics for passive plasma spectroscopy systems of the T-15MD tokamak is started using the Cherab software library for modeling fusion plasma spectroscopy. The calculation of signals in the visible wavelength range is performed with the Raysect code using the ray tracing technique, taking into account the reflection of light from the surfaces of the first wall specified as a full CAD model. The results of measuring the characteristics of light reflection by samples of the T-15MD graphite first wall before and after plasma exposure are used. The parameters of the model functions of graphite reflectivity are chosen by fitting the measurements. The results of edge plasma modeling by the SOLPS code are used to calculate the intensity of the spectral line Hα and the fraction of reflected light in the observed intensity depending on the geometry of the line of sight. It is shown that, during radial observation of the scrape-off layer plasma, the reflected light fraction in the observed intensity does not exceed 50

It is predicted that in ITER, due to high values of electron temperature and magnetic field strength, electron cyclotron (EC) radiation emitted by plasma will be a significant source (together with external EC radiation injected for auxiliary plasma heating and non-inductive current drive) of additional thermal and electromagnetic loads for microwave and optical diagnostics. The spectral distribution of plasma EC radiation is particularly important to consider in millimeter-wave diagnostics, namely for high- and low-magnetic-field side reflectometry, plasma position reflectometry, and collective Thomson scattering diagnostic, because the transmission lines of these diagnostics yield the transport of EC waves emitted by the plasma. The development of semi-analytical methods used to describe the spectral distribution of plasma-generated EC radiation in tokamaks, starting from the work of S. Tamor, is based on the dominance of multiple reflections of this radiation from the first wall in a toroidal axially symmetric vacuum chamber. Here, we present calculations using the CYNEQ code of the spectral intensity of the EC radiation emerging from the plasma to the first wall and port plugs for five scenarios of ITER operation. This code uses the symmetry-based effect of approximate isotropy and homogeneity of radiation intensity in a substantial part of the phase space and has been successfully tested by comparison with first-principles codes. The energy flux density in the range of 30–200 kW/m2 is predicted for wall reflectance in the range of 0.6–0.95. The possible effect of this radiation on in-vessel components and diagnostics is assessed by calculating the surface density of the energy absorbed by various materials of the ITER first wall.

A statistical analysis is carried out of the similarity of profiles of plasma parameters: electron temperature T e , density n e , and pressure P e , at the stage of quasi-stationary plasma current (the so called flat-top stage) of about 9000 discharges in the JET tokamak, among the discharges from #84458 to #99419 that covers almost the entire database of successful discharges in the JET ITER-like wall machine. For all these parameters, the existence of universal profiles is shown as functions of the normalized minor radius of the plasma column ρ . The discharge universal profiles (DUPs) are obtained by dividing the space-time-dependent profile by its value at the center of the plasma column or by the space-averaged value of this parameter in the region ρ ⩽ ρ max = 0.5–1, and subsequent averaging over time at the flat-top stage of each discharge. The machine universal profiles (MUPs) are obtained by averaging over time of flat-top stage of all discharges. It is shown that for 86% of the discharges, the time-averaged relative root-mean-square deviation of P e profile from the DUP is less than 20% at ρ ⩽ 0.8, and for the MUP, it is less than 17%. A similar picture is observed for the profiles of T e and n e /√ T e . An auxiliary heating, P aux ⩾ 1 MW or 10 MW, was present at, respectively, 44% and 21% of the total duration of the flat-top stage of discharges. It is shown that large, ∼100% in magnitude, jumps of T e at ρ = 0, caused by the switch-on of auxiliary heating, can be described with a ∼20% accuracy by the jumps of the space-averaged temperature if the DUP is used to describe the stationary shape of the temperature spatial profile. The results obtained illustrate the degree of plasma self-organization in tokamaks.

This article is devoted to the life and work of the outstanding theoretical physicist, creator of world-famous scientific schools in radiophysics, plasma physics, and controlled thermonuclear fusion, Academician Mikhail Aleksandrovich Leontovich (1903–1981). His achievements in basic physics contributed to significant advances in applied research in many fields. The uniqueness of the scientific style and civic appearance of M.A. Leontovich, the lessons of his creative biography remain significant today, especially for young scientists, setting an example of selfless service to science.

A sequence of fuel recovery methods was tested in JET, equipped with the ITER-like beryllium main chamber wall and tungsten divertor, to reduce the plasma deuterium concentration to less than 1% in preparation for operation with tritium. This was also a key activity with regard to refining the clean-up strategy to be implemented at the end of the 2nd DT campaign in JET (DTE2) and to assess the tools that are envisaged to mitigate the tritium inventory build-up in ITER. The sequence began with 4 days of main chamber baking at 320 °C, followed by a further 4 days in which Ion Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) were applied with hydrogen fuelling, still at 320 °C, followed by more ICWC while the vessel cooled gradually from 320 °C to 225 °C on the 4th day. While baking alone is very efficient at recovering fuel from the main chamber, the ICWC and GDC sessions at 320 °C still removed slightly higher amounts of fuel than found previously in isotopic changeover experiments at 200 °C in JET. Finally, GDC and ICWC are found to have similar removal efficiency per unit of discharge energy. The baking week with ICWC and GDC was followed by plasma discharges to remove deposited fuel from the divertor. Raising the inner divertor strike point up to the uppermost accessible point allowed local heating of the surfaces to at least 800 °C for the duration of this discharge configuration (typically 18 s), according to infra-red thermography measurements. In laboratory thermal desorption measurements, maintaining this temperature level for several minutes depletes thick co-deposit samples of fuel. The fuel removal by 14 diverted plasma discharges is analysed, of which 9, for 160 s in total, with raised inner strike point. The initial D content in these discharges started at the low value of 3%–5%, due to the preceding baking and conditioning sequence, and reduced further to 1%, depending on the applied configuration, thus meeting the experimental target.

The Monte-Carlo code ERO2.0 was used to simulate steady-state erosion and transport of beryllium (Be) in the ITER main chamber. Various plasma scenarios were tested, including a variation of the main species (hydrogen, deuterium, helium), plasma conditions (density, temperature, flow velocity) and magnetic configurations. The study provides valuable predictions for the Be transport to the divertor, where it is expected to be an important contributor to dust formation and fuel retention due to build-up of co-deposited layers. The Be gross and net erosion rates provided by this study can help identifying first wall regions with potentially critical armour lifetime.

The possibilities are considered of using the Thomson scattering diagnostics of core and edge plasmas in the tokamak with reactor technologies, which is under design. The problems are described that can be solved using the Thomson scattering diagnostics, including the possibility of controlling the plasma current profile. Technical requirements for the diagnostics are formulated. The possibilities are analyzed of its arrangement in the tokamak vacuum chamber. The accuracies are estimated of measuring the electron temperature and density of the plasma created in the tokamak. Particular attention is paid to ensuring the operability of the proposed diagnostics in the reactor regime of the tokamak operation.

A simple, computationally efficient 1D model is suggested for the recycling of atomic and molecular hydrogen between the main chamber first wall (FW) of a tokamak and the peripheral plasma. The model is based on iteratively solving the kinetic equation for the 1D velocity distribution function (VDF) of hydrogen neutral atoms in the projection of velocity on the direction normal to the FW. The model results in a fast-converging routine due to the domination of the contribution to the VDF from long-distance, ballistic flights of the neutral atoms produced by the charge exchange of colder atoms with hotter plasma ions. The proposed modification of the ballistic model (BM) enhances its original version along the following lines: the account of the inelastic reflections of the neutral atoms from the wall; parameterisation of the boundary condition for the VDF of the atoms produced by the reflection of ions from the wall with immediate recombination, without capture in the wall and thermalisation there; extension to a mixture of hydrogen isotopes; elaboration of algorithms for speeding up computation on graphical processing units. The complete set of equations of the BM is published for the first time. The model is verified by comparison with the EIRENE code simulations of the VDF of neutral atoms and molecules in the scrape-off-layer around mid-plane for typical conditions of ITER operation. The applicability of the BM to the synthetic diagnostics developed for the H-alpha high-resolution spectroscopy is demonstrated. That is, the simple 1D model reproduces reasonably well the 1D projection of the data computed with the 3D Monte-Carlo code on a 2D plasma background.

The present study addresses the uncertainties that affect the recently performed predictions of beryllium (Be) erosion and migration in ITER using the Monte-Carlo code ERO2.0. The focus of the study is a D–T baseline discharge with fusion power gain Q=10, scrape-off layer (SOL) input power PSOL=100MW, toroidal plasma current Ip=15MA, and central toroidal field Bt=5.3T. The parameter studies used to investigate uncertainties include variations of the radial extrapolation of plasma parameters in the far-SOL (scan A), the assumptions on impact angle distributions (scan B) and the anomalous transport of eroded Be (scan C). Variations by factors ∼3, ∼18 and ∼2 for scans A, B and C, respectively, are found.

The results of the analysis of the deviation of the force equilibrium for ions from the neoclassical theory prediction, calculated using the direct measurements of the radial electric field, in the view of its possible local and nonlocal correlation with the profiles of electron, Te, and ion, Ti, temperatures in the T-10 tokamak are presented. Local correlations are analyzed by means of the Pearson's correlation. Nonlocal correlations are treated with an inverse problem under the assumption of an integral equation relationship between the deviation and Te and Ti profiles. The discharges with zero, weak and strong auxiliary heating (electron cyclotron resonance heating) are analyzed. It is found that the electrons substantially (not less than ions) contribute to the deviation of the ion equilibrium from the neoclassical theory prediction both in the local and nonlocal models.

МОДЕЛИРОВАНИЕ СИГНАЛОВ ПАССИВНОЙ ПЕРЕЗАРЯДКИ ВОДОРОДОПОДОБНЫХ ИОНОВ БЕРИЛЛИЯ ДЛЯ АКТИВНОЙ СПЕКТРОСКОПИИ В ТОКАМАКЕ ИТЭР