Formation of a high-field-side high-density (HFSHD) regime and the role of the high-field-side (HFS) poloidal electric field in the scrape-off layer of the spherical tokamak Globus-M2 are analyzed using SOLPS-ITER edge plasma simulations. The dependence of the HFS poloidal electric field sign and, consequently, radial drift fluxes on the discharge density is discussed. It is demonstrated that the HFS poloidal electric field is the key element in the formation of a HFSHD regime in the Globus-M2 tokamak as in ASDEX Upgrade. It is demonstrated that the physics of HFSHD formation in a small spherical tokamak is similar to that suggested by Kaveeva et al (2009 36th EPS Conf. on Plasma Physics) and is in line with experimental observations and modeling, performed later on ASDEX Upgrade.

SOLPS-ITER modeling of EU-DEMO tokamak burning plasma with Ar seeding was performed. The modeling includes drifts, kinetic neutrals and current description switched on. The simulation results are compared with those without drifts. The power entering the edge plasma domain is 200 MW. The deuterium puff is 1.5 × 1023 at s−1 for all cases, which corresponds to neutral deuterium pressure of 10 Pa in the private flux region. The Ar seeding rates are 8.0×1019  at s−1 and 1.5×1020  at s−1, which correspond to the Ar concentration on the separatrix in the range of 0.5%–2%. It is demonstrated that with such a combination of parameters, it is possible to achieve power loads lower than 5 MW m−2 on both divertor targets. The temperature above 5 eV in the far scrape-off layer (SOL) of the outer target remains an issue. Together with drift and no drift cases, the impurity accumulation mechanism in the high-field-side SOL is discussed.

At the compact spherical Globus-M2 tokamak, a series of experiments was conducted to study the effect of the injection of nitrogen on the discharge parameters. The experiments were carried out in discharges in deuterium in the divertor configuration, and the auxiliary heating was performed by deuterium neutral beam injection. During the nitrogen seeding, a substantial decrease in electron temperature near the divertor was recorded as well as a sharp decrease of the heat flux onto the divertor plate, while the density and temperature of the main plasma changed insignificantly. Simulations by the SOLPS-ITER showed a satisfactory agreement with the experiment.

First experiment with nitrogen seeding has been performed at the compact spherical tokamak Globus-M2. Significant reduction of the electron temperature and the energy flux to the outer lower divertor target has been observed experimentally and reproduced in the modeling with the SOLPS-ITER code.

At the compact spherical Globus-M2 tokamak, a series of experiments was conducted to study theeffect of the injection of nitrogen on the discharge parameters. The experiments were carried out in dischargesin deuterium in the divertor configuration, and the auxiliary heating was performed by deuteriumneutral beam injection. During the nitrogen seeding, a substantial decrease in electron temperature near thedivertor was recorded as well as a sharp decrease of the heat flux onto the divertor plate, while the density andtemperature of the main plasma changed insignificantly. Simulations by the SOLPS-ITER showed a satisfactoryagreement with the experiment.

The diagnostic of the peripheral plasma parameters in terms of the relation of lines of neutral helium is included in the diagnostic complex of the tokamak Globus-M2. The first measurements of the peripheral plasma parameters are performed near the lower X point. The measured spatial distributions of the electron temperature and density are in satisfactory agreement with the simulation with the SOLPS-ITER code.

It was shown that the assumption of straight-line probing beams, which is usually used in reconstructing the plasma density profile from interferometric measurements, leads to unacceptably large errors in the case of a dense plasma. A fast method based on the assumptions of geometric optics is proposed for calculating the interferometric signal. This method takes into account the refraction of the probe wave and can be used in reconstruction of the plasma density profile. The method correctness is confirmed by comparison with the results of full-wave calculations.

Current structure in the scrape-off layer (SOL) of a tokamak is analyzed. It is demonstrated that poloidal currents measured in the experiments are a combination of several current types with different physical nature. Besides the known Pfirsch–Schlüter currents and thermoelectric currents, so-called plate closing currents flowing to/from the divertor plates are also analyzed. The latter close radial currents in the SOL and below/above the X -point in the SOL and private flux region (PFR). In particular, the current flowing to the outer plate in the PFR, opposite to the thermoelectric current, is predicted for the standard single-null configuration and favorable direction of ∇ B drift. In addition, a pair of currents should flow to and away from the outer plate. In the single-null configuration, they are often masked by a larger thermoelectric current. However, for the connected double null (CDN) case, where the thermoelectric current is strongly reduced due to smaller temperature asymmetry, these currents dominate. The suggested physical model is supported by the results of simulations performed with the SOLPS-ITER transport code. Simulations were done for ASDEX Upgrade (AUG), L- and H-modes, single-null configurations, and for Globus-M H-mode, both disconnected and CDN configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found. The role of parallel currents in the formation of the potential maximum/minimum in the vicinity of the X -point for strongly detached regimes is also analyzed.

Массив из десяти встроенных диверторных зондов был установлен на нижний купол сферического токамака Глобус-М. С его помощью проведены измерения профилей плавающего потенциала, электронной температуры, ионного тока насыщения и плотности потока тепла на дивертор. Продемонстрировано резкое изменение плавающих потенциалов зондов, находящихся вблизи внешней ветви сепаратрисы, при развитии периферийной неустойчивости. Ключевые слова: токамак, дивертор, пристеночная плазма, ленгмюровские зонды.

In plasma edge simulations using the SOLPS-ITER code, the simulated Scrape-Off Layer plasma domain has historically been restricted to magnetic flux surfaces contacting divertor targets at both ends. We present here a newly developed numerical solver for the B2.5 plasma solver in SOLPS-ITER, allowing the numerical grid to be extended to the true vessel boundaries. The new, unstructured Finite Volume scheme can deal with arbitrary grids and magnetic topologies in the 2D poloidal plane. It includes a correct numerical treatment of possibly misaligned faces and cells w.r.t. the magnetic field to cope with, for example, strong divertor target shaping. The solver combines the benefits of an accurate numerical separation of fast parallel and slow radial transport, with a realistic description of the wall geometry, and the possibility of local grid refinement to capture sharp features in the Scrape-Off Layer flows. Generalized sheath boundary conditions are presented that can be imposed at all vessel boundaries, removing an important modeling uncertainty related to the specification of ad hoc decay length boundary conditions at the outer flux surfaces. The resulting model is applied to an AUG single-null case, a standard benchmark case for SOLPS-ITER. We analyze in particular the impact of the extended plasma model on upstream and divertor plasma conditions, and the improved predictions of heat and particle loads to the main chamber wall. The extended solver also allows for a much improved qualitative agreement between fluid and kinetic neutral simulations, because the fluid neutral solution, which is obtained on the plasma grid, now also extends to the true main chamber and divertor vessel boundaries.

An array of ten built-in divertor probes was installed on the lower dome of the Globus-M spherical tokamak. It was used to measure the profiles of the floating potential, electron temperature, ion saturation current, and heat flux density to the divertor. It is demonstrated that there is a sharp change in the floating potentials of the probes located near the outer separatrix leg at the development of the peripheral instability.

Two Ohmic H-mode discharges of the tokamak Globus-M with similar parameters and different magnetic configurations were simulated with SOLPS-ITER code. One of the discharges has two separatrices and an active low X-point—the disconnected double null (DDN) configuration. The second discharge has a connected double null configuration (CDN). The modelled plasma parameters were matched to the experimentally measured values. The scrape-off layer (SOL) width and energy loads on the divertor targets were analysed for both cases. It is demonstrated that for the Globus-M CDN configuration discharge the energy peak load at the lower outer divertor plate is approximately half of that in the case of the DDN configuration. It is also shown that the electric currents flowing from the plasma to the target plates significantly change the energy flux to the plates in both cases with respect to the flux that would be expected for the floating potential at the plates. Two types of electric currents flowing to the plates were observed. The first is a well-known thermoelectric current flowing between magnetically connected plates with different electron temperatures. This current is more pronounced in the DDN case and is strongly reduced in the CDN case, where the difference between the temperatures at the lower and upper plates is smaller. Additionally, a new second type of electric current to the plates is observed which could be called the plate closing current (PCC). These currents close radial currents in the SOL and private region. For the DDN case the PCC is comparable with the thermoelectric currents, while for CDN case the PCC is dominant.

In according to a present understanding of Scrape-Off Layer (SOL) physics, future thermonuclear devices like ITER, DEMO and beyond, require high radiation regimes in order to reduce heat loads on tokamak divertor. Recent experiments at ASDEX Upgrade, JET and other tokamaks demonstrated that such regimes might be achieved by the seeding of the radiative impurities. In the present paper the modeling of the high radiation regimes and the transition to the detachment at the Globus-M2 spherical tokamak is performed by the SOLPS-ITER transport code. The obtained modeling results for GLobus-M2 tokamak demonstrate the trend similar to what is observed at larger machines, e.g. AUG and JET. The significant reduction of peak power density at the outer target plate and transition to the detachment with High Field Side High Density (HFSHD) formation at the inner plate was achieved with impurity seeding rate almost equal to the deuterium puff (in el/sec). However, unlike AUG, further increasing of the seeding rate leads not to a formation of the radiative X-point, but to a radiative collapse. This is caused by smaller machine size, which allows the impurity neutrals to penetrate easier into the confined region. It was noticed that starting with attached divertor the inner target transits to the detachment earlier than the outer one.

The reduction of heat fluxes to the tokamak divertor targets is a crucial problem of future thermonuclear devices, such as ITER and DEMO. According to the present understanding of the Scrape-Off Layer physics, such devices should operate in a detached divertor regime, when most of the exhaust power coming to a divertor region is dissipated by radiation. Recent experiments using ASDEX Upgrade (AUG), JET, and other tokamaks demonstrated that the transition to the detachment may be achieved by the intensive puff of radiative impurities. Spherical tokamaks can give a contribution to the understanding of mechanisms, which defines the impurity circulation in the tokamak volume. A fusion neutron source for a hybrid fusion–fission reactor is considered to be based on a spherical tokamak, and in a steady state, it can face the problem of critical heat loads. Simple estimates of power fluxes to the divertor of the Globus-M2 tokamak (which is an upgraded Globus-M tokamak) result in that they will exceed the limit of 10 MW/m2 at both inner and outer divertor targets, so the impurity seeding might be required. In the present paper, the modeling of different regimes of the Globus-M2 tokamak is performed by the SOLPS-ITER code with varying nitrogen seeding rates. It is demonstrated that with a seeding rate almost equal to the deuterium puff (as measured in electrons/s), a significant reduction of the peak power density at the outer target plate may be achieved, while the inner target plate goes to a detachment with a formation of High Field Side High Density. This result is similar to what is observed in the experiments using ASDEX Upgrade. However, in contrast to AUG, further increasing the seeding rate leads to a radiative collapse rather than to a formation of the radiative spot near the X-point. This is caused by a smaller machine size, which allows the impurity neutrals to penetrate easier into the confined region.

The structure of the scrape-off layer (SOL) of a tokamak with little or no turbulent transport is analyzed. The analytical estimates of the density and electron temperature fall-off lengths of the SOL are put forward. It is demonstrated that the SOL width could be of the order of the ion poloidal gyroradius, as suggested in Goldston (2012 Nuclear Fusion 52 013009). The analytical results are supported by the results of the 2D simulations of the edge plasma with reduced transport coefficients performed by SOLPS-ITER transport code.

The edge plasma of five Globus-M discharges was modeled by the B2SOLPS5.2 code. Plasma current varied in 114-198 kA range, and all discharges were in the H-mode. The modeled scrape-off layer (SOL) width appeared to be inversely proportional to the plasma current. Such a relation is observed in many other tokamaks. Heat flux to the outer divertor target and radiation power was examined for these discharges. Radiation was found to be responsible for 40% of the energy loss. Energy flux to the low outer divertor target was about 1/3 of the energy loss in the single-null low X-point discharges.

In this paper we present the fusion related activities of the Plasma Physics Division at the Ioffe Institute. The first experiments on lower hybrid current drive (LHCD) in a spherical tokamak performed at the Globus-M tokamak (R = 0.36 m, a = 0.24 m, B-t = 0.4 T, I-p = 200 kA) with a novel poloidally oriented grill resulted in an RF driven current of up to 30 kA at (100kW, 2.5 GHz), exceeding the modelling predictions. At the FT-2 tokamak (R = 0.56 m, a = 0.08 m, B-t = 3T, I-p = 30 kA) experiments with a traditional toroidally oriented grill revealed no strong dependence of the LHCD density limit on the H/D ratio in spite of LH resonance densities differing by a factor of 3. Microwave Doppler reflectometry (DR) at the Globus-M, and DR and heavy ion beam probe measurements at the tokamak TUMAN-3M (R = 0.53 m, a = 0.24 m, B-t = 1.0T, I-p = 190 kA) demonstrated geodesic acoustic mode (GAM) suppression at the L to H transition. Observations at FT-2 using Doppler Enhanced Scattering showed that the GAM amplitude is anti-correlated both spatially and temporally to the drift turbulence level and electron thermal diffusivity. For the first time turbulence amplitude modulation at the GAM frequency was found both experimentally and in global gyrokinetic modelling. A model of the L-H transition is proposed based on this effect. The loss mechanisms of energetic ions' (EI) were investigated in the neutral beam injection (NBI) experiments on Globus-M and TUMAN-3M. Empirical scaling of the 2.45 MeV DD neutron rate for the two devices shows a strong dependence on toroidal field B-t(1.29) and plasma current I-p(1.34) justifying the B-t and I-p increase by a factor of 2.5 for the proposed upgrade of Globus-M. Bursts of similar to 1 MHz Alfvenic type oscillations correlating with sawtooth crashes were observed in ohmic TUMAN-3M discharges. The possibility of low threshold parametric excitation of Bernstein and upper hybrid waves trapped in drift-wave eddies resulting in anomalous absorption in electron cyclotron resonance heating (ECRH) experiments in toroidal plasmas was identified theoretically. A novel method of radial correlation Doppler reflectometry is shown to be capable of measuring the turbulence wave-number spectrum in realistic 2D geometry. On the progress in design and fabrication of three diagnostics for ITER developed in the Ioffe institute is reported: neutral particle analysis, divertor Thomson scattering and gamma spectroscopy.

Recently a scheme for the coupling of the one-dimensional core transport code ASTRA and the two-dimensional edge transport code B2SOLPS was developed, thus providing the integrated modelling of tokamak discharge. Here, this scheme is improved by taking impurities into account and by considering a real flux surface shape using the equilibrium code SPIDER. This integrated modelling is applied to discharges of the spherical tokamak Globus-M to study the dependence of the scrape-off layer (SOL) width and divertor heat loads on the discharge power and the plasma current. Since these values, together with the magnetic field, are relatively small in Globus-M, this study can test the existing scaling against data in a wider range of tokamak operational parameters. The modelling results agree reasonably with Thomson scattering and Langmuir probe measurements and allow, in principle, the determination of the physical mechanisms responsible for the SOL structure formation. It is found that the SOL width is approximately inversely proportional to the plasma current, in agreement with existing experimental scaling, while its dependence on discharge power is found to be quite weak.

The first experiments on noninductive current drive (CD) using lower hybrid waves in a spherical tokamak are described. Waves at 2.45 GHz were launched by a 10 waveguide grill with 120 degrees phase shift between neighbouring waveguides. The experimental results for a novel poloidal slowing-down scheme are described. The CD efficiency is found to be somewhat larger than that predicted theoretically whilst at the same time being somewhat less than that for standard tokamak lower hybrid CD. Geodesic acoustic modes (GAM) have been discovered in Globus-M. GAMs are localized 2-3 cm inside the separatrix. The GAM frequency agrees with theory. The mode structures of plasma density and magnetic field oscillation at the GAM frequency have been studied. Fast particle confinement during neutral beam injection has been investigated and numerically simulated. Alfven instabilities excited by fast particles were detected by a toroidal Mirnov probe array. Their excitation conditions are discussed and the dynamics of fast ion losses induced by Alfven eigenmodes is presented. Preliminary experiments on the isotopic effect influence on global confinement in the ohmic heating (OH) regime are described. Scrape-off layer (SOL) parameters were measured and compared with results from self-consistent integrated transport modelling. Results showed that SOL width scales inversely proportional to plasma current. The behaviour of an a priori damaged tungsten divertor plate mock-up exposed to plasma flows was investigated. Preliminary conclusions are that the initial damage gives rise to a loose layer formation with low thermal conductivity right beneath the surface. Finally, engineering design issues of the next step-Globus-M2 (1T, 500 kA) and the status of component manufacture are described.

The reduction of heat loads to the material walls and divertor targets is the key issue of a future tokamak reactor. Recently a heuristic drift-based model of heat transport at the tokamak edge plasma was introduced [1], and then tested versus data from large machines like ASDEX Upgrade and JET [2]. The corresponding scaling predicts the power scrape-off width in the ITER H-mode to be 1 mm, which is critical for ITER operational performance. In this report we present transport study of the edge plasma of Globus-M spherical tokamak, which has smaller values of plasma current and power – the main parameters scrape-off width depends on according to Goldston's scaling [2]. Consequently, with the Globus-M data the Goldston's scaling can be verified against data in wider range of tokamak operational parameters. For analysis chosen were the NBI-heated shot #29076, I pl =185 kA, and the OH shot #30095, I pl =170 kA, (both in double-null divertor topology with active lower Xpoint) so we can study the power dependence of SOL width λq . Unfortunately, the more important plasma current (I pl) scan still was not performed on Globus-M. The transport processes were modeled by coupling of 1D core and 2D edge transport codes (namely ASTRA and B2SOLPS5.2) [3]. This coupling procedure, briefly described in [3], results in continuous self-consistent profiles of density, temperature, particle and heat fluxes from the magnetic axis up to divertor targets. Values of anomalous transport coefficients were fitted in order to reach satisfactory agreement between calculated profiles and experimental data, coming from Thomson scattering, Langmuir probes, infrared camera and Doppler reflectometry. According to previous results of Globus-M transport modeling [4], anomalous ion heat conductivity was set to zero being much less than the neoclassical one.