Search

The tokamak scrape-off layer (SOL) model is discussed on the assumption of drift mechanisms of electron heat transport and dominant role of recycling outside the separatrix comparing to the ion outflow from the core. The SOL width estimate is given. It is shown that drift mechanisms give minimal SOL width corresponding to present day scalings which is close by the order of magnitude to the experimental values.

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λq) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λq, which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ion flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (Btor and Bpol) are conducted, showing minimal λq dependence on either component of the field. Insensitivity to Btor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λq dependence on Bpol, however, is inconsistent with both the HD model and experimental results. The inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model. Keywords: Power scrape-off width, SOLPS-ITER, Drift effects, SOL flows

SOLPS-ITER is a code package for tokamak plasma boundary modeling comprising the B2.5 plasma fluid code and the EIRENE neutral kinetic code. The form of the momentum balance equation, used in the SOLPS-ITER code before version 3.0.6 (and in all previous versions of SOLPS codes), was simplified due to numerical reasons—the electric field term was replaced by the electron gradient pressure term; friction and thermal force terms were derived using the trace impurity assumption. It has now been decided to switch back to the general Braginskii version of this equation. Derivation of the new form of the parallel friction and thermal force terms has been required to allow the use of the general form of the equation and allow more accurate modeling of multicomponent plasmas. The new form is based on the theoretical description of multicomponent plasma derived by V. M. Zhdanov [Plasma Phys. Controlled Fusion 44(10), 2283 (2002)]. In the present paper, a detailed derivation of these terms is presented, and their implementation into the SOLPS-ITER code package is discussed.

Currents 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 of different physical nature. Besides known Pfirsch-Schluter (PS) currents and thermoelectric currents, so-called PCC (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 particular, current flowing to the outer plate in the private flux region (PFR), opposite to thermolelectric current is predicted for the standard single-null configuration and favorable direction of ∇B drift (∇B drift of ions is directed towards active X-point). In addition, a pair of currents to and away from the outer plate should flow. In the single-null configuration they are often masked by a larger thermoelectric current, however for the connected double null (CDN) case, where thermoelectric current is strongly reduced due to less temperature asymmetry, these currents dominate. The suggested physical model is supported by results of simulations performed with SOLPS-ITER transport code. Simulations were done for ASDEX-Upgrade (AUG), L-mode , single-null configurations, and for Globus-M H-mode, both disconnected and connected double null configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found.

By modeling done with the SOLPS-ITER transport code, it is demonstrated that moderate radiating impurity seeding leads to the increase in the radial electric field by absolute value, while radial electric field remains close to the neoclassical one. In the seeded detached regime with a highly radiating X-point, the radial electric field deviates considerably from the neoclassical electric field. The radial dimension of the zone with increased poloidal E × B rotation shear is larger by absolute value than that corresponding to the moderate seeding. This effect might lead to the shift of the transport barrier further to the core and can partially compensate energy losses rise inside the separatrix, so the pedestal improvement might take place.

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.

Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.

Impurity seeding into a tokamak divertor for radiative cooling is considered as a tool for achieving detached/semi-detached regimes required to meet the condition of acceptable heat loads on divertor plates. Experiments aimed at searching for an operational window with a significant reduction of poloidal heat fluxes due to the impurity radiation and without decrease of confinement are performed on many tokamaks. A critical issue in these experiments is how large a fraction of impurities is retained in the divertor region and how much is extracted upstream to the scrape-off layer. In the present paper a physical mechanism of impurity transport from a divertor towards upstream and back to the divertor is analyzed. It is demonstrated that the widespread concept that the impurity leaks if the parallel thermal force exceeds the friction due to main ions and is retained otherwise—is not correct. In this paper, we contend that the impurity leaks if it crosses the stagnation point of the impurity ion poloidal velocity profile before being ionized, and is retained if it ionizes closer to the target than the location of that stagnation point. Thus the leakage efficiency depends on the relative spatial positions of the impurity atom ionization source and the stagnation point of the impurity ion poloidal velocity profile. The impurity ion poloidal velocity is to large extent the sum of the poloidal projection of its parallel velocity and the E × B drift velocity, where the former is derived from the parallel impurity force balance equation. It is demonstrated that the solution of this equation may be approximated by the balance of friction and thermal forces in all regimes, while other terms are smaller. This allows for expressing the impurity parallel velocity through the main ion one and makes the distribution of the parallel (poloidal) fluxes of the main ions, including Pfirsch–Schluter fluxes and E × B drift fluxes, to be an important element of the impurity transport. It is shown that the impurity distribution in the edge plasma is rather sensitive to the value of the impurity ion ionization rate. This analysis is supported by simulation results obtained for the ASDEX Upgrade tokamak with various seeding rates of N and Ne with the SOLPS-ITER code. The importance of the inclusion of self-consistent drift flows is demonstrated by comparison to results of corresponding simulations with the drifts turned off.

Reported is the first modeling of ITER H-mode. Simulations were made with the code B2SOLPS5.2 with drifts and currents, which was specially developed for simulation of tokamak H-regimes. The emphasis is made on the pedestal and edge transport barrier regions. It is demonstrated that the specific feature of the ITER discharges is the density profile in the edge barrier region typical for the low particle sources in the main part of the barrier. The density, temperature and radial electric field profiles are calculated for two values of the edge barrier width (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A new version of the B2SOLPS5.0 transport code, which is free from numerical problems in the barrier region, has been used to simulate H-mode shots from ASDEX-Upgrade and MAST. The radial electric field inside the edge transport barrier and in the pedestal region is close to the neoclassical prediction. The shear of poloidal E → × B → drift at the inner side of the barrier is close to the value before the transition, while inside the barrier it is significantly larger. It is demonstrated that to match the experimental density and temperature radial profiles the drop in the diffusion coefficient within the barrier should be significantly larger than the drop in the electron heat conductivity.

25th Fusion Energy Conference (FEC), Govt Russian Federat, St Petersburg, RUSSIA, OCT 13-18, 2014; International audience; 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.

A method of reduction of the two-dimensional equation for the potential to a one-dimensional ordinary equation is suggested and implemented in the B2SOLPS5.0 transport code to simulate a divertor tokamak. The one-dimensional version, which gives similar results as the full two-dimensional version, provides better understanding of the role of various mechanisms of perpendicular conductivity and reduces computational time. The scheme of poloidal (parallel) current calculation is presented and compared to the two-dimensional code results both for the core region and for the scrape-off layer.

The impurity transport in the edge plasma of a divertor tokamak is simulated by means of B2-SOLPS5.0 2D multi-fluid code where diamagnetic and E → × B → drifts for all species are taken into account. It is demonstrated that these drifts to large extent determine the poloidal distribution of impurities near the separatrix and hence penetration of impurities to the core region. The influence of the drifts on the low ionized impurity distribution in the divertor region is more modest.

Mechanisms of the impurity transfer inside a tokamak separatrix have been studied. It is shown that poloidal rotation significantly influences the poloidal distribution of impurities. An analysis of the neoclassical component of the radial impurity flux showed that this fraction is described by an expression that is more complicated than the standard neoclassical variant.

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 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° 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 (1 T, 500 kA) and the status of component manufacture are described.

Simulation of an H-mode ASDEX Upgrade shot with boron impurity was done with the B2SOLPS5.2 transport code. Simulation results were compared with the unique experimental data available for the chosen shot: radial density, electron and ion temperature profiles in the equatorial midplanes, radial electric field profile, radial profiles of the parallel velocity of impurities at the low-field side (LFS) and high-field side (HFS), radial density profiles of impurity ions at LHS and HFS. Simulation results reproduce all available experimental data simultaneously. In particular strong poloidal HFS?LFS asymmetry of B5+ ions was predicted in accordance with the experiment. The simulated HFS B5+ density inside the edge transport barrier is twice larger than that at LFS. This is consistent with the experimental observations where even larger impurity density asymmetry was observed. A similar effect was predicted in the simulation done for the MAST H-mode. Here the HFS density of He2+ is predicted to be 4 times larger than that at LHS. Such a large predicted asymmetry is connected with a larger ratio of HFS and LFS magnetic fields which is typical for spherical tokamaks. The HFS/LFS asymmetry was not measured in the experiment, however modelling qualitatively reproduces the observed change of sign of He+parallel velocity to the counter-current direction at LFS. The understanding of the asymmetry is based on neoclassical effects in plasma with strong gradients. It is demonstrated that simulation results obtained with account of sources of ionization, realistic geometry and turbulent transport are consistent with the simplified analytical approach. Difference from the standard neoclassical theory is emphasized.

The fluid simulation of a divertor tokamak edge plasma by the B2-SOLPS5.0 transport code gives the dependence of the radial electric field on the local and global plasma parameters. The shear of the radial electric field, which is responsible for the transition to an improved confinement regime, is a linear function of a local ion temperature and the local average toroidal velocity and is inversely proportional to the toroidal magnetic field. The scaling for the L-H transition threshold agrees with the experimental H-mode scaling of ASDEX Upgrade. The radial electric field shows no bifurcation and is close to the neoclassical electric field with the toroidal rotation contribution determined by the radial anomalous transport of the toroidal momentum. The fine structure of the electric field at the separatrix and its dependence on the toroidal magnetic field inversion is analysed.

The biasing experiments and the problem of the effective perpendicular conductivity in a tokamak are studied both analytically and numerically. The simulations were performed by means of the B2SOLPS5.0 transport code [V. A. Rozhansky et al., Nucl. Fusion 41, 387 (2001)] where the fluid set of equations is solved for the real divertor geometry. It is demonstrated that there exist three regimes, which correspond to the different values of the effective perpendicular conductivity and to different profiles of the radial electric field and toroidal and poloidal rotations. Simulation results are compared with existing and developed analytical models.

The radial electric field in the vicinity of the separatrix and its shear is studied in detail by means of the B2-SOLPS5.0 transport code. Calculations are performed for various regimes of ASDEX Upgrade. The dependencies of the radial electric field on the local temperature, density, toroidal and poloidal magnetic fields and mean toroidal rotation are investigated.

A complete system of transport equations with all the important perpendicular currents is derived for the simulation of tokamak edge plasma. These transport equations are implemented in the B2.5 code and solved for the parameters of the ASDEX Upgrade tokamak. The relative roles of different mechanisms of transverse conductivity in the formation of the potential profile are studied. It is demonstrated that a reasonable potential distribution in the tokamak edge plasma can be obtained without an ad hoc assumption of the existence of the anomalous perpendicular conductivity. The role of E ? B drifts in the redistribution of edge plasma and closing of the currents in the plasma is analysed.

The numerical codes ASTRA and B2SOLPS5.2 are coupled to perform an integrated modeling of particle and energy transport and to obtain continuous self-consistent profiles of the main plasma parameters from the magnetic axis up to target plates. The unique distinguishing feature of the new coupling scheme is the presence of a region of overlap of the 1D and 2D computational domains, where the 1D solution coincides with the 2D one at the equatorial midplane. In the 2D transport equation system, all relevant drift flows and currents are taken into account, which allows us to calculate the poloidal variation of the density, temperatures and electrostatic potential, and obtain neoclassical radial fluxes in a self-consistent manner. Such an approach allows us to model tokamaks for which neoclassical effects give a significant contribution to the ion heat transport, and in particular, spherical tokamaks.

The radial electric field and toroidal rotation of tokamak plasma near a magnetic island have been calculated. Outside the magnetic island, the radial electric field varies from a value determined by the rotation of this island to the neoclassical value over a scale dependent on the anomalous viscosity and collisionality of the plasma. Inside the magnetic island, the radial electric field is constant.

SOLPS-ITER, a comprehensive 2D scrape-off layer modeling package, is used to examine the physical mechanisms that set the scrape-off width () for inter-ELM power exhaust. Guided by Goldston's heuristic drift (HD) model, which shows remarkable quantitative agreement with experimental data, this research examines drift effects on in a DIII-D H-mode magnetic equilibrium. As a numerical expedient, a low target recycling coefficient of 0.9 is used in the simulations, resulting in outer target plasma that is sheath limited instead of conduction limited as in the experiment. Scrape-off layer (SOL) particle diffusivity (D SOL) is scanned from 1 to 0.1 m2 s−1. Across this diffusivity range, outer divertor heat flux is dominated by a narrow (~3–4 mm when mapped to the outer midplane) electron convection channel associated with thermoelectric current through the SOL from outer to inner divertor. An order-unity up–down ion pressure asymmetry allows net ion drift flux across the separatrix, facilitated by an artificial mechanism that mimics the anomalous electron transport required for overall ambipolarity in the HD model. At m2 s−1, the density fall-off length is similar to the electron temperature fall-off length, as predicted by the HD model and as seen experimentally. This research represents a step toward a deeper understanding of the power scrape-off width, and serves as a basis for extending fluid modeling to more experimentally relevant, high-collisionality regimes.

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.

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 radial electric field Er at the tokamak plasma edge is simulated both with a two-dimensional (2D) fluid code solving the most complete system of transport equations and with five-dimensional (three-dimensional in configuration space and 2D in velocity space) Monte Carlo particle following code. At low to high confinement transition conditions, the E→r×B→ shearing rate is found to be high enough for turbulence suppression even though the field is essentially neoclassical. Here, B→ is the magnetic field. No bifurcation of Er is found.

Plasma rotation and radial electric field in the presence of resonant magnetic perturbations (RMPs) is considered. The situation is discussed in detail where no change of poloidal rotation is observed while the radial electric field and toroidal rotation are changing self-consistently when RMPs are switched on. Recent experimental results from DIII-D are discussed in connection with this model.

The non-linear reduced four-field RMHD model in cylindrical geometry was extended to include plasma rotation, neoclassical poloidal viscosity and two fluid diamagnetic effects. Interaction of the static resonant magnetic perturbations (RMPs) with the rotating plasmas in tokamaks was studied. The self-consistent evolution of equilibrium electric field due to RMP penetration is taken into account in the model. It is demonstrated that in the pedestal region with steep pressure gradients, mean flows perpendicular to the magnetic field, which includes and electron diamagnetic components plays an essential role in RMP screening by plasma. Generally, the screening effect increases for lower resistivity, stronger rotation and smaller RMP amplitude. Strong screening of central islands was observed limiting RMP penetration to the narrow region near the separatrix. However, at certain plasma parameters and due to the non-linear evolution of the radial electric field produced by RMPs, the rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands. Typical plasma parameters and RMPs spectra on DIII-D, JET and ITER were used in modelling examples presented in the paper.

Steady-state screening of resonant magnetic perturbations (RMPs) in a tokamak is analysed taking into account a self-consistent electric field. On the one hand, the self-consistent radial electric field is determined by the balance of the electron radial conductivity in a stochastic magnetic field screened by the plasma and by the neoclassical ion conductivity. On the other hand, the parallel current of electrons, the radial projection of which is balanced by the ion current, determines the screening of RMPs. In this work, the self-consistent electric field and RMP screening are calculated. Two different regimes of screening are found: the ‘ion’ branch which corresponds to the negative radial electric field and the ‘electron’ branch for which the electric field is positive. Predictions of the model are compared with the experimental data and results of the simulation with various codes. The corresponding toroidal rotation and pump-out effect are discussed.

A new B2SOLPS5.2 transport code has been developed and implemented for the simulation of H-mode shots. A new equation system is proposed, which is equivalent to the system which was used in B2SOLPS5.0 previously. The main idea is to replace the major part of the large radial ∇B driven convective fluxes by poloidal fluxes with the same divergence both in the particle balance and in the energy balance equations. This is of special importance for the H-mode where the diffusion coefficient is strongly reduced inside the barrier and large radial convective flows are strongly undesirable from the numerical point of view. The H-mode shots of ASDEX-Upgrade and MAST have been simulated with the new version with reasonable time steps and convergence. It is demonstrated that the radial electric field inside the edge transport barrier and in the pedestal region is close to the neoclassical electric field as in previous simulations of Ohmic shots. The toroidal rotation is co-current directed as in L-mode but is significantly larger in absolute value. It is shown that the shear of the poloidal drift at the inner side of the barrier is close to the value of the shear before the transition, while inside the barrier the value of the shear is significantly bigger. This fact determines self-consistently the width of the edge transport barrier. It is demonstrated that to match the experimental density and temperature radial profiles the drop in the diffusion coefficient within the barrier needs to be significantly larger than the drop in the electron heat conductivity coefficient. For the H-mode the pedestal region usually corresponds to the collisionless regime, so several corrections were introduced into the transport coefficients to extend the applicability of the code to the plateau and banana regimes in the inner regions of the simulation domain.

A theoretical model for the toroidal rotation spin-up and generation of a positive radial electric field during the stochastization of plasma edge is put forward. Equations for the toroidal velocity and for modification of the core radial electric field have been derived. A detailed comparison of the model results with the radial electric field and plasma potential measurements during MHD activity in the TUMAN-3M tokamak is presented.

Account of drifts and currents dramatically decreases the accessible time step for the integration of time dependent equations of SOLPS-ITER code for edge modeling. Running the code with sophisticated EIRENE Monte-Carlo model for neutrals and large number of fluid equations for multiple ion species makes the computation time unacceptably long. In the paper the main mechanisms leading to the time step limitations caused by drifts are discussed. Several methods of the suppression of these mechanisms are suggested and the results of numerical scheme tests with the applied corrections are presented. Application of these schemes decreases the time of convergence to steady state solution by more than an order of magnitude.