Showing papers in "Nuclear Fusion in 1997"

Theory for avalanche of runaway electrons in tokamaks

Abstract: An analysis is presented of runaway electron formation and its evolution during disruptions in large tokamaks, where avalanche phenomena play a crucial role A simplified, but sufficiently accurate, analytical model suitable for one dimensional (1-D) transport codes is proposed Validation of the model was done by comparison with Monte Carlo calculations

H mode confinement in Alcator C-Mod

Abstract: A series of experiments, examining the confinement properties of ion cyclotron range of frequencies (ICRF) heated H mode plasmas, has been carried out on the Alcator C-Mod tokamak. Alcator C-Mod is a compact tokamak that operates at high particle, power and current densities at toroidal fields up to 8 T. Under these conditions the plasma is essentially thermal with very little contribution to the stored energy from energetic ions (typically no more than 5%) and with Ti~Te. Most of the data were taken with the machine in a single null `closed' divertor configuration with the plasma facing components clad in molybdenum tiles. The data include those taken both before and after the first wall surfaces were coated with boron, with emphasis on the latter. H modes obtained from plasmas run on boronized walls typically had a lower impurity content and radiated power and attained a higher stored energy than those run on bare molybdenum. Confinement enhancement, the energy confinement time normalized to L mode scaling, for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the Alcator C-Mod device provided a means for extending the tests of global scaling laws to parameter ranges not previously accessible. For example, the Alcator C-Mod edge localized mode (ELM)-free data were found to be 1.1 to 1.6 times the ITERH93 scaling and the ELMy data almost 2.0 to 2.8 times the ITERH92 ELMy scaling law, suggesting that the size scaling in both scalings may be too strong. While both ELM-free and ELMy discharges were produced, the ELM characteristics were not easily compared with observations on other devices. No large, low frequency ELMs were seen despite the very high edge pressure and temperature gradients that were attained. For all of our H mode discharges, a clear linear relationship between the edge temperature pedestal and the temperature gradient in the core plasma was observed; the discharges with the `best' transport barriers also showing the greatest improvement in core c

ITER L mode confinement database

Abstract: This special topic describes the contents of an L mode database that has been compiled with data from Alcator C-Mod, ASDEX, DIII, DIII-D, FTU, JET, JFT-2M, JT-60, PBX-M, PDX, T-10, TEXTOR, TFTR and Tore Supra. The database consists of a total of 2938 entries, 1881 of which are in the L phase while 922 are ohmically heated only (ohmic). Each entry contains up to 95 descriptive parameters, including global and kinetic information, machine conditioning and configuration. The special topic presents a description of the database and the variables contained therein, and it also presents global and thermal scalings along with predictions for ITER. The L mode thermal confinement time scaling, determined from a subset of 1312 entries for which the τE,th are provided, is τE,th = 0.023Ip0.96BT0.03R1.83(R/a)0.06 κ0.64ne0.40Meff0.20P-0.73 in units of seconds, megamps, teslas, metres, -, -, 10-9 m-1

Tokamak disruption alarm based on a neural network model of the high- beta limit

Abstract: An artificial neural network, combining signals from a large number of plasma diagnostics, was used to estimate the high- beta disruption boundary in the DIII-D tokamak. It is shown that inclusion of many diagnostic measurements results in a much more accurate prediction of the disruption boundary than that provided by the traditional Troyon limit. A trained neural network constitutes a non-linear, non-parametric model of the disruption boundary. Through the analysis of the input-output sensitivities, the relative statistical significance of various diagnostic measurements (plasma parameters) for the determination of the disruption boundary is directly assessed and the number of diagnostics used by the neural network model is reduced to the necessary minimum. The neural network is trained to map the disruption boundary throughout most of the discharge. As a result, it can predict the high- beta disruption boundary on a time-scale of the order of 100 ms (much longer than the precursor growth time), which makes this approach ideally suitable for real time application in a disruption avoidance scheme. Owing to the relative simplicity of the required computations, the neural network is easily implemented in a real time system. A prototype of the neural network disruption alarm was installed within the DIII-D digital plasma control system, and its real time operation, with a typical time resolution of 10 ms, was demonstrated

High beta tokamak operation in DIII-D limited at low density/collisionality by resistive tearing modes

Abstract: The maximum operational high beta in single-null divertor (SND) long pulse tokamak discharges in the DIII-D tokamak with a cross-sectional shape similar to the proposed International Thermonuclear Experimental Reactor (ITER) device is found to be limited by the onset of resistive instabilities that have the characteristics of neoclassically destabilized tearing modes. There is a soft limit due to the onset of an m/n=3/2 rotating tearing mode that saturates at low amplitude and a hard limit at slightly higher beta due to the onset of an m/n=2/1 rotating tearing mode that grows, slows down and locks. By operating at higher density and thus collisionality, the practical beta limit due to resistive tearing modes approaches the ideal magnetohydrodynamic (MHD) limit

Ideal MHD stability limits of low aspect ratio tokamak plasmas

Abstract: The ideal magnetohydrodynamic (MHD) stability limits of low aspect ratio tokamak plasmas are computed numerically for plasmas with a range of cylindrical safety factors q*, normalized plasma pressures beta , elongations kappa and central safety factors q(0). Four distinct regimes are optimized, namely: (a) low-q* plasmas with q(0)=1.1 with and without a stabilizing wall, (b) low-q* plasmas with no wall and 1.1

Radiative divertor experiments in DIII-D with D2 injection

Abstract: Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: Ip=1.0-2.0 MA, q95 approximately= 2.4-6.0 and total input power (20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached' divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q95 equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

Poloidal distribution of impurities in a rotating tokamak plasma

Abstract: Toroidal rotation of a tokamak plasma causes an outward major radial displacement of impurities. The physics of this effect is described and the distribution of impurities on a magnetic surface is calculated for the cases of practical interest

Neoclassical islands on COMPASS-D

Abstract: Neoclassical magnetic islands are observed to limit the achievable beta in COMPASS-D low collisionality single null divertor tokamak plasmas with ITER-like geometry (R0=0.56 m, B0=1.2 T, Ip=120-180 kA, k=1.6, epsilon =0.3). The limiting beta is typically well below that expected from ideal instabilities with maximum beta N in the range of 1.6 to 2.1. The plasma is heated with up to 1.8 MW of 60 GHz electron cyclotron resonance heating (ECRH) at the second harmonic with X mode polarization. The time history of the measured island width is compared with the predictions of neoclassical tearing mode theory, with good agreement between theory and experiment. The measured islands have a threshold width below which the mode will not grow. The density scaling of the point of onset of the measured instabilities is compared with two theories that predict a threshold island width for the onset of neoclassical tearing modes. Applied resonant helical error fields are used to induce islands in collisionality regimes wherein the neoclassical islands do not occur naturally, allowing the study of the behaviour of neoclassical tearing modes in this regime. The critical beta for the onset of neoclassical tearing modes is seen to be ~3 times higher in the naturally stable region. This observation is compared with the predictions of both threshold theories. A simple expression for the q scaling of the maximum achievable beta N in the presence of neoclassical tearing modes is derived on the basis of the assumption of a maximum allowable island width. The predicted q scaling of this beta limit is compared with data from a q scan, and the results are in good agreement

Liquid first walls for magnetic fusion energy configurations

Abstract: Liquids (~7 neutron mean free paths thick), with certain restrictions, can probably be used in magnetic fusion designs between the burning plasma and the structural materials of the fusion power core. If this works there would be a number of profound advantages: a cost of electricity lower by as much as a factor of 2; removal of the need to develop new first wall materials, saving over 4 billion US dollars in development costs; a reduction of the amount and kinds of wastes generated in the plant; and the wider choice of materials permitted. The amount of material that evaporates from the liquid which can be allowed to enter the burning plasma is estimated to be less than 0.7% for lithium, 1.9% for Flibe (Li2BeF4 or LiBeF3) and 0.01% for Li17Pb83. The ability of the edge plasma to attenuate the vapour by ionization appears to exceed this requirement. This ionized vapour would be swept along open field lines into a remote burial chamber. The most practical systems would be those with topological open field lines on the outer surface, as is the case with a field reversed configuration (FRC), a spheromak, a Z pinch or a mirror machine. In a tokamak, including a spherical tokamak, the field lines outside the separatrix are restricted to a small volume inside the toroidal coil making for difficulties in introducing the liquid and removing the ionized vapour, i.e. the configuration is not open ended

Interaction of stochastic boundary layer with plasma facing components

Abstract: To alleviate the plasma-wall interaction problems in magnetic confinement devices, a stochastic layer is used at the edge of the Tore Supra tokamak (ergodic divertor). A very important point is to determine the power deposition on the plasma facing components. Two different kinds of transport can be identified in such a configuration: stochastic transport surrounding the confined plasma, with a random walk process, and scrape-off layer (SOL) like transport, a laminar transport, near the plasma facing components. The laminar regime is investigated in terms of a simple criterion, namely that the power deposition is proportional to the radial penetration of the laminar zone flux tubes over a finite parallel length. The magnetic connection properties of the first wall components are then determined. The connection lengths are quantified with two characteristic scales. The larger corresponds to one poloidal turn and appears to be the characteristic parallel length for laminar transport. A field line tracing code MASTOC (magnetic stochastic configuration) is used to compute the complex topology and the statistics of the connection in the real tokamak geometry. The numerical simulations are then compared with the experimental heat deposition on the modules and neutralizer plates of the Tore Supra ergodic divertor. Good agreement is found. Further evidence of laminar transport is also provided by the tangential view of such structures revealed from Halpha structures in detached plasma experiments

An international pellet ablation database

Abstract: The contents are described of an international pellet ablation database (IPADBASE) that has been assembled to enable studies of pellet ablation theories that are used to describe the physics of an ablating fuel pellet in a tokamak plasma. The database represents an international effort to assemble data from several tokamaks of different magnetic configurations and auxiliary heating methods. In the initial configuration, data from JET, Tore Supra, DIII-D, FTU, TFTR, ASDEX Upgrade, JIPP T-IIU, RTP and T-10 have been included. The database contains details of measurements of deuterium and hydrogen pellet ablation, including pellet mass and speed, plasma electron density and temperature profiles, and pellet ablation light emission. A summary of the database contents and a scaling analysis of the data are presented

Feedback control of the current profile on Tore Supra

Abstract: Recent results on feedback control of the global shape of the current density profile in discharges with lower hybrid current drive (LHCD) on Tore Supra are presented. The global shape of the current density profile is characterized by its internal inductance li, and feedback controlled through the LH power or the launched LH wave spectrum. Feedback control of the flux on the plasma boundary has allowed for exploration of regimes with partial current drive (constant loop voltage) and full current drive (zero loop voltage). In stationary, steady state discharges at 0.8 MA (qa=7), the current profile shape is characterized by an li between 1.5 and 1.7 and a safety factor on-axis (q0) between 1 and 1.4. The energy confinement is 1.3 to 1.6 times higher than the value predicted by the Rebut-Lallia-Watkins scaling law

X-ray observations of central toroidal rotation in ohmic Alcator C-Mod Plasmas

Abstract: Impurity toroidal rotation has been observed in the centre of ohmic plasmas in the Alcator C-Mod tokamak from the Doppler shifts of argon and molybdenum X-ray lines. The rotation is highest (~6 × 106 cm/s) in the early portion of the discharges, when the loop voltage is highest and the electron density is lowest, and then typically settles to values ≤ 2 × 106 cm/s during the steady state period. The impurity rotation is in the same direction as the electron toroidal drift, opposite to the plasma current, and reverses direction when the plasma current direction is reversed. Molybdenum and argon ions rotate with the same velocity. These observations are in qualitative agreement with neoclassical theory

Analysis of coupled MHD modes with Mirnov probes in ASDEX Upgrade

Abstract: With the help of a simulation of the perturbation field and a comparison with Mirnov measurements, it becomes possible to investigate coupled MHD modes. The structure and the width of the different coupled islands can be estimated. For a reliable simulation, it is necessary to take into account the induced screening currents in the vessel wall and in some of the in-vessel components. The fields of these screening currents have a large influence on the measurements. This improved analysis method is demonstrated for the development of coupled modes during a density limit disruption. It is clearly shown that an increase of the coupled islands leads to minor disruptions. Owing to mode locking, reliable results are not possible before major disruptions

Impurity induced neutralization of megaelectronvolt energy protons in JET plasmas

Abstract: A neutral particle analyser was deployed on JET for a measurement of the distribution function of megaelectronvolt energy protons driven by the high power ion cyclotron resonance frequency heating of deuterium plasmas in the hydrogen minority scheme. Unexpectedly, efficient neutralization of megaelectronvolt energy protons was observed in the plasma centre without recourse to injection of a beam of atoms to provide electron donors for charge exchange neutralization reactions. A model is presented that elucidates the role of carbon and beryllium, the main intrinsic impurities in JET plasmas, in this process and establishes charge exchange between hydrogen-like ions of the impurities and protons as the main neutralization process. A model calculation for deducing the proton energy distribution function f(Ep) from the measured hydrogen flux is described. The effects of uncertainties in neutralization cross-sections on the inferred f(Ep) are examined. The validity of this model of impurity induced neutralization (IIN) is tested by using it to describe the measured hydrogen flux in different conditions of plasma heating and fuelling. A crucial point in making these tests was to use measured local values of bare impurity ion densities; the required measurements were realized using charge exchange spectroscopy. Using IIN modelling and a procedure in which a known change in the density of deuterium atoms at the plasma centre was made by applying neutral beam injection, we have deduced the background thermal deuterium atom density at the plasma centre, which is an important new diagnostic result. Concerning future experiments, the model predicts that carbon and beryllium impurities will be major contributors to neutralization of hydrogenic ions (protons, deuterons and tritons) in ITER plasmas, for ion energies E

High density operation close to Greenwald limit and H mode limit in ASDEX upgrade

Abstract: The L mode and H mode density operational window in the vicinity of the density limit has been investigated with a combination of gas puff refuelling and improved fine tuning of neutral beam injection (NBI) heating power. In this way, a novel strategy is achieved by means of a parallel increase of density and heating power. As the density limit is approached, H modes degrade into L modes independently of heating power; this is in contrast to the generally accepted L to H mode threshold scaling PheatL-H varies as neB. Furthermore, contrary to the well known heating power independent Greenwald limit, the L mode density limit increases moderately with rising heating power, neDL varies as Pheat0.3+or-0.1, if a simple power law is assumed. The power dependence becomes more obvious when analysed in terms of edge densities and powers flowing across the separatrix into the scrape-off layer, nesep varies as Psep0.6+or-0.2. The corresponding H mode studies show that before an H mode quenches into an L mode the maximum achievable density (i.e. The H mode density limit) is practically independent of the heating power, as observed on many machines

Finite-β magnetic field line tracing for Helias configurations

Abstract: A new code, Magnetic field solver for Finite-Beta Equilibria (MFBE), has been developed. It computes the magnetic field in a form suitable for field line tracing inside and outside the plasma boundary of finite-β Helias equilibria calculated with the NEMEC free boundary equilibrium code. Coupling of the NEMEC and MFBE codes allows a determination of the last closed magnetic surface. For various beta values up to (β) 0.04, the variations of the island sizes and positions outside the plasma boundary and the ergodization of the edge region are studied for a Helical Advanced Stellarator (Helias) configuration (Nuhrenberg and Zille, Phys. Lett. A 114 (1986) 129)

Observation of dust particles by a laser scattering method in the JIPPT-IIU tokamak

Abstract: Laser scattering signals that indicate the presence of small dust particles (diameter ≤ 2 μm) have been occasionally observed in the JIPPT-IIU tokamak chamber. This phenomenon was reproduced deliberately by spreading carbon dust from the top of the vacuum chamber. No noticeable effect on the plasma was observed for dust falls of up to at least 106 particles (10 μg) in 20 ms during discharges. Dust falling just before plasma startup seemed to be confined but it was soon ejected (in less than 30 ms)

Correlation between potential well structure and neutron production in inertial electrostatic confinement fusion

Abstract: The electrostatic potential well in inertial electrostatic confinement (IEC) is studied using two approaches. First, the equilibrium potential profile is obtained by solving the charge neutrality condition, i.e. ni=ne, assuming the appropriate distribution functions for the ions and the electrons. The formation of a double well structure is demonstrated, with a depth depending upon the ratio between the focus radii of the electrons and the ions. The correlations between the well depth and the volume integrated neutron production due to deuterium-deuterium (DD) reactions are obtained. Second, in order to study the stability of the well, the dynamic behaviours of the potential well are calculated by performing time advancing numerical simulations on the basis of the particle in cell method. Single, double and triple wells, depending on the amount of injected ion current, are observed to be formed for ions with a monoenergetic distribution. The well in the centre of the multiwell structure is unstable and oscillates with a period much longer than the inverse ion plasma frequency. A double well structure can be formed even for ions with a spread out energy distribution when the ion current is larger than the threshold value. The time averaged neutron production by DD fusion events is proportional to a power of the ion current involved in forming the double well structure. The results strongly suggest that the high neutron production rate should be attributed to not only the well depth but also the unstable behaviour of the potential, i.e. the intermittent peaking of the density in the centre region. A numerical simulation reveals that IEC possesses a favourable dependence of fusion reactions on the injected ion current for the application to a neutron source or a fusion reactor

Effect of plasma radial electric field on motional Stark effect measurements and equilibrium reconstruction

Abstract: The motional Stark effect (MSE) has become the leading diagnostic technique for measurement of the internal poloidal field profile in neutral beam heated tokamak plasmas. The technique relies upon the measurement of the polarization angle of Stark broadened neutral beam Dalpha emission due to the strong E= nu b*B motional electric field. In many recently discovered enhanced confinement regimes in tokamaks, a strong radial electric field Er is also present in the plasma. It is shown that in these cases, the effect of Er on the interpretation of MSE measurements is significant and cannot be neglected. The importance of this effect is illustrated with the analysis of the q profile in a high performance discharge from the DIII-D tokamak

Analysis of eta i mode by reactive and dissipative descriptions and the effects of magnetic q and negative shear on the transport

Abstract: Ion temperature gradient driven modes are studied in the short wavelength region by using a three pole reactive fluid model and an adiabatic E*B convection model, as well as the gyro-Landau fluid model. In all of the above models the parallel ion dynamics is included and the results have been compared both analytically and numerically. Furthermore, the effects of the magnetic safety factor q on the growth rate and transport coefficient chi i have been investigated. The maximum of chi i as a function of the mode number is computed, and the stabilizing effect of the negative magnetic shear has been studied

Comparison of the CREATE-L plasma response model with TCV limited discharges

Abstract: Experiments have been performed on the TCV tokamak to evaluate the response of ohmic, L mode, limited, vertically unstable plasmas to changes in all the poloidal field coil voltages. The resulting closed loop plasma responses have been compared with the CREATE-L linearized MHD equilibrium model of the TCV tokamak. All the responses in both the time domain and the frequency domain show excellent agreement both for directly measured quantities and for derived parameters. No modifications to the CREATE-L model were made to achieve this quality of agreement, indicating that the underlying physical assumptions are appropriate.

Tearing mode formation and radiative edge cooling prior to density limit disruptions in ASDEX Upgrade

Abstract: Tearing mode formation is investigated for ohmic density limit discharges on the ASDEX Upgrade tokamak at medium and high safety factors (q95=3.8, 4.9, 6.0). Low electron temperatures inside magnetic islands and the observation of localized C III impurity radiation suggest that a thermal instability, as proposed by Rebut and Hugon, destabilizes the (m, n)=(3,1) and (4,1) islands, which grow during the current profile contraction phase. In contrast, the (2,1) islands appear to be thermally stable. Minor disruptions lead to step-wise loss of confinement, first localized at the 4 2 surface, then after a time delay comparable to the resistive time-scale, at the q=3 surface and after a second time delay, at the q=4 surface. It is found that (3,1) islands, unlike (2,1) islands, are quenched by the high heat flux during minor disruptions

Burn performance of fast ignited, tritium-poor ICF fuels

Abstract: The potentials of fuels that are predominantly deuterium, with a small tritium content (tritium-poor deuterium), are examined in the context of the inertial confinement fusion (ICF) scheme known as `fast ignitor', which is based on external heating of a portion of a precompressed target. For this purpose, the burn performance of compressed spheres of pure or nearly pure deuterium with an off-centre deuterium-tritium (DT) seed, ignited by an ultrapowerful external beam, has been studied using two dimensional (2-D) numerical simulations. The dependences of the fuel energy gain on the fuel mass and energy, and on the tritium fractional content FT, have been studied; limiting gain curves have been obtained for fixed values of FT and of the isentrope parameter alpha . It is found that assemblies with FT=0.5-1%, fuel mass m approximately=20 mg, compressed at density rho approximately=1000 g/cm3, can achieve tritium self-sufficiency (i.e. net tritium production). At alpha approximately=1.5, the corresponding fuel energy at ignition is of the order of 1 MJ and the fuel energy gain can be about 1000. The potentials of tritium-poor configurations are also compared with those of equimolar DT fuels with different ignition configurations

The role of rotation in tokamak internal transport barriers

Abstract: Internal transport barriers have been observed in several tokamak operating regimes and the reduced transport has been shown to be qualitatively consistent with E*B rotational shear suppression of turbulence. The interaction of the toroidal rotation with the diamagnetic and poloidal rotation contributions of the electric field is explored. Toroidal momentum injection counter to the current is predicted to lower the power threshold for improved confinement if the thermal ion density is peaked on-axis. Peaking of the density or power deposition profile near the magnetic axis is also found to lower the power threshold, in agreement with experimental trends

Experimental demonstration of tokamak fuelling by compact toroid injection

Abstract: The most promising concept for deep fuelling a reactor is by the injection of compact toroid (CT) plasmoids. The first results showing CT fuelling of a tokamak plasma, without any adverse perturbation to the tokamak discharge, are reported. The Compact Toroid Fueller (CTF) device was used to inject a CT-spheromak plasmoid into the TdeV tokamak. Following the CT penetration, the tokamak particle inventory increased by 16%, the loop voltage and the plasma current did not change, and there was no increase in magnetohydrodynamic (MHD) activity. The number of injected impurities was low and dominated by non-metallic elements. The plasma diamagnetic energy and the energy confinement time increased by more than 35%

Energy scaling of inertial confinement fusion targets for ignition and high gain

Abstract: The dependence of the ignition threshold on the velocity vimp and compressibility of an imploding fuel mass is central to establishing the driver requirements and implosion strategy for inertial confinement fusion (ICF). Using a series of LASNEX calculations, it is found that keimp varies as nu imp- alpha beta a, where keimp is the kinetic energy in the imploding fuel at the ignition threshold, alpha =5.5+or-0.5, a=1.7+or-0.2 and vimp is the implosion velocity. Here, the compressibility parameter beta is related to the pressure P and density rho of the DT fuel by the relation P= beta p53/. These results are obtained by starting at the peak implosion velocity for a fuel shell of a high gain ICF capsule and scaling the isentrope, mass and velocity of the fuel shell. In the presence of a mix of hot and cold material at the edge of the central hot spot, it is also found that the results can be fitted by assuming that the reduced clean fuel radius for a mixed capsule requires a velocity increase of the same magnitude as that which would be required if the entire capsule had been rescaled in size by the same ratio

Plasma convection induced by toroidal asymmetries of the divertor plates and gas puffing

Abstract: In the open field line region of the scrape-off layer (SOL), the plasma potential is to a considerable degree determined by the boundary conditions on the divertor plates. The idea is explored here that, by making the divertor plates slightly `wavy' in the toroidal direction, toroidally varying potentials can be produced throughout the SOL and thereby convective plasma motion induced. This motion should lead to a broadening of the SOL and to a reduction of the heat load on the divertor plates. An attractive feature of this technique is that it induces strong convection only on the open field lines without causing enhanced transport inside the separatrix. The possibility is also considered of inducing plasma convection by varying the composition of the divertor plate in the toroidal and radial directions, so as to affect the plasma potential by virtue of the toroidal variation of the secondary emission coefficient. Mutual biasing of adjacent elements of the plates is the third option. Finally, the use of toroidally asymmetric gas puffing to induce toroidal asymmetry in the potential distribution is discussed

The m=2, n=1 mode suppression by ECRH on the T-10 tokamak

Abstract: Experiments on m=2, n=1 tearing mode suppression and on avoidance of density limit disruptions by electron cyclotron resonance heating (ECRH) were performed on the T-10 tokamak. Partial suppression of the m=2, n=1 mode by the high frequency (HF) power deposition in the vicinity of the q=2 surface was observed. Development of external kink modes with HF power injection can result in m=2, n=1 mode destabilization under specific operating conditions. ECRH suppresses m=2, n=1 mode activity at extremely high values of electron densities and prevents the density limit disruptions practically independently of EC resonance position. Complete compensation of the additional peripheral heat losses near the density limit by ECRH should be responsible for this result. No effect of electron cyclotron current drive (ECCD) on m=2, n=1 mode stability has been observed because of insufficient values of HF driven current in the vicinity of the q=2 surface under the operating conditions of the experiment