Table of contents

For high density H modes with constant gas puff on ASDEX Upgrade a slow evolution of the density profile is observed with a time constant much larger than τ_E. While the edge density remains constant the central density increases as does the stored energy until sawteeth are lost and impurity influx sets in. This peaking process is only observed with NBI heating. Substituting half of the heating by central ICRH leads to completely flat profiles even though half of the original NBI particle fuelling remains. This behaviour, as well as the slow timescale of the peaking process with pure NBI heating, is successfully modelled with a pinch of the order of the neoclassical Ware pinch and a proportionality between the particle diffusion coefficient D and the heat conductivity χ. Such an assumption links D to the heat flux profile q_heat, since the temperature profile is observed to be stiff. Such a model implies flat density profiles for a centrally heated fusion reactor.

The enhanced current drive (ECD) efficiency mode, which is characterized by a spontaneous increase of current drive efficiency η_CD from (0.3-0.4) × 10¹⁹A/W m^-2 to (0.6-1.0) × 10¹⁹ A/W m^-2, is observed in long pure LHCD plasmas on TRIAM-1M. The energy confinement time is also improved due to the increase of line averaged electron density, and of the ion and electron temperatures. The current drive efficiency is proportional to the electron density. The transition to ECD mode occurs at a critical density, which depends slightly on the refractive index in the toroidal direction N_|| of the injected wave.

A new spherical tokamak, TST-2, was constructed at the University of Tokyo and started operation in September 1999. Reliable plasma initiation is achieved with typically 1 kW of ECH power at 2.45 GHz. Plasma currents of up to 90 kA and toroidal fields of up to 0.2 T have been achieved during the initial experimental campaign. The ion temperature is typically 100 eV. Internal reconnection events are often observed. The internal magnetic field measured at r/a = 2/3 indicated growth of fluctuations up to the fourth harmonic, suggesting the existence of modes with several different mode numbers. In the presence of a toroidal field and a vertical field with positive curvature, non-inductively driven currents of the order of 1 kA were observed with 1 kW of ECH power. The driven current increased with decreasing filling pressure, down to 4 × 10^-4Pa. A study of high harmonic fast wave excitation and propagation has begun. Initial results indicate highly efficient wave launching. The spatial distribution of the RF magnetic field was qualitatively consistent with the result of a full-wave calculation.

It has been shown that an m = 0 instability of a Z pinch carrying a current of the order of 10 MA with a rise time of less than 10 ns can generate a spark capable of igniting a fusion detonation in the adjacent DT plasma channel. A possible method for generating such currents, necessary for the implosion of an initial large radius, low temperature Z pinch, can be a radial implosion of a cylindrical fast liner. The problem has been addressed in previous publications without considering the role played by an initially impressed m = 0 perturbation, a mechanism indispensable for the generation of a spark. The liner-Z pinch dynamics can be solved at several levels of physical model completeness. The first corresponds to a zero dimensional model in which the liner has a given mass per unit length and a zero thickness, the plasma is compressed adiabatically and is isotropic, and there are no energy losses or Joule heating. The second level is one dimensional. The Z pinch plasma is described by the full set of MHD, two-fluid equations. The liner is treated first as thin and incompressible, and subsequently it is assumed that it has a finite thickness and is composed of a heavy ion plasma, having an artificial but realistic equation of state. Both plasma and liner are considered uniform in the Z direction and only DT reactions are considered. It is shown that, given sufficient energy and speed of the liner, the Z pinch can reach a volume ignition. The third level is two dimensional. Plasma and liner are treated as in the second level but either the Z pinch or the liner is perturbed by an m = 0 non-uniformity. Provided the liner energy is high enough and the initial m = 0 perturbation is correctly chosen, the final neck plasma can act as a spark for DT ignition. It is also shown that the liner energy required for generating a spark and the subsequent detonation propagation are considerably less than in the case of volume ignition.

Experiments in JET have concentrated on steady state discharges with internal transport barriers (ITBs). The ITBs are formed during the current rise phase of the discharge with low magnetic shear ( = r/q(dq/dr)) in the centre and with high additional heating power. In order to achieve stability against disruptions at high pressure peaking, which is typical for ITB discharges, the pressure profile can be broadened with an H mode transport barrier at the edge of the plasma. However, the strong increase in edge pressure during an ELM free H mode weakens the ITB owing to a reduction of the rotational shear and pressure gradient at the ITB location. In addition, type I ELM activity during the H mode phase leads to a collapse of the ITB with the input powers available in JET (up to 28 MW). The best ITB discharges are obtained with input power control to reduce the core pressure, and with the edge pressure of the plasma controlled by argon gas dosing. These discharges achieve steady conditions for several energy confinement times (τ_E) with H97 confinement enhancement factors (τ_E/τ_{E,ITER97 scaling}) of 1.2-1.6 at line averaged densities of around 30-40% of the Greenwald density. Increasing the density by using additional deuterium gas dosing or shallow pellet fuelling leads to a weakening of the ITB. In order to sustain ITBs at higher densities, type III ELMs should be maintained at the plasma edge, giving scope for future experiments in JET.

The plasma current profile driven by coaxial direct current helicity injection in a low aspect ratio toroidal configuration is investigated by applying the principle of minimum energy dissipation rate. It is shown that current profile modes are mainly determined by the Lagrange multiplier β. Some critical values β_c are found. Different current profiles are obtained in different β ranges. Three typical current profiles are presented. The key features of the first case with β< 7.1 agree well with current experiments. Larger driven plasma current and the typical current profiles of normal tokamaks can be obtained in the region of β from 7.1 to 9.65. There exist reversals of both j_φ and B_φ in the central part of the plasma when β becomes higher than β_c ≈ 9.65. For a selected geometry the values of β_c depend weakly on other parameters. The different β values and corresponding current profiles can be achieved by adjusting certain parameters. There exist critical values of plasma temperature, bias voltage and vacuum toroidal magnetic field that induce the transformation of the current profile mode. The profile of λ = μ₀j_φ/B_φ is non-uniform in the plasma and a larger deviation from force free states appears when β becomes higher.

The current drive capability of negative ion based neutral beam injection (N-NBI) in JT-60U has been extended to the reactor relevant regime. The driven current profile and current drive efficiency have been evaluated in a high electron temperature regime T_e(0) ≈ 10 keV, and reasonable agreement with the theoretical prediction has been confirmed in this regime. The N-NB driven current reached 1 MA with an injection power of 3.75 MW at a beam energy of 360 keV. A current drive efficiency of 1.55 × 10¹⁹A m^-2 W^-1, approaching the ITER requirement, was achieved in the high β_p H mode plasma with T_e(0) ≈ 13 keV. This current drive performance permitted sustainment of a high beta (β_N = 2.5) and high confinement (HH_y2 = 1.4) plasma in the full current driven condition at a plasma current of 1.5 MA. The influence of instabilities on the N-NBI current drive was studied. When a burst-like instability driven by N-NBI occurred in the central region, reductions in loop voltage near the magnetic axis and in the neutron production rate due to loss of beam ions were observed although the lost driven current was at most ∼7% of the total driven current. When a neoclassical tearing instability appeared in high beta plasmas, the loss of beam ions was enhanced with increasing instability activity.

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5τ_E with q_min>1.5. (The normalized performance is measured by the product β_N H₈₉, indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite q_min > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈ 75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with β_N H₈₉ ≈ 7 for up to 6.3 s or ≈ 34τ_E. These discharges appear to have stationary current profiles with q_min ≈ 1.05, in agreement with the current profile relaxation time ≈ 1.8 s.

Research in NSTX has been conducted to establish spherical torus plasmas to be used for high β, auxiliary heated experiments. This device has a major radius R₀ = 0.86 m and a midplane halfwidth of 0.7 m. It has been operated with toroidal magnetic field B₀ ⩽ 0.3 T and I_p ⩽ 1.0 MA. The evolution of the plasma equilibrium is analysed between discharges with an automated version of the EFIT code. Limiter, double null and lower single null diverted configurations have been sustained for several energy confinement times. The plasma stored energy reached 92 kJ (β_t = 17.8%) with neutral beam heating. A plasma elongation in the range 1.6 ⩽ κ ⩽ 2.0 and a triangularity in the range 0.25 ⩽ δ ⩽ 0.45 have been sustained, with values of κ = 2.6 and δ = 0.6 being reached transiently. The reconstructed magnetic signals are fitted to the corresponding measured values with low errors. Aspects of the plasma boundary, pressure and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events, characterized by Mirnov oscillations, and a perturbation of the I_p, κ and l_i evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

High density plasmas (n₀ ≈ 8 × 10²⁰m^-3) featuring steady improved core confinement have been obtained in FTU up to the maximum nominal toroidal field (8 T) by deep multiple pellet injection. These plasmas also feature high purity efficient electron-ion coupling and peaked density profiles sustained for several confinement times. Neutron yields in excess of 1 × 10¹³ n/s are measured, consistent with the reduction of the ion transport to neoclassical levels.

Soft β limiting phenomena have been observed in T-10 in ECRH heated plasmas. Neoclassical tearing modes are supposed to be responsible for the β limitation. MHD onset was observed at high β_p values but low β_N values. The critical β has been found to be almost independent of the collisionality parameter ν_e*. Sawtooth stabilization by ECCD does not result in an increase of critical beta. A dependence of the critical β on the q(r) profile (modified by ECCD) has been observed.

Efforts have been made on the HT-7 tokamak to extend the stable operation boundaries. Extensive RF boronization and siliconization have been used and a wider operational Hugill diagram has been obtained. The transit density reached 1.3 times the Greenwald density limit in ohmic discharges. A stationary high performance discharge with q_a = 2.1 has been obtained after siliconization. Confinement improvement was obtained as a result of the significant reduction of electron thermal diffusivity χ_e in the outer region of the plasma. An improved confinement phase was also observed with LHCD in the density range of 70-120% of the Greenwald density limit. Off-axis LH wave power deposition was attributed to the weak hollow current density profile. Code simulations and measurements showed good agreement with the off-axis LH wave deposition. Supersonic molecular beam injection has been successfully used to achieve stable high density operation in the region of the Greenwald density limit.

The theoretical model of transport reduction by E × B shear decorrelation is tested experimentally for ASDEX Upgrade discharges. The radial force balance is used to determine the radial electric field from charge exchange recombination spectroscopy measurements. As the effective rate coefficient for photon emission of the charge exchange process depends on the collision energy, the alignment of the lines of sight with respect to the neutral beam gives rise to apparent velocities and temperatures. In addition, the gyro-motion of the observed species along with the finite lifetime of the observed excited state leads to lineshifts in spectra measured in the poloidal direction. Both effects require corrections, which will be discussed. The corrections are tested using the measurements of a discharge with a locked mode. From the profiles of an H mode discharge with improved confinement and a discharge with an internal transport barrier (ITB) the ion heat transport coefficients, E × B shearing rates and the maximum linear growth rates of the instabilities are calculated. Comparison of these results supports the assumption that turbulent transport due to the ion temperature gradient instability is suppressed inside the transport barrier in the ITB discharge. However, due to the dominant influence of the toroidal rotation velocity on the central E × B shear, they do not prove the shear decorrelation model, because E_r naturally rises during the phases with improved confinement if unbalanced neutral beam heating is applied.

Modulation of the plasma current has been used successfully to suppress MHD activity. This was achieved in discharges near the density limit where large MHD m = 2 tearing modes were suppressed by sufficiently large plasma current oscillations. The modulation of the plasma current must be large enough to move the resonance q = 2 surface outside the island width on a timescale faster than the growth time of the instability. These observations resemble those predicted in a previous study of non-linear effects on the m = 1 mode.

In order to mitigate the effect of disruptions in tokamaks, it is proposed to inject quickly a relatively large amount of helium; first experiments on this topic have been performed on TEXTOR. For this purpose, a fast valve has been developed which releases 10 mbar L of helium gas within 1 ms; the valve is located at a vessel flange such that a fast response is guaranteed even if it is triggered at the onset of the disruption. The amount of gas is sufficient to exceed the density limit even with low density discharges. The intention of the proposal is to shorten the plasma current decay phase, to reduce halo currents, to suppress runaway electrons and to provide good conditions for the start of the following discharge. In particular, for achieving the last goal, helium is the optimum choice of all the elements. The experiments performed on TEXTOR have proven various of these mitigation aspects: the current decay time is shortened, runaway electrons are expelled by the gas puff and the conditions for the start of the next discharge have neither deteriorated with respect to gas release from wall components nor with respect to excessive impurity production.

The effect of plasma shaping (triangularity and elongation) on sawtooth stability is addressed as well as the technique of current profile broadening using off-axis ECH to enlarge the operational range towards higher elongations. The plasma shape strongly influences the sawtooth period and amplitude. This effect is emphasized by ECH, with the sawtooth period becoming shorter at low triangularity or at high elongation; for these plasma shapes, the pressure profile inside the q = 1 surface remains essentially flat throughout the sawtooth cycle. A comparison of the sawtooth response with marginal Mercier stability shows that the critical pressure gradient at q = 1 is particularly low for plasma shapes where the increased sawtooth repetition frequency prevents the peaking of the pressure profiles. For these shapes, stability calculations show that the ideal internal kink is also unstable. The stability of highly elongated plasmas depends largely on the current profiles. The operational range at low currents has been extended towards higher elongations using ECH discharges. Far off-axis second harmonic X mode ECH power deposition proves to be an efficient tool for current profile tailoring, allowing a significant elongation increase at constant quadrupole field.

Temperature measurements in tokamak edge plasmas suffer frequently from outliers of unknown origin. Such outliers have an important unwanted influence on the estimation of parameters for edge temperature model functions in conventional least squares fits. Bayesian probability theory is applied to deal with such outliers and develop a robust procedure which performs highly satisfactorily.

The interactions of a compact toroid (CT) plasma with an external magnetic field and a tokamak plasma have been studied experimentally on the FACT and JFT-2M devices. Fast framing camera and soft X ray emission profile measurements indicate shift and/or reflection motions of the CT plasma. New electrostatic probe measurements indicate that the CT plasma reaches at least up to the separatrix for discharges with toroidal field strengths of 1.0-1.4 T and that there exists a trailing plasma behind the CT. A large amplitude fluctuation on the ion saturation current and magnetic coil signals is observed. Power spectrum analysis suggests that this fluctuation is related to magnetic reconnection between the CT plasmoid and the toroidal field. The CT, including much of the trailing plasma, may be able to move across the external magnetic field more easily in the drift region of the injector owing to the Hall effect.

A 2-D numerical simulation of the edge plasma in ASDEX Upgrade is used to show that above a critical density a bifurcation occurs, leading to strong poloidal in-out asymmetry of the electron temperature in the divertor. At densities well below and above the critical one, the asymmetry is less pronounced. Although drift terms are not included in the model, these findings can explain experimental observations from Alcator C-Mod and ASDEX Upgrade: (a) the clamping of the temperature in the inner divertor after the bifurcation and (b) the large temperature asymmetry immediately above the critical density. The position of the colder divertor leg is determined by a small asymmetry which is amplified by the bifurcation. In the simulation the basic asymmetry is due to different lengths of the flux tubes in the plasma edge. In the experiment the asymmetry effect of different field line lengths is combined with the action of E × B drifts. Comparison of X point top and bottom measurements then produces evidence that the E × B drift effects are dominant. The effect of the difference in field line length is still present in the experiment but E × B produces a stronger effect.

A review of the LH current drive experiments carried out on the Tore Supra tokamak is presented. The issues for an effective application of the LH wave at high power in reactor relevant conditions are highlighted. A promising performance has been obtained with the new launcher that is designed to couple up to 4 MW during 1000 s at a power density of 25 MW m^-2. The heat load on the guard limiter of the antenna and the fast electron acceleration in the near electric field of the grill mouth remain at a low level, while the mean reflection coefficient never exceeds 10%. The powerful diagnosis capabilities of hard X ray (HXR) fast electron bremsstrahlung tomography has led to significant progress in the understanding of LH wave dynamics. The role of the fastest electrons driven by the LH wave is clearly identified. From HXR measurements, an increase of LH current drive efficiency with plasma current is predicted and confirmed by a direct determination at zero loop voltage. Lower hybrid power absorption is observed to be off-axis under almost all plasma conditions, and its radial width clearly depends on the antenna phasing conditions. A correlation between the HXR profiles and the onset of an improved core confinement is identified in fully non-inductive discharges. This regime ascribed to some vanishing of the magnetic shear is found to be transient and usually ends when the minimum of the safety factor becomes very close to 2, leading to a large MHD activity. Experimental observations and numerical simulations suggest that LH power is absorbed after a few transits across the plasma. However, besides toroidal mode coupling, additional mechanisms probably contribute to a spectral broadening of the LH wave.

Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. A novel technique is presented of early prediction of plasma disruption in tokamak reactors which uses neural networks and `fuzzy' inference. The studies carried out in the work make use of an experimental database of disruptive shots made available by the ASDEX Upgrade Team. The main result of the work is that, in the limit of the available database, it is possible to predict the onset of the disruptive event sufficiently in advance in order to put the control system into action. The proposed system is a modular scheme that exploits a decomposition of the original database carried out in a proper way.

The non-linear 3-D toroidal gyrokinetic simulation code PG3EQ is used to study toroidal ion temperature gradient (ITG) driven turbulence - a key cause of the anomalous transport that limits tokamak plasma performance. Surveys of χ_i versus E × B and toroidal shear show that (a) the maximum growth rate is not a good predictor of the E × B shear required to suppress turbulent transport, (b) there is often a `plateau region' in which E × B shear significantly reduces the maximum linear growth rate but not the transport and (c) the parallel velocity shear component of toroidal velocity shear can negate much of the transport reduction by the E × B shear. Simulations in which the ion temperature gradient T_i(r) initially varies with radial position evolve towards a state without strong radial variations in T_i(r) while approximately preserving the total (E × B + diamagnetic) ion flow. If the electrostatic potential is initialized to zero, then sheared E × B flows result which can significantly reduce or quench the transport. If, however, the total ion flow profile (or equivalently the radial force) is initialized to be radially uniform, then the initial radial variations in T_i(r) do not result in a significant reduction in the transport. Surveys of χ_i versus magnetic shear S show a (often very sharp) peak at S ≅ 0.5 - the value at which the orientation of the ITG modes is in the major radius direction and is independent of poloidal angle in the region near the outer midplane. Surveys of χ_i versus T_i show that in most cases the thermal flux Q (χ_i T_i) has a linear dependence on T_i, with an intercept value of T_i which is often significantly larger than than linear critical value.

Report on the 7th IAEA Technical Committee Meeting held at Cannes, France, 13-16 June 2000.