Table of contents

Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak.

Alfvén instabilities with rapidly decreasing frequency are observed in experiments with super-Alfvénic (up to V_beam/V_A ≅ 3) NBI in the spherical tokamak START (Culham Science Centre, UK). The unstable modes appear as high frequency, f₀ ≅ 70-400 kHz, `bursts' detected by magnetic pick-up coils and soft X ray cameras, and `sweep' down by a factor of two in frequency in a single 0.2-0.3 ms burst. A strong correlation between the initial frequency of the frequency sweeping modes f₀ and the parameter V_Aβ_T^-3/4, where V_A is the Alfvén velocity and β_T the toroidal beta, is found.

Tore Supra experiments are at present devoted to the study of high density regimes with radiofrequency heating. Recently, an improved L mode confinement regime has been observed in plasmas heated by ion cyclotron hydrogen minority heating, at relatively high densities up to 80% of the Greenwald limit. The quality of energy confinement is as good as that of ELMy H mode. The main physical mechanism of this regime has not been clearly identified. However, some features very similar to those of previous improved confinement modes using neutral beam heating in other tokamaks have been observed.

A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-45 (6 mm) deuterium pellets doped with a small (0.0005%) concentration of krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modelling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR.

Stationary helical fields with toroidal mode numbers n = 1 and n = 3 are applied to beam heated DIII-D plasmas. Measurements of the 14 MeV neutron emission monitor the confinement of the 1 MeV tritium fusion product. To within ∼15% uncertainty, static magnetic fields with vacuum amplitudes of δB/B ∼ O(10^-3) have no impact on fusion product confinement.

Approximate criteria for a Z pinch spark of an axial detonation in DT plasma are derived. On the basis of these, 2-D simulations are perfomed that suggest correct combinations of parameters for Z pinch sparks. The I_z current required for ignition is of the order of 10 MA and the radii of the sparks less than 10 μm. Such μZ pinches may be produced by a fast implosion of a cylindrical liner.

The prediction of the thermal energy confinement time, τ_th, for ITER size experiments is based on power law scalings obtained using data sets of present tokamak results in specific regimes, the most relevant being the ELMy H mode regime. A thorough statistical approach has provided a best fit to these data with an estimation of the error bars which forms the basis for the ITER EDA design parameters. In this article the range of variation of the main parameters in the database are studied individually and it is observed in particular that τ_th depends linearly on the combined variable aκB. Taking advantage of this linear dependence and of the dependence on the plasma current, it is shown that the four variables aκB, n, P* = P_L/V and q_eng, or equivalently I, , n and P*, are good parameters which provide a simple fit to the data, namely: τ_th = 0.0307 (aκB)n^1/2/(P*^2/3 q_eng) = 2π × 10^-3In^1/2/(P*^2/3), which satisfies the high-β or Kadomtsev constraint. This fit is as good as the best log-linear fit, using the eight variables I, B, R, , P, κ, M and n, over the full set of devices used in the databank and it yields the same prediction for the ITER EDA confinement time. It turns out that the simple best fit used is a gyro-Bohm scaling. It is also shown that small changes in the density, n and P* exponents can give a Bohm-like scaling, which is less accurate and pessimistic, but can be used as a lower bound prediction. The use of this simple scaling law is illustrated by a proposal to slightly reduce the size and magnetic field of the ITER EDA design, while taking advantage of the favourable dependence of τ_th on κ.

A computationally efficient modelling procedure for the calculation of neutral atom densities in the plasma edge is described and applied to calculate the recently measured neutral densities in DIII-D. The agreement of the calculated and measured neutral densities in the plasma edge inside the separatrix is quite good. The implications of these results for neutral transport in the plasma edge are discussed.

In the JET tokamak the perpendicular velocity distribution function of ICRF heated ions is inferred in the energy range 0.3 ⩽ E (MeV) ⩽ 1.5 from neutral particle analyser measurements of the emitted high energy atomic flux. The measurements are integrated along a vertical line of sight through the plasma centre, with a finite viewing solid angle. A method is presented to deduce the ICRF heated minority ion density in the plasma centre taking into full consideration the geometry of the measurements. The results are verified against independent measurements of fast ion density and perpendicular energy content, and quantitative agreement is obtained. These results further validate the model of impurity induced neutralization, which was developed to interpret the high energy atomic flux measurements in JET. They give access to fundamental properties of the ICRF heated ion distribution function, and provide improved tools for understanding the behaviour of fast ion populations in forthcoming fusion plasma experiments.

The interaction of plasma with the walls has been one of the critical issues in the development of fusion energy research. On the one hand, plasma induced erosion can seriously limit the lifetime of the wall components, while, on the other hand, eroded particles can be transported into the core plasma where they lead to dilution of the fusion plasma and to energy losses due to radiation. Low-Z wall materials induce only small radiation losses in the plasma core but suffer from large physical sputtering rates. Carbon based materials in addition suffer from chemically induced erosion. High-Z wall materials show significantly smaller erosion but lead to large radiation losses. One of the main goals of present plasma-wall studies is to find a special choice of wall materials for steady state plasma scenarios that will provide an optimum with respect to fuel dilution, radiation losses, wall lifetime and fuel inventory in the walls. To obtain a better understanding of the processes and to estimate the plasma-wall interaction behaviour in future fusion devices the 3-D Monte Carlo code ERO-TEXTOR, based originally on the ERO code, has been developed. It models the plasma-wall interaction and transport processes in the vicinity of a surface positioned in the boundary layer of TEXTOR. The main aim is to simulate the erosion and redeposition behaviour of different wall materials under various plasma conditions and to compare this with experimental results. This contribution describes the main features of the ERO-TEXTOR code and gives some examples of simulation calculations to illustrate the application of the code.

Many theoretical models show a direct connection between energy transport and particle transport. For a complete understanding of transport processes both energy and particle transport must be understood. A discussion is presented of how energy and particle transport should be related in terms of the relative importance of the diagonal and off-diagonal terms in the equation for the fluxes. A model for particle transport is discussed and it is shown that this model can describe measured density profiles in the DIII-D tokamak under a wide range of DIII-D parameters. This model is obtained from a Lagrangian formulation of the kinetic equation and utilizes the assumption that transport takes place while approximately conserving the first and second adiabatic invariants. The measured results utilize the improved diagnostic capability of DIII-D with an improved capability of measuring current density profiles and particle density profiles. This model is then extended to include energy transport. It is experimentally observed that the particle diffusivity and the thermal diffusivity do not differ greatly and have roughly the same radial dependence. This is discussed and compared with the model. A brief discussion of the effect of internal transport barriers is included.