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A unique sequence of 120 almost identical plasmas in the Joint European Torus (JET) recently provided two orders of magnitude more statistically equivalent data than ever previously available. The purpose was to study movement of eroded plasma-facing material from JET's new Beryllium wall, but it has allowed the statistical detection of otherwise unobservable phenomenon. This includes a sequence of resonant-like waiting times between edge-localised plasma instabilities (ELMs), instabilities that must be mitigated or avoided in large magnetically confined plasmas such as those planned for ITER. Here, we investigate the cause of this phenomenon, using the unprecedented quantity of data to produce a detailed picture of the plasma's behaviour. After combining the data, oscillations are clearly observable in the plasma's vertical position, in edge losses of ions, and in Beryllium II (527 nm) light emissions. The oscillations are unexpected, are not obvious in data from a single pulse alone, and are all clearly correlated with each other. They are likely to be caused by a small vertical oscillation that the plasma control system is not reacting to prevent, but a more complex explanation is possible. The clearly observable but unexpected link between small changes in the plasma's position and changes to edge-plasma transport and stability suggest that these characteristics cannot always be studied in isolation. It also suggests new opportunities for ELM mitigation and control that may exist.

Optimization of the higher order moments—specifically the squareness—of a tokamak cross-section can significantly enhance the stability to ideal magnetohydrodynamic ballooning and kink modes. At conventional aspect ratios, it is shown that access to the second regime of ballooning stability is facilitated by moderate squareness. In a low aspect ratio, fully bootstrap current driven spherical torus, optimization of the squareness results in an increase in β of the order of 10%, leading to a configuration stable to ballooning, axisymmetric, and ideal n⩽5 kink modes at β≃67%.

This article reviews the present status of research into controlled fusion using magnetic confinement of ionized gas or plasma. It concentrates on the world's leading magnetic configurations for this purpose, comprising the conventional tokamak, stellarator, reversed field pinch and spherical tokamak. These configurations are described in some detail, along with an introduction to the basics of this line of fusion research, the progress so far and the principal scientific and technical problems outstanding. The environmental implications are briefly addressed, in the context of some power generation alternatives and every-day radiation exposures.

For a two week period during the Joint European Torus (JET) 2012 experimental campaign, the same high confinement plasma was repeated 151 times. The dataset was analysed to produce a probability density function (pdf) for the waiting times between edge-localised plasma instabilities ("ELMS"). The result was entirely unexpected. Instead of a smooth single peaked pdf, a succession of 4-5 sharp maxima and minima uniformly separated by 7-8 millisecond intervals was found. Here we explore the causes of this newly observed phenomenon, and conclude that it is either due to a self-organised plasma phenomenon or an interaction between the plasma and a real-time control system. If the maxima are a result of "resonant" frequencies at which ELMs can be triggered more easily, then future ELM control techniques can, and probably will, use them. Either way, these results demand a deeper understanding of the ELMing process., Comment: 10 pages, 4 figures

One of the most promising approaches to achieving fusion of the light elements, as a potential large-scale energy source for the next century, is based on the magnetic confinement of an ionised high temperature plasma. Most of the current research in magnetic confinement makes use of toroidal plasma configurations in experiments known as tokamaks. Theoretical results have predicted that the characteristics of a tokamak plasma can be made more favourable to fusion if the cross-section of the plasma is appropriately shaped. However, the accurate generation of such plasmas, and the real-time control of their position and shape, represents a demanding problem involving the simultaneous adjustment of the currents through several control coils on time scales as short as a few tens of microseconds. In this paper, we present results from the first use of neural networks for the control of the high temperature plasma in a tokamak fusion experiment. This application requires the use of fast hardware, for which we have developed a fully parallel custom implementation of a multilayer perceptron, based on a hybrid of digital and analogue techniques. Our results demonstrate that the network is indeed capable of fast plasma control in accordance with the predictions of software simulations.

Electron cyclotron current drive experiments have been carried out in Compass-D using a power of approximately 500 kW at a frequency of 60 GHz. The fundamental resonance was used and the waves were launched from the high field side of the torus through four mirror-antennae. Significant asymmetry in the loop voltage behaviour is observed when comparing co and counter current drive cases, suggesting a driven current approximately 15 kA (in a plasma current of 130 kA). The BANDIT-3D Fokker Planck code has been used to model these discharges revealing a similar value for the predicted driven current at zero loop voltage but strong synergistic effects with the toroidal electric field at the approximately 0.4 Volts/turn typical of the experiments. This synergy appears to be sufficient to explain the inferred net 'co' current drive in most of the 'counter' current drive shots. In H-mode discharges featuring a strongly asymmetric single null separatrix configuration, discrete ELM's sometimes produce sufficient relaxation of the current profile to cause loss of vertical control, leading to vertical displacement events. Application of resonant magnetic perturbations (RMPs) of predominantly n=1 (with various low m values) is seen reliably to increase the frequency of the ELM's, although the exact mechanism underlying this effect has still to be elucidated. A real-time electronic neural network has been used for the first time for feedback control of the equilibrium in a tokamak. The neural network was used for feedback control of the plasma elongation throughout the discharge, while simultaneously monitoring the plasma vertical and horizontal position.

The poloidal distribution of the halo current density on the top dump plate in JET can now be measured thanks to a new set of Rogowskii coils. These are the first measurements in JET able to offer an insight in the width of the halo current interaction with the wall. Therefore they offer both validation of the assumption made for JET disruption design criteria and one additional point in the extrapolation of the expected halo current width, and hence halo current density (and related local electro-mechanical loads on in-vessel components) for ITER. During upward events, the measured current density is consistent with the measured total poloidal halo current. The halo footprint extends over most of the upper dump plate, converting to a halo current flux tube width of ∼100 mm. A set of four toridal field pick-up coils installed 90° apart now allows a more accurate measurement of the poloidal halo current, in particular its toroidal peaking factor, and direct comparison between halo and plasma asymmetries.

Global models are useful in the analysis of fusion reactors due to the facility in computing and presenting the results in terms of comprehensive parameters. A general figure of merit, which encompasses all the relevant tokamak parameters, is introduced in this paper by a convenient normalization of the global power balance equation. In this way, different hypothetical tokamak reactors can be compared in terms of their figure-of-merit value. This criterion is applied to analyse the performance of both ITER-like reactors and a class of newly proposed low-power reactors.

This paper covers two novel features used on the COMPASS experiment at Culham. The first part of the paper describes the function, design and performance of the toroidal field coil dynamic vertical pre-load device and the second part describes a new technique for accurately aligning the poloidal field coils using a fixed coil array.

Simple scaling laws for the ideal-MHD β-limit (βc) as a function of plasma shape parameters (inverse aspect ratio e, elongation κ, and triangularity δ) in the range covered by recent high-β (≤ 4.5%) experiments in Doublet III are derived. β-limits are obtained by optimizing the current profile. A large class of profiles is considered. β-limits are presented for a single Gaussian profile as they do not significantly improve with more elaborate profiles. Excluding the region dominated by sawtooth activity, experimental values do not exceed the n = ∞ ballooning mode β-limit, but do exceed the n = 1 kink mode limit if no wall stabilization is assumed, even in the presence of a cool mantle. If wall stabilization is assumed, the kink limit is above the ballooning limit when the safety factor at the plasma surface (qs) is greater than two. Even with a fairly close wall, the kink is still unstable when qs 2: βc(%) = 27 e1.3κ1.2 (1 + 1.5δ) for qs > 2 and βc = 0 for qs < 2. This expression shows that triangularity is almost as important as elongation to reach high β-values.

The CLEO device has been used to compare the confinement properties of a variety of ohmically heated toroidal configurations at the same magnetic field. These include RFP and OHTE configurations, conventional tokamaks, a novel helically assisted low-q tokamak, and an l = 3 stellarator. The plasma current and density vary over 2 orders of magnitude for the different configurations. The stellarator exhibits the best energy confinement time, tau/sub E/, but the tokamak achieves the best ..beta..tau/sub E/ product.

In current-free stellarators, the parallel current density is normally too weak to drive global external kink modes. However, at finite values of beta, the bootstrap current (BC) can provide sufficient free energy to trigger this class of mode in some stellarator systems. The effect of the BC in the collisionless 1/nu regime has been investigated in several different types of stellarator reactor systems all with a volume V similar to 1000 m(3). In quasiaxisymmetric and quasihelically symmetric stellarators, the BC is large at, finite beta and this can cause low order resonances to move into and emerge out of the plasma which in turn can destabilize global internal and external kink modes. In a six-field period system with poloidally closed contours of the magnetic field strength B, the BC is small and decreases the rotational transform only slightly. As a result, only intermediate to high n modes can become weakly destabilized. Furthermore, it is demonstrated in this system that the contours of the second adiabatic invariant J(parallel to) close poloidally for all trapped particles at finite beta* similar to 6%. This condition leads to the loss of a very small fraction of the collisionless alpha-particle orbits. In Sphellamak configurations with peaked toroidal currents required to generate nearly isodynamic maximum-B confining field structures, the BC accounts only for a small fraction of the total current. The loss of a-particles born within the inner quarter of the plasma volume is negligible while about 1/3 of those born at half volume escape the device within a slowing down time.