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A 2D-tomographic reconstruction of D α - and C III-emission distribution in L-mode, density limit discharge has been performed and compared with results from other diagnostics (KL2, KS3). The observed emission patterns in a poloidal plane show a strong in–out asymmetry, which depends on local plasma parameter in the divertor. It is shown that the major contribution to D α -emission, under the observed conditions, lies in excitation processes and that low-energy charge-exchange reactions with neutral deuterium can influence the ionisation balance of carbon and the recorded C III-emission line. The evolution of the intrinsic hydrogen and carbon impurity profiles has been analysed with respect to its dependence on the location of methane fuelling.

Computer simulations of a design for a plasma ion mass spectrometer (PIMS) head for diagnosing the edge plasma on the Joint European Torus (JET) magnetic fusion experiment are reported. An overview of the design issues is presented along with results of these fully three-dimensional ion optical calculations. By modelling the behaviour of the different impurity ions anticipated in JET, an evaluation of the proposed PIMS geometry has been carried out. The effects of misalignment of the electric and magnetic fields within the PIMS on mass resolution, and the effect of the range of ion temperature and sheath voltage expected in JET on the transmission are described.

The probability density function (PDF) of turbulent transport has been investigated in the plasma edge region of tokamak (JET) and stellarator (TJ-II) fusion devices. PDFs can be re-scaled using a functional form, PDF(ΓE×B) = L−1g(ΓE×B/L), where L is directly related with the level of fluctuations in the turbulent flux. This kind of re-scaling holds at different timescales in which the functional form of the PDF changes. The empirical similarity in the PDF of turbulent transport in the edge region in both the JET tokamak and the TJ-II stellarator supports the view that turbulent transport displays universality in fusion plasmas. These results emphasize the importance of the statistical description of transport processes in fusion plasmas as an alternative approach to the traditional way to characterize transport based on the computation of effective transport coefficients (i.e. diffusion coefficients) and on average quantities (i.e. average correlation lengths).

Deposition and H-isotope retention in JET is highly asymmetric, with deposition predominantly in the inner divertor, where flaking deposits form on water-cooled louvres shadowed from the plasma. The asymmetry implies drift in the SOL of JET from outboard to inboard under normal operating conditions, which has been measured. In order to model the amounts of deposition, assumptions have to be made about the transport at the inner target. Analysis of divertor tiles shows that material from the main chamber travels along the SOL to the inner divertor wall, from where carbon is preferentially removed leaving a beryllium-rich film. The carbon travels to shadowed areas (such as the louvres) where deposits with high H-isotope content accrue. The analysis indicates that chemical processes must be important.

The successful operation of a popup Langmuir probe system, which was installed in the JET MkIIa divertor, is described. The system utilises the ambient magnetic field in tokamak plasmas to act on a current carrying coil and pop up a rail containing Langmuir probes. Measurements were made using pin-plate probes which, owing to their relatively large exposed area, are ideally suited for use with such a system. Details of the design, testing, measurements and potential applications of JET’s popup system are given.

L-mode and H-mode density limits with the ITER-like wall (ILW) have been investigated in the recent experimental campaign and compared with experiments in the JET carbon material configuration. The density limit is up to 40% higher in the JET-ILW than in the JET-CFC machine. This is linked to the formerly higher radiation fraction and, correspondingly, to earlier divertor detachment in the JET-CFC. In the ILW configuration, the discharge demonstrates a stable operation with a completely detached outer divertor in L- and H-mode. In contrary to the well-known "heating power independent" Greenwald limit, the L-mode densities limit increases moderately with rising heating power (∼ P0.4 heat) independently of the wall material. The H-L transition constitutes an effective undisruptive density limit for an H-mode plasma. Detachment itself does not trigger the H-L back transition and does not present a limit on plasma density. In the range of neutral beam heating 8-10.5 MW, no dependence of the H-mode density limit on the heating power was observed. © 2013 Euratom. Published by Elsevier B.V. All rights reserved.

Plasma detachment from a divertor target is defined as the state in which large gradients in total plasma pressure (static plus dynamic) are observed parallel to the magnetic field with consequent reductions in the plasma power and ion fluxes to the limiting surfaces. The origins of this concept and the first experiments in a linear divertor simulator are described. Experimental evidence for this phenomenon from the JET, JT60-U, DIII-D, Alcator C-Mod, and ASDEX-U tokamaks is reviewed along witi recent advances in the theory and modelling of detached plasmas applicable to current experiments. Plasma detachment from limiters was first reported by TFTR in 1985, shortly followed by DITE and TEXTOR. This phenomenon occurs in low power discharges near the density limit and is characterised by a shrinking of the layer where most of the ionisation occurs away from the limiter. The physics, phenomenology and relevance to fusion of detachment from limiters and divertors are constrasted.

We show that charge-exchange and other ion-neutral collisions can reduce the power transmission factor of the plasma sheath, thereby lowering the ion impact energy and target plate sputtering. The power transmission factor relates the heat flux reaching the divertor target to the plasma density and temperature just in front of the surface: δ=Qsurf/Jewk Te. Experimental data from the DIII-D tokamak suggests that δ could be as low as 2–3 near the region of peak divertor particle flux, instead of the 7–8 expected from usual sheath theory. Several effects combine to allow ion-neutral interactions to be important in the divertor plasma sheath. The shallow angle of incidence of the magnetic field (1–3° in DIII-D) leads to the spatial extension of the sheath from approximately ρi∼1 mm normal to the plate [R. Chodura, Phys. Fluids 25 (1982) 1628] to several centimeters along the field lines. Ionization reduces the sheath potential, and collisions reduce the ion impact energy.

Recent advances in electrostatic particle analysers enable a detailed analysis of impurity and fuel ions in the tokamak boundary. The paper presents measurements of impurity content, charge state mix, ion temperature, ion energy distribution functions, sheath voltage and secondary electron yields obtained in the DITE tokamak. The consequences of these results for the understanding of impurity production processes are discussed.

In this paper, recent developments in ion diagnostic probes for tokamak boundary plasmas are discussed. Three areas are covered: retarding field analysers, sniffer probes and plasma ion mass spectrometers. The contribution of these diagnostics to our understanding of plasma surface interactions is summarised.

A new diagnostic, combining both electrical and surface collection techniques has been designed in an effort to characterize the SOL with respect to the charge and energy of impurity ions. The probe utilizes the tokamak magnetic field to obtain spatial profiles of impurity deposition due to the gyroradii of particles passing through small apertures. Measurements have been made of the spatial distributions of iron and carbon on high purity aluminium substrates. The deposited profiles are compared with the results of a simple 2-D Monte Carlo model leading to estimates of the impurity ion temperature and the determination of approximate charge state distributions.

This paper describes direct experimental investigation of the region of plasma perturbed by a probe. Two adjacent probes have been mounted on the DITE tokamak. The large “disturber” probe is a rectangular graphite plate that creates the main particle flow pattern and can be biased with respect to the torus. The smaller “search” Langmuir probe is located 0.5 m away on the same magnetic field line. It can be swept back and forth mechanically through the shadow of the main probe many times during each discharge. Through measurements of ion saturation current at many distinct radii, a complete map of the perturbed region has been built up and transformed into radial and poloidal coordinates. A numerical solution of the two-dimensional diffusion equations is compared with experimental data in the limiter shadow. Two zones of substantially different interconnection length are included in the model. The assumption of equal radial and poloidal cross-field diffusion coefficients (D⊥ = 0.18 m2s−1) is found to give a good fit with the data.

Measurements of secondary-electron-emission (SEE) yield are reported for samples of clean 5890 PT graphite and for the same material exposed for some months in the JET Tokamak. The samples exposed in the JET Tokamak give substantially higher values than the clean samples. Controlled experiments show that both implantation with H2+ ions and deposition of thin metal films increase the SEE yield. The experimental data have been integrated over a 3D Maxwellian distribution to obtain the values of effective SEE yield expected from a plasma as a function of electron temperature. From these SEE yields, theoretical values of sheath potential and sheath energy transmission factor have been calculated. The implications for ion sputtering and power loss are discussed.

During the past year electron cyclotron resonance heating (ECRH) has been applied for the first time to the DITE tokamak. A new edge diagnostic, operable both as a Langmuir probe and a retarding field analyser, has been used extensively to investigate the variation of plasma boundary parameters during ECRH heated discharges. Arrays of Langmuir probes embedded in a pumped and inner bumper limiter have also been operated during additionally heated discharges. Measurements have been made over a wide range of ECRH input powers, as a function of global plasma density and for many values of the toroidal magnetic field. The behaviour of the edge plasma in response to the additional heating is observed to be strongly dependent on the major radial position at which resonant absorption of the ECRH pulse occurs. Radial profiles of electron density and temperature have been obtained, together with observations of the time response of these edge parameters to the ECRH pulse. Characteristic lengths in the boundary are observed to remain approximately unchanged throughout heated discharges. In general, the edge particle flux exhibits a well defined, reproducible behaviour, increasing during ECRH to ~1.5 times its ohmic value with rise times ≤ 2 ms. The percentage increase in the edge temperature is smaller with rise times ≤ 10 ms. Reasonable global power balance is obtained using data from the limiter Langmuir probes in both ohmic and ECRH phases of a typical discharge.

Experimental measurements of the energy distributions of ions and electrons arriving at a surface in the boundary layer of a tokamak have been made with a retarding field analyser. The analyser uses a 9 μm entrance slit which is less than the Debye length for the plasma conditions investigated. The electron energy distributions are Maxwellian and the ion energy distributions are displaced Maxwellians, as expected. The sheath potential is derived from the ion energy distribution and found to be ∼2kTe, slowly decreasing with electron temperature. Measurements of secondary electron emission yield have been made on materials exposed in a tokamak as functions of incident electron energy and angle. This data have been integrated to obtain the effective yield of a full three-dimensional Maxwellian distribution. Using these results the theoretical sheath potentials have been calculated as functions of electron temperature and have been found in good agreement with the values that were measured directly.

For typical reactor conditions lower hybrid (LH) waves are a strong candidate for bulk heating and for profile control via current drive in the outer regions of the plasma. However, limited accessibility, strong electron damping and the need to avoid damping on fusion‐produced alpha particles probably exclude the use of such a scheme in the plasma core. Therefore, it is likely that lower hybrid heating/current drive in the outer regions of the plasma will be combined with either ECRF, ICRF or NBI in the core. Electron cyclotron (EC) waves are particularly attractive since they offer the possibility of localised absorption and in addition may be used to facilitate low voltage start‐up of the tokamak. Recent experimental and theoretical research on current drive at Culham has concentrated on some of the outstanding critical issues concerning the reactor applications of LH and EC waves. These include studies of the interaction of lower hybrid waves with energetic ions including alpha particles, the effect of...