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The JET scrape-off layer has been characterized with a reciprocating probe in inner wall, IW, and outer wall, OW, limited plasmas. Broad SOL profiles are observed for IW limited plasmas with power e-folding length substantially larger (by a factor of ∼5-7.5) than in OW limited plasmas. The properties of the fluctuations in the SOL parameters indicate larger turbulent transport for IW limited plasmas. The striking differences observed between IW and OW limited plasmas on the power e-folding length, parallel flow, turbulent transport as well as the characteristics of the fluctuations support the existence of a poloidally localized region of enhanced radial transport near the outboard midplane. The dependence of the SOL power e-folding length on the main plasma parameters was also investigated for IW limited plasmas and a modest negative dependence on both the plasma current and the line-averaged density found. © 2013 Euratom. Published by Elsevier B.V. All rights reserved.

JET carbon screening experiments were performed using methane gas injection. L-Mode experiments scanned parameters influencing the JET scrape-off-layer (SOL) and/or intrinsic impurity level. Scaling relations are derived to describe methane injected into L-Mode plasmas from the JET horizontal mid-plane. L-Mode screening was 3–20 times better for plasmas connected to the divertor than for similar limited plasmas. The screening was worse for methane injection from the mid-plane and best for injection from the divertor. The screening was 1.5–2 times worse for H-Mode than L-Mode. Both ELM-averaged and inter-ELM H-Mode screening was documented. The screening results were used to understand the intrinsic impurity levels. Zeff reduced at higher densities partly due to better carbon screening at the higher density, and partly due to decreased carbon influxes. Diverted L-Mode intrinsic carbon levels arose from both main chamber and divertor sources, while H-mode carbon primarily originated from the divertor. DIVIMP and EDGE2D were used to model the observed screening. The modelling indicated that carbon removal to the divertor required lower temperatures for Coulomb collisions to couple the impurity ions to the SOL deuterium flows. The carbon removal occurred primarily in the outer SOL regions.

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 dynamical interplay between turbulent transport, density gradients and radial electric fields has been investigated in the plasma boundary of magnetic confined plasmas. Experimental results show that there is a dynamical relationship between gradients and turbulent transport. Fluctuations on gradients, which turns out to have a rather Gaussian probability density function (PDF), and turbulent transport, which has a non-Gaussian PDF, are strongly coupled. Also, the size of turbulent events increases when the plasma deviates from the average gradient. This increase is particularly significant when the gradient increases above its average value. The system relaxes to the most probable state in a time comparable to the correlation time of turbulence. The present results emphasize the importance of the characterization of fluctuations in the terms of PDFs to test critically edge turbulence models.

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.

Hydrocarbon injection into the JET gas box divertor in ohmic and L-mode discharges has been used to investigate the intrinsic hydrocarbon erosion yields and to determine the variation of the photon efficiency with the edge ion flux. Constant methane erosion yields were measured (5% for CD 4 , 3% for CH 4 ) with no sign of any flux dependence. The erosion yield of ethane/ethene was lower but increased with plasma density (increasing edge ion flux, decreasing edge electron temperature) such that it dominated the molecular carbon source at high density. The measured photon efficiencies (the number of molecular dissociations per emitted photon) were not constant, but were found to increase with increasing density (increasing edge ion flux, decreasing edge electron temperature).

High density, high radiated power fraction and small ELMs are key elements of the current ITER design. In JET, these conditions are shown to be associated with high ELM frequency, low pedestal pressure and correspondingly reduced global energy confinement time. The article reviews the current understanding of the connections between these parameters and the mechanisms for the saturation of the density with gas fuelling.

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.

For the design and the operation of large fusion devices a detailed understanding of the power exhaust processes is necessary. This paper will give an overview of the current research on the divertor power load mechanisms. The results shown are obtained in JET with the ITER like wall and ASDEX Upgrade with tungsten coated plasma facing components. The challenges of infrared thermography on an ITER like bulk tungsten divertor are presented. For the steady state heat load, the power fall-off length q in JET-ILW is compared to an empirical scaling found in JET and ASDEX Upgrade with carbon plasma facing components. A first attempt to scale the divertor broadening S in ASDEX Upgrade with tungsten plasma facing components is shown. The ELM duration tELM in JET-C and JET-ILW is compared. For similar pedestal conditions (Te,ped and ne,ped), similar ELM durations are found in JET-C and JET-ILW. For higher ne,ped at the same pedestal pressure pe,ped, longer ELM durations are found in JET-ILW. The pedestal pressure pe,ped is found to be a good qualifier for the ELM energy fluency in both JET-C and JET-ILW. Improved diagnostic capabilities reveal ELM substructures on the divertor target occurring a few milliseconds before the ELM crash.

The ITER-Like Wall recently installed in JET comprises solid beryllium limiters and a combination of bulk tungsten and tungsten-coated carbon fibre composite divertor tiles without active cooling. During a beryllium power handling qualification experiment performed in limiter configuration with 5MW NBI input power, accidental beryllium melt events, melt layer motion and splashing were observed locally on few beryllium limiters at the plasma contact areas. The Lorentz force is responsible for the observed melt layer movement. To move liquid beryllium against the gravity force, the current flowing from the plasma perpendicularly to the limiter surface must be higher than 6kA/m2. The thermo-emission current at the melting point of beryllium is much lower. The upward motion of the liquid beryllium against gravity can be due to a combination of the Lorentz force from the secondary electron emission and plasma pressure force.

The first priority for the 2013 JET campaigns will be to provide relevant data for the ITER decision on the day one divertor material selection. In particular, a tungsten melt experiment using a specially modified divertor module will study the characteristics and impact of ELM induced melting on plasma operation ­ only JET has large enough ELMs to perform such a test. The second priority is to better understand why the change to an all metal wall has restricted access to scenarios with high normalised energy confinement (H-factor). This unexpected result may be partly due to the impact of the higher fuelling needed to control tungsten levels but in some scenarios there is also an effect related to the loss of a beneficial effect of carbon as a plasma impurity. In parallel with pushing JET's core plasma parameters as close as possible to those in ITER, we also intend to develop ITER demonstration pulses including as many of ITER's requirements as possible but not focusing on core performance. Another high priority item is the study of the generation and mitigation of runaway electrons which are so far not created in JET disruptions but are potentially very challenging for JET and ITER. Finally, there is still important work to be continued on key plasma-wall interaction issues ranging from material migration to fuel removal studies.

R. Neu, S. Brezinsek, M. Beurskens, V. Bobkov, P. de Vries, C. Giroud, E. Joffrin, A. Kallenbach, G.F. Matthews, M.-L. Mayoral, G. Pautasso, Th. Puetterich, F. Ryter, J. Schweinzer, ASDEX Upgrade Team and JET EFDA contributors ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs in both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically the Be PFCs of JET led to a very small low-Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by co-deposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N2) (AUG and JET) or by combining N2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fuelling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however in both devices theWcontent is increased during ICRH operation most probably due to increasedW sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not yet clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N2 seeding observed in both devices hints to the fact, that low-Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.

Disruptions are a critical issue for ITER because of the high thermal and magnetic energies that are released on short time scales, which results in extreme forces and heat loads. The choice of material of the plasma facing components (PFCs) can have significant impact on the loads that arise during a disruption. With the ITER-like wall (ILW) in JET made of beryllium in the main chamber and tungsten in the divertor, the main finding is a low fraction of radiation. This has dropped significantly with the ILW from 50­100% of the total energy being dissipated during disruptions in CFC wall plasmas, to less than 50% on average and down to just 10% for VDEs. All other changes in disruption properties and loads are consequences of this low radiation: long current quenches, high vessel forces caused by halo currents and toroidal current asymmetries as well as severe heat loads. Temperatures close to the melting limit have been locally observed on upper first wall structures during deliberate VDE and even at plasma currents as low as 1.5MA and thermal energy of about 1.5MJ only. A high radiation fraction can be regained by massive injection of a mixture of 10%Ar with 90%D2. This accelerates the current quench thus reducing the halo current and sideways impulse. The temperature of PFCs stays below 400oC. MGI is now a mandatory tool to mitigate disruptions in closed-loop operation for currents at and above 2.5MA in JET.

A complete global balance for material transport in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust in the divertor region. Following the end of the first JET ITER-Like Wall campaign a set of tiles has been removed from the main chamber and the divertor. This paper describes the initial tile surface profiling results for evaluating the erosion in the main chamber and deposition in the divertor. Tile profiling was performed on upper dump plate tiles, inner wall guard limiters made of beryllium and on inner divertor tiles made of tungsten coated carbon-fibre composites. Additionally, the mass of dust collected from the JET divertor is also reported. Present results are compared with JET-C campaign results with the all carbon wall.

Following the first JET ILW operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics have been removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases are drawn. The amount of deposition on inner and outer base divertor tiles and remote divertor corners has reduced by more than an order of magnitude. Deuterium retention ratio in mainly beryllium deposits has reduced by a factor of five and carbon concentrations are low. There is however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the midplane inner limiters is found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition.

This paper reports on the first post-mortem analyses of tiles removed from JET after the first campaigns with the ITER-like Wall (ILW) [1]. Tiles from the divertor have been analysed by the Ion Beam Analysis (IBA) techniques Rutherford Back-scattering (RBS) and Nuclear Reaction Analysis (NRA) and by SIMS to determine the amount of beryllium deposition and deuterium retention in the tiles exposed to the scrape-off layer. Films 10-20 microns thick were present at the top of Tile 1, but only very thin films (

G.P. Maddison, C. Giroud, B. Alper, G. Arnoux, I. Balboa, M.N.A. Beurskens, A. Boboc, S. Brezinsek, M. Brix, M. Clever, R. Coelho, J.W. Coenen, I. Coffey, P.C. da Silva Aresta Belo, S. Devaux, P. Devynck, T. Eich , R.C. Felton, J. Flanagan, L. Frassinetti, L. Garzotti, M. Groth, S. Jachmich, A. Järvinen, E. Joffrin, M.A.H. Kempenaars, U. Kruezi , K.D. Lawson, M. Lehnen, M.J. Leyland, Y. Liu, P.J. Lomas, C.G. Lowry, S. Marsen, G.F. Matthews, G.K. McCormick, A.G. Meigs, A.W. Morris, R. Neu, I.M. Nunes, M. Oberkofler, F.G. Rimini, S. Saarelma, B. Sieglin, A.C.C. Sips, A. Sirinelli , M.F. Stamp , G.J. van Rooij, D.J. Ward , M. Wischmeier and JET EFDA contributors The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER. Attention is focused upon a common high-triangularity, single-null divertor configuration at 2.5MA, q95 3.5 yielding the most robust all-C performance. Contrasting results between the alternative linings are found firstly in unseeded plasmas, for which purity is improved and intrinsic radiation reduced in the ITER-like wall (ILW) but normalised energy confinement is 30% lower than in all-C counterparts, owing to a commensurately lower (electron) pedestal temperature. Divertor recycling is also radically altered, with slower, inboard-outboard asymmetric transients at ELMs and spontaneous oscillations in between them. Secondly, nitrogen seeding elicits opposite responses in the ILW to all-C experience, tending to raise plasma density, reduce ELM frequency, and above all to recover (electron) pedestal pressure, hence global confinement, almost back to previous levels. A hitherto unrecognised role of light impurities in pedestal stability and dynamics is consequently suggested. Thirdly, while heat loads on the divertor outboard target between ELMs are successfully reduced in proportion to the radiative cooling and ELM recurrence effects of N in both wall environments, more surprisingly, average power ejected by ELMs and estimated scrape-off-layer pressure ratio during the approach to outboard detachment also both decline in the same proportion for the ILW. Finally, inter-ELM W sources in the ILW divertor tend to fall with N input, although core accumulation possibly due to net ELM erosion still leads to significantly less steady conditions than in all-C plasmas. This limitation of ILW H-modes so far will be readdressed in future campaigns to continue progress towards a fully-integrated scenario suitable for D-T experiments on JET and for "baseline" operation on ITER. The diverse changes in behaviour between all-C and ILW contexts demonstrate essentially the strong impact which boundary conditions and intrinsic impurities can have on tokamak-plasma states.

During the C28-C30 campaigns the ITER-like Wall (ILW) was periodically inspected by IVIS (InVessel Inspection System) without breaking the vacuum. After these campaigns the JET chamber was opened and the wall was inspected with a high resolution camera. The marker tiles were removed from the chamber and they have been inspected in the beryllium handling facility. The overall picture of the ILW after more than 3000 pulses looks good. However there are a few modifications such as deposition of Be on some W coated tiles, local destruction of the W coating by runway electrons in an emergency shutdown, arcing on particular W coated tiles, small delaminations on G6 and G8 divertor tiles. These aspects together with the results of high heat flux tests carried out with W coated CFC samples in Magnum-PSI are discussed in the paper.

The ITER-like wall (ILW) at JET is a unique opportunity to study the combination of material (beryllium and tungsten) that will be used for the plasma facing components (PFC) in ITER. Both the limiters (Be) and divertor (CFC W coated and bulk W) have been designed to maximise their power handling capability. During the last experimental campaign (October 2010 to July 2011) this capability has been assessed and even challenged in the case of the beryllium wall. The Be limiters are composed of highly shaped, castellated tiles. They are optimised for a range of scrape-off layer (SOL) power decay length 5 < lq,design < 20mm. Its power handling capability (19MW/m2s1/2), predicted with a simple model, has been proven to be robust by the experiments despite unexpected power loads pattern. However, this capability has been pushed to its limit leading to Be melt events, which revealed that the power load is toroidally asymmetric. The toroidal asymmetry could be explained by a misalignment of the limiters by 1mm. The protection system of the ILW has now been partially commissioned and the main chamber protection was operational. It did not prevent melt events mainly because the protection. The bulk W divertor target performed as predicted. Operations were constrained by: 1) an energy load limit of 60MJ per stack ( 60MJ/m2), and 2) the limited number of cycles of the surface temperature above 1200oC in order to prevent thermal fatigue. This latter limit has been exceeded about 300 times and no signs of damage or thermal fatigue have been observed by the photogrammetric survey.

Transient impurity events leading to unexpected increase in radiated power have been studied in JET from the installation of the ITER-like wall. A total of 1800 events over 2800 plasma discharges have been detected. None have led to permanent changes in plasma conditions. Of all events 60% show traces of W, 25% of either Ni, Fe or Cr from either Inconel or steel structures. They occur mainly in diverted magnetic configuration, independently of strike-point position. The effect of disruptions on dust redistribution has been investigated using the Thomson scattering diagnostic and correlated with TIE occurrence. The number of dust events detected increases with disruption force and, in comparison to the full-C wall, the amount of dust mobilized is found to be about an order of magnitude lower. Their time evolution correlates well with that of the transient impurity events.

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.

In this paper an experimental investigation of asymmetries in the DITE plasma boundary will be presented for a limiter configuration consisting of a pumped limiter centred on the outside midplane and an inner bumper limiter displaced ~ 70 ° toroidally. A number of experiments have been performed in which the poloidal variation of the edge T e , n e and j sat has been monitored as a function, principally, of the direction of plasma current and toroidal magnetic field. Poloidal asymmetries in particle flux are observed with the measured flux increasing towards the outside midplane. The poloidal variation of T e is less marked. Considerable e-i side asymmetries are evident in the particle flux which, when interpreted as indicative of net plasma flow, give poloidally varying Mach numbers and flow directions dependent on the toroidal magnetic field direction and line-averaged central density. The persistence of the observed asymmetries when the magnetic field and plasma current directions are reversed show the effect to be related to plasma properties and not simply a consequence of varying connection lengths.

The edge plasma response to Lower Hybrid Current Drive (LHCD) and Electron Cyclotron Heating (ECH) has been studied in DITE with input powers of up to 250 and 300 kW, respectively. During LHCD significant reductions in ne(edge), Te(edge) and the scrape-off layer (SOL) particle flux are observed. These effects, together with an observed decrease in the fluctuation levels E ~ θ and ne, and in the Hα emission, are strong evidence for an increase in particle confinement. During ECH precisely the opposite occurs and particle confinement decreases. Measurements have also been made during Lower Hybrid Heating (LHH). Global power balance has been studied by comparing the power conducted and convected to the edge, computed using Langmuir probes in the limiter, with the total power input less the radiated power. Reasonably good power accounting is observed in ohmic, ECRH and LHH phases but, during LHCD, virtually none of the injected RF power is observed in that reaching the boundary. This is consistent with a very local power loss and/or power loss via a high-energy component in the electron distribution which would not be registered in the Langmuir probe bias range (—100 to ~ 30 V).

The transport of carbon and oxygen impurities has been studied in TEXTOR by introducing the gases CH, and CO through a small hole in a test limiter. The toroidal distributions of different charge states of the impurities have been measured using a CCD camera with optical filters. Local impurity ion temperatures have been calculated from the Doppler broadening of line emission measured with a high resolution spectrometer. The spatial distributions and the ion temperatures have been modelled using the LIM Monte Carlo impurity code, with experimentally measured plasma profiles. Good agreement is obtained for both sets of measurements. The comparison shows the breakup energies of the atomic fragments to be < 1 eV. The fuelling efficiency of different gas species is discussed.

Using a new application of the retarding field analyser (RFA), it is possible to measure, in-situ, the coefficients of secondary electron release from chosen, conducting target substrates. Since the measurements are made in the tokamak, the yields are due to primary particles characterized by velocity distributions which are close to those experienced by solid surfaces intercepting the edge plasma. Operating the RFA in the standard mode also permits measurement, under identical discharge conditions, of Ti, Te, ne and Vsheath at the probe. Measurements of the coefficients of ion and electron secondary electron emission (δi, δe) are presented for both graphite and molybdenum target substrates. For electron energies in the range 0 to ~60 eV, values of δe near to unity have been measured for Mo and C with the graphite yields being slightly lower than those for molybdenum. δi is found to be 0.3–0.5 for molybdenum, in good agreement with ion beam measurements at low energy (~300–500 eV).

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137-2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Material migration has received renewed interest due to tritium retention associated with carbon transport to remote vessel locations. Those results influence the desirability of carbon usage on ITER. Subsequently, additional experiments have been performed, including tracer experiments attempting to identify material migration from specific locations. In this paper, EDGE2D models a well-diagnosed JET{sup 13}C tracer migration experiment. The role of SOL flows upon the migration patterns is identified.

Scaling to larger tokamaks of high confinement plasmas with radiating edges, induced by impurities, is being studied through internationally collaborative experiments on JET. In campaigns till the end of 2000, three different regimes have been explored. A small number of limiter L-mode discharges seeded with neon have most closely repeated the approach used on TEXTOR-94, but different collisionality and particle transport in JET impede central peaking of the density associated with improved confinement. Divertor L-modes at intermediate density, again with neon injection, have pursued transiently enhanced states found on DIII-D. Confinement up to H-mode quality, together with radiation fractions of approximate to40%, have briefly been obtained, though central Z(eff) quickly increases. Most effectively, neon and argon seeding of higher density ELMy H-modes formed mainly at low triangularity on the septum of the MkIIGB divertor, resembling a pumped-limiter arrangement, have been examined. Good confinement has been sustained at densities close to the Greenwald level in 'afterpuff' (AP) phases following the end of main gas fuelling, for little change of central Z(eff) but up to approximate to60% radiation. Outstanding normalized properties up to H-97 = 0.99 at f(Gwd) = 0.94 have thus been achieved, above the conventional H-mode density limit for diverted plasmas. Stationarity of states has also been extended to many energy confinement times by including low, extra gas inputs in the 'AP', suggestive of an optimized fuelling scheme. Further development in 2001 is reported separately in [1]. Accompanying ELMs are generally reduced in frequency though not evidently in size, electron pedestal pressure being almost unchanged from unseeded behaviour. There are indications of the most favourable impurity species scaling with plasma parameters, performance, radiation and its concentration within a mantle all increasing with argon compared to neon in JET. These benefits in terms of integrated properties are just as required for long burning pulses in ITER, supporting its use of a radiating mantle to assist not only power exhaust but performance too. Impurity-seeded H-modes can therefore contribute directly to next-step scenario development.

The spatial dependence of the power transmission factor, d, associated with an ion-electron pair passing through the sheath at the DIII-D divertor plate has been determined by sweeping the edge plasma across Langmuir probe detectors. Our results show that d decreases from the classically expected value of eight near the low density edge of the scrape-off-layer plasma to values less than unity at the peak of the profile. 13 refs., 16 figs.

The injection of hydrocarbons into the JET Gas Box divertor in ohmic and L-mode discharges has been used to calibrate the intrinsic hydrocarbon erosion yields and to determine the variation of the photon efficiency with the edge ion flux. Methane erosion yields of 5% for CD4 and 3% for CH4 were measured, with no sign of any flux dependence. The erosion yield of ethane/ethene was lower but increased with the edge ion flux (increasing edge plasma density, decreasing edge electron temperature) such that it dominated the molecular carbon source at high density. The methane and ethane photon efficiencies (the number of molecular dissociations per emitted photon) were both not constant, but were found to increase with increasing density (increasing edge ion flux, decreasing edge electron temperature).

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.