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Two possible techniques for the measurement of erosion/redeposition in the new JET divertor have been assessed; colour fringe analysis (CFA) and speckle interferometry. CFA is a simple technique in which the target is viewed with a colour camera, and can study optically transparent deposited films up to 1 μm in thickness. Speckle interferometry is a powerful technique for following changes in surface topology, but its sensitivity of >1 μm makes it more ideally suited to the next generation of tokamaks.

An analytical model of the SOL/divertor magnetic geometry is applied to JET. Exponential decay lengths, λ, are related to differences in magnetic fluxes and are expressed in terms of λ at midplane. Consistent values of λ are usually obtained from Langmuir probes in the SOL or in the divertor, and with Lyman-α and Balmer-α profiles in the divertor. Scaling of λ is presented: it is only slightly affected by 〈ne〉, by X-point to target distance and by input power (other than the usual changes Ω⇒L⇒H); it increases strongly with BΦ, especially in H-mode.

The dependence of impurity production and retention on the divertor density, on the power flow into this region as well as on the X-point to target distance are investigated. Model predictions suggest a good impurity retention above a certain divertor (scrape-off) density threshold, which is dependent on heating power. In our experiments pre-programmed midplane or X-point gas puffs were used to scan the density, as well as to avoid the depletion of particles from the divertor and the scrape-off during H-modes. The gas puffs reduce T e and increase N e in particular at the outer strike zone. In general the Be as well as the C influx increases with density, which is understood from the T e ( T i ) dependence of the sputtering yields. The impurity retention shows the expected improvement with increasing scrape-off (divertor) density as well as with increasing X-point to target distance (connection length).

In this paper we describe the simulation of edge diagnostics in JET using the DIVIMP (divertor impurity) Monte Carlo code. We concentrate on two ohmic pulses and show how the results are influenced by a variety of modeling assumptions. Our results show that a wall source must be included to explain the diagnostic signals. The wall source is shown to be a significant source of impurity in the discharges studied and more generally.

A study of density control in JET has revealed that dynamic retention in the beryllium surface of the first wall is an important factor in controlling recycling. A self-consistent global particle balance model that simulates the recycling in JET has been devised. The assumption is that particle transport is effectively diffusive in the materials encompassing the plasma and particle release is by recombination to molecules. The effect of neutral particle screening in the plasma edge is also globally taken into account. Comparison with experimental results of discharges of normal duration (20 s) shows good agreement. Discrepancies appear for 60 s long discharges, suggesting changes of material properties.

Systematic analysis of currents passing between the two strike zones in the scrape-off layer of JET single-null divertor discharges has been performed. Using a set of Langmuir probes, stretched across the target tiles in poloidal direction, the relation between the asymmetry in plasma parameters at the strike zones and parallel currents has been established. It has been found that the direction of edge currents was entirely determined by the plasma density and electron temperature distributions at the two strike zones, with the thermoelectric current being the major contribution to the net edge current. However, to match the whole set of data, the existing theory of Staebler and Hinton (Nucl. Fusion 29 (1989) 1820), which accounts for the thermoelectric current, had to be extended to comprise the effect of electron pressure imbalance between the two strike zones. The remarkable phenomenon of pressure imbalance in favour of ion strike zone is discussed.

An edge database is presented for divertor discharges in JET with various configurations: single null upper/lower (SNU/SNL), double null (DN), Δ B drift towards or away from the target and various divertor target materials (Be and C). We present the scaling of the scrape-off layer (SOL) parameters (i.e. density, temperature and their radial profiles) with main plasma parameters for divertor discharges with: I p =3.2 MA, 409-1 and B t =2.2–3.4 T. Results are presented for ohmic, L-Mode and H-mode discharges with additional heating (NBI and ICRH) up to 10 MW. The parameters were measured in the main SOL and at the divertor plates. A comparison is made for discharges when gas puffing was used either at the midplane or at the X-point. The results are compared with similar scalings obtained in limiter plasmas and with the predictions by simple SOL models.

A comparison is presented between a 1D SOL/divertor model (EDGE1D) and results from a set of L mode, low power NI heated single null divertor experiments run on the Be target plates. Two gas puff scans were run, one with outer midplane puffing, and the other with divertor puffing. Divertor and SOL parameters were measured with Langmuir probes and a CCD camera. Very good agreement and consistency were found between measurements in the SOL, outer divertor, inner divertor, and the code predictions, for the deuterium plasma, when the code predictions of electron pressure ratio between outer divertor and SOL were used to help determine the precise location of the reciprocating probe relative to the separatrix. Such features as inner-outer divertor asymmetries, parallel temperature gradients, and Be influx from the targets were all correctly modelled, within the error bars, and detached plasmas were observed both experimentally and numerically. However, the Be content of the SOL appears to scale differently than predicted.

Deuterium recycling during and after ohmic limiter discharges in the Be-phase of JET are compared with that during the C-phase. For the Be-phase, effective transport parameters for diffusion and recombination of deuterium at surfaces of in-vessel components of JET are derived using plasma discharge data only. This and the analysis of the outgassing behaviour after the discharge suggest that deuterium release is controlled by recombination with D/k (D is the diffusion, k the recombination coefficient) of about 1021−1022 m−2, larger than expected from laboratory measurements.

Measurements of flux and electron temperature across the divertor region in JET have been made using Langmuir probes mounted in the target tile surfaces, a reciprocating probe in the scrape-off layer, and CCD camera observations. It is found that flux density radial profiles steepen in going from L to H mode, and this is observed as a fall in flux and Dα light along a chord at large radii. Fluxes integrated across the strike zone region of the tiles only exhibit a transient fall, however, and then an exponential rise back to the L-mode value. From these measurements particle confinement times in L and H mode have been calculated, they rise from typically 0.15 in the L-mode to 0.8 s in the H-mode. Calculations of the sputtering of the divertor target tiles show little difference between the two modes.

The rate of production of impurities by physical sputtering is discussed in terms of a self-consistent zero-dimensional model based on energy balance. The importance of the effective self-sputter yield is emphasised and the effect of the uncertainty in this parameter is examined. The model is used to calculate how the plasma edge temperature, impurity influx, impurity concentration and total radiation depend on plasma average density and total input power. These results are compared with experimental results obtained using graphite and beryllium limiters in JET.

The distribution of metal impurities on the surface of the graphite limiters in JET after exposure to discharges in the operational periods 1983, 1984 and 1985 have been examined. A simple model for erosion and redeposition is proposed which is in broad agreement with the observed distributions. It is concluded that impurity deposition on the limiter from all sources is rapidly eroded and redeposited on the limiter surface during normal tokamak discharges so that the final distribution is dominated by this mechanism. Mechanisms are also identified for net removal of material from the limiter.

Wall long term samples (LTS) which were exposed during the 1984 and 1985 operational periods of JET have been investigated by means of surface analysis techniques. The results indicate that deposition of wall material (Fe, Cr, Ni) and its distribution on the vessel wall were strongly affected by wall erosion on bellows protection plates during unstable discharge phases. Erosion sources and deposition sinks are spatially correlated indicating that most of the eroded material is deposited close to its source.

The interaction of the boundary plasma with the limiters has been analyzed for two 5 MA, 20 s Ohmic discharges in JET by Langmuir probes and by an all carbon collector probe. The deposition of D, C, Ni and Cr and the erosion on the probe have been measured as functions of the distance to the last closed magnetic flux surface. The values are compared to the particle fluences and their energies as calculated from the measured plasma parameters. The calculated deuterium fluences are about a factor of 5 to 10 larger than the deposition on the probe and the calculated e-folding length is steeper than the measured values. The carbon deposition and erosion at the probe has been determined with the aid of 13 C implants as well as from the depth of the metal and deuterium depositions and the interference colours of the deposited films. The distribution of deposited carbon is similar to the trapped deuterium with a ratio D : C of about 0.3, indicating that D trapping is dominated by saturation of the deposited C layer. The metal deposition is small; it amounts to a few times 10 −3 of the carbon deposition, indicating the dominance of carbon in the boundary layer.

This paper presents a review of hydrogen and helium recycling phenomena in tokamaks with limiters and walls largely made out of carbon (graphite, a-C:H layers). The key points of interest are the plasma fuelling efficiency, the wall pumping phenomena as observed in JET, TFTR, TEXTOR and other machines under various fuelling schemes (gas, neutral beams, pellets), the release of hydrogen/helium from material surfaces during and after plasma discharges and the long term retention (total particle inventory) of hydrogen in graphite or carbonised structures in tokamaks. The effect of a combined hydrogen/helium plasma on recycling is also discussed. It is shown that only part of the above phenomena can be understood in terms of processes between hydrogen/helium and carbon as known from simulation experiments (ion beams, gas discharge facilities) and that others (in particular the JET wall pumping phenomenon) have still to be explained. Possible mechanisms are outlined and discussed by means of global models.

Profiles of density and temperature have been measured in the limiter scrape-off layer on JET using Langmuir probe arrays at the torus top and mid-plane positions. When ICRF heating is applied the local density, electron temperature and floating potential rise. Both particle flux and electron temperature show the characteristic exponential dependence on distance from the plasma edge but the e-folding length increases when ICRH starts. The particle flux e-folding length rises rapidly with ICRH power. Data from top and mid-plane probe arrays show good agreement when the field line compression is taken into account. The time response of particle flux to applied RF power is fast, 0.25–3 ms, however only low level RF voltages have been detected on the Langmuir probes suggesting that sheath rectification is not distorting the data. A change in perpendicular diffusion due to ICRH and ionisation of hydrogen molecules desorbing from the torus wall are considered as possible causes of the profile flattening.

The amounts of H and D trapped in the surface layers of samples made out of C, Ni, Inconel and Si, installed at the vessel walls of JET and removed after the 1984 and 1985 periods of operation were measured. After the 1984 operation period the D concentration in carbon probes was of the order of 5 × 10 19 /m 2 . while the hydrogen was about a factor of 40 higher. This can be attributed to a final glow discharge cleaning in hydrogen at the end of the operation period. After the 1984 period of operation the amounts of deuterium and hydrogen were nearly equal at about 5 × 10 21 /m 2 . A comparison of the amounts of deuterium and hydrogen with the amount of carbon deposition on the Inconel wall samples shows a ratio of about 0.3 to 0.4 (H + D) per C-atom. The hydrogen concentrations trapped in the vessel wall correspond to more than 100 times the amount of hydrogen isotopes in the plasma during a discharge.

The surface layers of tiles from the carbon limiters in octant 1, ocant 4 and the RF-antenna limiters were analysed after the end of the first three experimental sessions of JET, 1983, 1984 and 1985. On all tiles metallic contaminants were found from the vessel walls and the RF antennae. In general on all areas deposition and erosion was observed. At the part of the tiles in direct contact with the plasma the erosion dominates, whereas deposition dominates on the outer parts of the limiter radially further away from the plasma. In the small gaps between adjacent limiter tiles, deposited layers of carbon having a thickness of several μm were found, which were partly flaked and contained metallic impurities. Electrical are tracks were observed on the limiter tiles and in the area between limiter tiles.

After five years of construction JET started its operation in June 1983. In December plasmas with peak currents of 3 MA were produced. Flat top times of a few seconds were obtained for currents of 2.4 MA. The plasmas were slightly elongated ( ~ 20%) with a minor radius of about 1 m. Line densities (peak values) reached 7 × 10 19 m −2 , electron temperatures on the axis 1.5–2 keV and energy confinement times 0.3 s. The paper describes the state of the machine, the methods employed for impurity control and summarises the experimental results obtained during 1983.

Investigation of the scrape-off layer (SOL) at different toroidal and poloidal positions has been carried out in JET for limiter discharges with and without ion cyclotron resonance heating (ICRH). A wide range of plasma conditions ( I p ≈1−5 (rmMA), B T ≈1.8−3.5 T , 〈 n e 〉 ≈ (1−5) × 10 19 m −3 ), ICRH heating scenarios (H and He3 minority, quadrupole, monopole, dipole antenna phasing) and different machine configurations were investigated. The total ICRH power coupled to the JET plasma was ≤ 7.2 MW, when three antennae were used, and ≤ 13 MW using eight antennae operation. Limiter discharges were produced with either eight discrete graphite limiters or two full toroidal belt limiters. Density profiles in the SOL were found to decay exponentially with an e-folding length, λ n e , which changes with total input power, wall conditioning and plasma species. The modification of the plasma edge depends on the antenna phasing. The scaling of edge temperature, edge density and impurity production with 〈 n e 〉 and P T for different scenarios is presented. Changes in SOL parameters are also compared with different torus wall and antenna conditioning.

Ohmic heating studies in hydrogen and deuterium up to currents of Ip = 5 MA have been completed and additional heating experiments by means of ICRH and NBI are now in process. With ICRH powers up to 7 MW, the global energy confinement time τE is observed to deteriorate with increasing heating power. NBI experiments are at a more preliminary stage, but also show deterioration in τE. Detailed studies of confinement time scaling in ohmically heated discharges show a strong dependence of plasma size, τE ∼ R1.7a1.3, but a much weaker dependence on q and ne than seen in smaller experiments. Boundary phenomena have been studied from early 1984. Data on the impurity coverage of limiters and wall have been collected during three successive experimental campaigns. Substantial amounts of wall material have consistently been found on the limiters and a large inventory of hydrogen isotopes concentrated on the colder edges. The wall in the shadow of protective elements shows non-uniform erosion. Langmuir probe measurements of edge plasma parameters have been taken in OH, RF and NB discharges. Additional heating leads generally to a temperature increase of the scrape-off layer. Model calculations using data from these diagnostics seem to produce a consistent picture of impurity production in the scrape-off layer which agrees reasonably well with spectroscopic observations.

The JET scrape-off plasma during X-point discharges with L- and H-modes is described. There is a flow of current in the scrape-off layer which can influence equilibrium, transport, heating and impurity generation in the scrape-off layer. A local measurement of D ⊥ is obtained in the divertor region. The saturated flux of ions flowing parallel to the magnetic field, I SAT is ∼10 A/cm2 at the divertor separatrix in ohmic discharges. In L-mode with NBI it increases by a factor ⩽ 2, depending on injected power level ( P NBI MW ) and decreases with H-mode. The e-folding decay length, λ I SAT , is the same in ohmic and L-mode but decreases with H-mode. The electron temperatures at the divertor target are ~ 30 eV (OH) and ~ 50–60 eV (H-mode). Measurements of T e during L-mode are not reported because of the high level of fluctuations. An enhancement in the erosion of carbon due to ion-stimulated desorption at the divertor target is discussed. Flow reversal in the scrape-off layer is demonstrated.