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This report summarizes the contributions presented at the 3rd IAEA technical meeting on divertor concepts, held in Vienna, Austria, 4–7 November 2019. The meeting brought together more than 70 experts from nuclear fusion research sites worldwide to discuss the different aspects that the divertor design and fusion machine operation involve, from ITER divertor developments to innovative technologies for future DEMO divertor. The main topics of the meeting were: divertor and confinement; radiative power exhaust; scrape-off layer (SOL) and divertor physics; steady state operation and transient heat loads; plasma facing components materials and heat exhaust for steady state operation; and divertors for DEMO and future power reactors.

The cathode spots of vacuum arcs emit material in the form of plasma as well as droplets. Generated by arcs burning at the first wall of fusion devices, the droplets may effectively contaminate the fusion plasma. Essential characteristics of the droplets (diameter, velocity, and emission direction) and their interrelations are poorly known so far. In this paper, a new approach of optical droplet investigation is presented. Emitted from a pulsed vacuum arc, the droplets fly vertically inside a drift tube against gravity, finally passing two consecutive light beams. The time-of-flight and the detected intensity of scattered light allow a simultaneous determination of droplet velocity and size. Different solid angle directions have been realized by turning the cathode with respect to the flight tube axis. Using this technique, the parameters of droplets emitted from a Cu cathode into different directions were obtained. Two distinct groups could be identified at smaller ( 20°) angles between the surface plane and emission direction, respectively. They exhibit different velocity distributions and different relations between the particle diameter and the emission velocity.

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6 × 1020 Li atoms were launched with a speed of 600 m s−1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.

Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

The observation in JET of co-deposition of tritium with carbon has led to a broad discussion on the replacement of graphite by a high-Z material for the first wall coverage. Moreover, due to the high erosion rate, carbon plasma facing components (PFCs) appear to be unacceptable for a commercial fusion reactor. Research in this area has subsequently gained increased attention. This paper describes the status of investigations on the use of tungsten as a first wall material. It discusses on the physical side the plasma wall interaction, the transport of tungsten in the plasma boundary and in the core. As an intermediate step on the technological side, graphite is often coated with tungsten layers. For highly loaded surfaces in a fusion reactor finally bulk tungsten components will have to be developed.

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to aid the design and prepare operation. In order to increase its exhaust capabilities and operational flexibility a new bulk W divertor as well as an adjustable cryo-pump had been installed prior to the 2014 campaign. In experiments with high field side pellet injection central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power a large normalized heat flux Psep/R=10 MWm−1 has been reached, representing 2/3 of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm−2. The divertor load could be further reduced by increasing the core radiation still keeping the confinement in the range of H98y≈1. ELM suppression at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER Q=10 baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, q95 and the distance the L-H threshold. It turned out that the ELMs frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative an improved H-Mode scenario has been developed achieving a similar performance at lower plasma current (and consequently higher q95). Experiments using ECCD with feedback controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationary achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI) a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at the strengthening of the power supplies in order to allow full use of the installed heating power, the exchange of two ICRH antennas to reduce the W influx during ICRH and the upgrading of the ECRH system to 7–8 MW for 10s.

Experiments dealing with plasma-wall interactions and first-wall materials comprise a significant part of the work program of ASDEX Upgrade. To elucidate carbon chemical erosion under reactor-relevant conditions, dedicated spectroscopic measurements were performed. These investigations are complemented with long-term erosion and deposition probes consisting of various materials, which are mounted at numerous locations inside the vacuum vessel. The codeposition of hydrogen with carbon below the divertor is studied in detail with long-term samples as well as with quartz microbalance measurements, which allow a discharge-resolved measurement of the layer growth. In parallel to the investigations on carbon, the behavior of tungsten plasma facing components (PFCs) and their influence on plasma performance is studied. In several experimental campaigns, the divertor as well as large parts of the PFCs in the main chamber were equipped with tungsten-coated tiles. Surface conditioning by applying a silicon layer (siliconization) was performed as a preexperiment of the tungsten program, and the results are compared to those of surface conditioning with boron (boronization)

Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

Tungsten divertor plates have been used in ASDEX Upgrade for a full experimental campaign of approximately 800 discharges. Key issues investigated were the plasma performance, tungsten erosion, redeposition and migration, hydrogen isotope retention and the consequences of the simultaneous use of carbon and tungsten as plasma facing materials. Tungsten behavior was investigated by spectroscopic observation of line and quasi-continuum radiation and by surface analysis of material probes and of samples taken from plasma facing components after the experimental campaign. Code simulations were used to gain further understanding of the observed tungsten erosion properties and of tungsten plasma transport processes. The results obtained so far are presented and the implications with respect to the construction of future fusion devices are discussed.

The major aspects linked to the use of the JET auxiliary heating systems: NBI, ICRF and LHCD, in the new JET ITER-like wall (JET-ILW) are presented. We show that although there were issues related to the operation of each system, efficient and safe plasma heating was obtained with room for higher power. For the NBI up to 25.7MW was safely injected; issues that had to be tackled were mainly the beam shine-through and beam re-ionisation before its entrance into the plasma. For the ICRF system, 5MW were coupled in L-mode and 4MW in H-mode; the main areas of concern were RF-sheaths related heat loads and impurities production. For the LH, 2.5 MW were delivered without problems; arcing and generation of fast electron beams in front of the launcher that can lead to high heat loads were the keys issues. For each system, an overview will be given of: the main modifications implemented for safe use, their compatibility with the new metallic wall, the differences in behavior compared with the previous carbon wall, with emphasis on heat loads and impurity content in the plasma., 21 pages, 17 figures

Several experiments were conducted in ASDEX Upgrade to prove the suitability of tungsten as a divertor target material under the conditions of a high density and low temperature divertor. The observed fluxes from a tungsten tile into the plasma are low, in keeping with the extremely low sputtering yields. In addition, the very favourable effect of `prompt redeposition' (redeposition during the first gyration) could be confirmed by the experiments. Cooling of the edge region by neon injection seems permissible, i.e. neon impurity sputtering did not increase the eroded fluxes of tungsten. The transport and accumulation behaviour were investigated by means of the laser blow-off technique. No accumulation effects could be observed in ohmic discharges. In discharges with NBI heating but without ICRH, strong accumulation can occur. High heat flux tests were performed on graphite tiles coated with plasma sprayed tungsten, which withstood a thermal load of 15 MW/m2 lasting 2 s as well as 1000 cycles of 10 MW/m2 for 2 s without disabling damage. Owing to the encouraging results, an experiment using a tungsten divertor is planned in ASDEX Upgrade

JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0-2.4 T, δ = 0.2-0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 1021, 1.37 × 1021 and 1.97 × 1021 D s-1 down to 4.8 × 1019, 7.2 × 1019 and 16 × 1019 D s-1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 1020 D s-1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device. © 2013 IAEA, Vienna.

The use of refractory metal plasma-facing components (PFCs) requires intensive research in all areas, i.e., in plasma-wall interaction, in the physics of the confined plasma, diagnostic, and in material development. Only a few present-day divertor tokamaks-mainly Alcator C-Mod (C-Mod) and ASDEX Upgrade (AUG)-gained experience with the refractory metals molybdenum and tungsten, respectively. AUG was stepwise converted from graphite to tungsten PFCs. In line with this transition, a reduction of the deuterium retention by almost a factor of ten has been observed due to the strong suppression of D codeposition with carbon. The deuterium retained in W is in line with laboratory results in contrast to C-Mod, where the D retention in Mo is more than a factor of ten larger than that in corresponding laboratory experiments. As expected from the sputtering threshold of Mo and W, negligible erosion by the thermal divertor background plasma is found in these experiments under low-temperature divertor conditions. However, erosion by fast particles and intrinsic impurities, which additionally might be accelerated in rectified electrical fields observed during ion cyclotron frequency heating, plays an important role. The Mo and W concentrations in the plasma center are strongly affected by plasma transport, and variations up to a factor of 50 are observed for similar influxes. However, it could be demonstrated that sawteeth and turbulent transport driven by central heating can suppress central accumulation. The inward transport of high-Z ions at the edge can be efficiently reduced by ?flushing? the pedestal region caused by frequent edge instabilities. Extrapolations to ITER and DEMO are difficult since the physics of the plasma transport is not yet completely understood, the particle and energy fluxes are orders of magnitude higher, and the technical boundary conditions in DEMO strongly differ from those of present-day devices.

The use of refractory metal PFCs requires intensive research in all areas, i.e. in plasma wall-interaction, in the physics of the confined plasma, diagnostic, and in material development. Only a few present day divertor tokamaks - mainly Alcator C-Mod and ASDEX Upgrade - gained experience with the refractory metals molybdenum and tungsten, respectively. ASDEX Upgrade was stepwise converted from graphite to tungsten PFCs. In line with this transition a reduction of the deuterium retention by almost a factor of ten has been observed due to the strong suppression of D co-deposition with carbon. The deuterium retained in W is in line with laboratory results in contrast to Alcator C-Mod, where the D retention in Mo is more than a factor of ten larger than in corresponding laboratory experiments. As expected from the sputtering threshold of Mo and W, negligible erosion by the thermal divertor background plasma is found in these experiments under low temperature divertor conditions. However, erosion by fast particles and intrinsic impurities, which additionally might be accelerated in rectified electrical fields observed during ion cyclotron frequency heating, plays an important role. The Mo and W concentrations in the plasma centre are strongly affected by plasma transport and variations up to a factor of 50 are observed for similar influxes. However, it could be demonstrated that sawteeth and turbulent transport driven by central heating can suppress central accumulation. The inward transport of high-Z ions at the edge can be efficiently reduced by ‘flushing’ the pedestal region caused by frequent edge instabilities (ELMs). Extrapolations to ITER and DEMO are difficult since the physics of plasma transport is not yet completely understood, the particle and energy fluxes are orders of magnitude higher and the technical boundary conditions in DEMO strongly differ from those of present day devices.

ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.

Scanning electron microscopy images and energy-dispersive x-ray spectra were recorded for a total of about 4×104 dust particles collected on the same position within the vacuum vessel via silicon wafers during four consecutive full-tungsten first wall campaigns of ASDEX Upgrade between 2007 and 2009. By careful analysis of the elemental composition and shape of the sampled particles, seven statistically relevant classes of dust were identified. The particle flux and area coverage of each class were normalized to the total plasma duration of each sampling period, revealing a high sensitivity of the dust composition to device conditioning. According to the present results, particles produced by arcing on divertor tiles with delaminated coatings were transported to the main chamber first wall.

Spectroscopy has always been an integral part of the diagnostic systems of super-hot fusion plasmas. The numerous results derived from studies of the emitted spectra have been highly important for elucidating key physical properties of fusion plasmas; moreover, plasma spectroscopy has provided crucial input for development of new tokamaks, stellarators and other advanced devices. Many concepts of spectroscopic diagnostics in magnetic confinement fusion (MCF) are well established and have been thoroughly tested over decades. However the recent advances on the existing machines (e.g. JET, ASDEX Upgrade and LHD), linked to the accelerating development of the international ITER project, call for new and improved spectroscopic methods. These are required to determine the key plasma parameters in all regions of the plasma volume while coping with complex materials, new advanced operating scenarios and a hostile radiation environment. It is such advanced methods and techniques that are the focus of this special issue. The papers collected here provide an extensive representation of the state of the art in spectroscopic diagnostics of MCF plasmas. On the experimental side, several contributions from the existing tokamaks and stellarators demonstrate how sophisticated spectroscopic methods are used to derive information on temporal evolution of electron temperature and density, particle velocity in peripheral plasmas, edge ion temperature, and many other quantities of interest. A group of papers addresses development of new experimental techniques for future measurements, including specific spectroscopic tools for ITER diagnostics. As accurate atomic data are at the cornerstone of reliable diagnostics, several papers describe newly calculated spectroscopic and collisional data for species and processes of highest importance in fusion devices, such as, for example, charge exchange for neutral beam injection diagnostics. Finally, a group of contributions address various issues related to theoretical modelling of plasma emission spectra including collisional-radiative simulations and line profile modelling. We hope that the papers contributed to this special issue will serve as a valuable resource for the MCF community.

For the investigation of high-Z impurities in the ASDEX Upgrade tokamak a sublimation probe was developed and tested. With this probe it is possible to inject materials, that sublimate at temperatures from about 50 to 150 °C into the plasma through a controllable valve. For the investigation of the tungsten transport in ASDEX Upgrade the probe was operated with tungsten carbonyl. The flux of tungsten, which is difficult to determine directly because of the uncertain atomic data, can be determined using the fluxes of oxygen and carbon, the atomic data of which are better known. In this article the setup of the probe and first experiments are described. Here the layer of deposited tungsten was investigated and the number of emitted photons per ionization (the S/XB ratio) for the 400.8 nm line of W I was estimated.