According to current knowledge and understanding, nuclear fusion can be developed to a sustainable energy technology. Fuel is abundant and key points as fusion power production and alpha particle heating have already been demonstrated. The next-step device ITER (International Thermonuclear Experimental Reactor) is designed to demonstrate net power production and to address most of the technological issues on the way to a power reactor. There is, however, a

Demonstrating improved confinement of energetic ions is one of the key goals of the Wendelstein 7-X (W7-X) stellarator. In the past campaigns, measuring confined fast ions has proven to be challenging. Future deuterium campaigns would open up the option of using fusion-produced neutrons to indirectly observe confined fast ions. There are two neutron populations: 2.45 MeV neutrons from thermonuclear and beam-target fusion, and 14.1 MeV neutrons from DT reactions between tritium fusion products and bulk deuterium. The 14.1 MeV neutron signal can be measured using a scintillating fiber neutron detector, whereas the overall neutron rate is monitored by common radiation safety detectors, for instance fission chambers. The fusion rates are dependent on the slowing-down distribution of the deuterium and tritium ions, which in turn depend on the magnetic configuration via fast ion orbits. In this work, we investigate the effect of magnetic configuration on neutron production rates in W7-X. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from DD fusion and, particularly, on the 14.1 MeV neutron production rates. Despite triton losses of up to 50 %, the amount of 14.1 MeV neutrons produced might be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges.

The transport of silicon has been investigated for various heating scenarios in ASDEX Upgrade H-mode discharges. Inside of r ≈ a/4, the diffusion coefficient D is either mainly neoclassical or anomalous depending on the heating method. For all investigated scenarios with NBI-heating and off-axis ECRH or off-axis ICRH, the diffusion coefficient is approximately neoclassical, and the effective heat diffusion coefficient χ eff is below the neoclassical ion heat diffusion χ i,neo in the plasma core. When central ECRH is added, χ eff is above χ i,neo , and D strongly increases by a factor of 3-10, i.e. becomes predominantly anomalous. For central ICRH, D is above the neoclassical level by a factor of 2.For radii outside of r ≈ a/4, D is always anomalous and increases towards the plasma edge. For r a/4, we find a clear scaling of D in terms of χ eff , where D is about equal or above χ eff . A strong inward drift parameter v/D is only observed in the core and only for cases, when the diffusion coefficient is neoclassical. With central wave heating, the drift parameter decreases to small values.

The elements of transport into and across the scrape-off layer in the poloidal divertor tokamak ASDEX Upgrade are analysed for different operational regimes with emphasis on enhanced confinement regimes with an edge barrier. Utilizing the existing set of edge diagnostics, especially the highresolution multi-pulse edge Thomson scattering system, in combination with long discharge plateaus, radial sweeps and advanced averaging techniques, detailed radial mid-plane profiles of diverted plasmas are obtained. Profiles are smooth across the separatrix, indicating strong radial correlation, and there is no remarkable variation across the second separatrix either. Together with measured input, recycling, pumping and bypass fluxes, a corrected separatrix position is determined and transport characteristics are derived in the different radial zones generally identified in the profile structure. Transport in the steep gradient region inside and across the separatrix shows typical ballooning-type critical electron pressure gradient scaling and, in parallel, even a clear correlation between radial electron density and temperature decay lengths (e.g. η e = d(ln T )/d(ln n) ∼ 2 for type-I ELMy H-modes). These findings indicate the importance of stiff profiles in this region, while diffusion coefficients are secondary parameters, determined essentially by the source distribution. The outer scrape-off layer wing exhibits a more filamentary structure with preferential outward drift especially in high-performance discharges, with formal diffusion coefficients far above the Bohm value in agreement with results on the old ASDEX experiment. A basic mechanism involved there seems to be partial loss of equilibrium and fast curvaturedriven outward acceleration, in principle well known from theory, investigated decades ago in pinch experiments and utilized recently in high-field-side pellet fuelling.

Injection of cryogenic deuterium pellets has been successfully applied in ASDEX Upgrade for external edge localized mode (ELM) frequency control in type-I ELMy H-mode discharge scenarios. A pellet velocity of 560 m s −1 and a size of about 6×10 19 D-atoms was selected for technical reasons, although even lower masses were found sufficient to trigger ELMs. A moderate repetition rate close to 20 Hz was chosen to avoid over-fuelling of the core plasma. Pellet sequences of up to 4 s duration were injected into discharges close to the L-H threshold, intrinsically developing large compound ELMs at a rate of 3 Hz. With pellet injection, these large ELMs were completely replaced by smaller type-I ELMs at the much higher pellet frequency, accompanied by a slight increase of density and even of stored energy. This external ELM control could be repeatedly switched on and off by just interrupting the pellet train. ELMs were triggered in less than 200 µs after pellet arrival at the plasma edge, at which time only a fraction of the pellet has been ablated, forming a rather localized, three-dimensional plasmoid, which drives the edge unstable well before the deposited mass is spread toroidally. The pellet controlled case has also been compared with a discharge at a somewhat lower density, but with otherwise rather similar data, developing spontaneous 20 Hz type-I ELMs. Despite the different trigger mechanisms, the general ELM features turn out to be qualitatively similar, possibly because of the similarity of the two cases in terms of ELM relevant parameters. The scaling with background plasma, heating power, pellet launch parameters, etc over a larger range still remains to be investigated.

At the central column of ASDEX Upgrade, an area of 5.5 m 2 of graphite tiles was replaced by tungsten-coated tiles representing about two-thirds of the total area of the central column. No negative influence on the plasma performance was found, except for internal transport barrier limiter discharges. The tungsten influx W stayed below the detection limit only during direct plasma wall contact or for reduced clearance in divertor discharges spectroscopic evidence for W could be found. From these observations a penetration factor of the order of 1% and effective sputtering yields of about 10 −3 could be derived, pointing to a strong contribution by light intrinsic impurities to the total W-sputtering. The tungsten concentrations ranged from below 10 −6 up to a few times 10 −5 . Generally, in discharges with increased density peaking, a tendency for increased central tungsten concentrations or even accumulation was observed. Central heating (mostly) by ECRH led to a strong reduction of the central impurity content, accompanied by a very benign reduction of the energy confinement. The observations suggest that the W-source strength plays only an inferior role for the central W-content compared to the transport, since in the discharges with increased W-concentration neither an increase in the W-influx nor a change in the edge parameters was observed. In contrast, there is strong experimental evidence, that the central impurity concentration can be controlled externally by central heating.

Plasma operation with high-Z plasma facing components is investigated with regard to sputtering, core impurity contamination and scenario restrictions. A simple model based on dimensionless quantities for fuel and high-Z ion sources and transport to describe the high-Z concentration in the plasma core is introduced. The impurity release and further transport is factorized into the sputtering yield, the relative pedestal penetration probability and a core confinement enhancement factor. Since there are quite large uncertainties, in particular, in the sputtering source and the edge transport of high-Z impurities, very different scenarios covering a wide parameter range are taken into account in order to resolve the experimental trends. Sputtering of tungsten by charge exchange neutrals in the energy range 0.5-2 keV is comparable to the effect of impurity ion sputtering, while the impact of thermal fuel ions is negligible. Fast ions produced by neutral beam injection as well as sheath acceleration during ICR heating may cause considerable high-Z sources if the limiters on the lowfield side have high-Z surfaces. The critical behaviour of the central high-Z concentration in some experimental scenarios could be attributed to edge and core transport parameters in the concentration model. The improved H-mode with off-central heating turns out to be the most critical one, since a hot edge is combined with peaked density profiles.

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald-like scaling, , for the RFP and the ohmic tokamak, a mixed scaling, , for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, are taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.