Power‐to‐gas is a storage technology aiming to convert surplus electricity from renewable energy sources like wind and solar power into gaseous fuels compatible with the current network infrastructure. Results of CO2 dissociation in a vortex‐stabilized microwave plasma reactor are presented. The microwave field, residence time, quenching, and vortex configuration were varied to investigate their influence on energy‐ and conversion efficiency of CO2 dissociation. Significant deterioration of the energy efficiency is observed at forward vortex plasmas upon increasing pressure in the range of 100 mbar towards atmospheric pressure, which is mitigated by using a reverse vortex flow configuration of the plasma reactor. Data from optical emission shows that under all conditions covered by the experiments the gas temperature is in excess of 4000 K, suggesting a predominant thermal dissociation. Different strategies are proposed to enhance energy and conversion efficiencies of plasma‐driven dissociation of CO2.

A suite of diagnostics is proposed to characterize microwave plasma dissociation of CO2: laser scattering, Fourier transform infrared spectroscopy, and passive emission imaging. It provides a comprehensive performance characterization as is illustrated on the basis of experiments in a 2.45 GHz, 1 kW microwave reactor with tangential gas injection. For example, two operating regimes are identified as function of pressure: the diffuse and constricted plasma mode. Their occurrence is explained by evaluation of microwave propagation, which changes with the electron‐heavy particle collision frequency ve−h. In the diffuse mode, gas temperatures of 1500–3500 K are determined. The measured conversion degree, specific energy input, and temperature are summarized in a two‐temperature thermal model, which is solved to obtain the gas temperature at the periphery of the reactor and the size of the hot zone. Solutions are found with edge temperatures of hundreds of K, and hot zone fractions which agree with the measured behavior. The agreement shows that non‐thermal processes play only a marginal role in the measured parameter space of the diffuse discharge. In the constricted mode, the radial plasma size is independent of power. A skin depth equal to the plasma size corresponds to electron densities of 1018–1019 m−3. Temperatures in the central filament are in the range 3000–5000 K. Both discharge modes are up to 50% energy efficient in CO production. Rayleigh signals increase in the afterglow, hinting at rapid gas cooling assuming that the gas composition remains unchanged.

Wendelstein 7-X is the first comprehensively optimized stellarator aiming at good confinement with plasma parameters relevant to a future stellarator power plant. Plasma operation started in 2015 using a limiter configuration. After installing an uncooled magnetic island divertor, extending the energy limit from 4 to 80 MJ, operation continued in 2017. For this phase, the electron cyclotron resonance heating (ECRH) capability was extended to 7 MW, and hydrogen pellet injection was implemented. The enhancements resulted in the highest triple product (6.5 × 1019 keV m−3 s) achieved in a stellarator until now. Plasma conditions [Te(0) ≈ Ti(0) ≈ 3.8 keV, τE > 200 ms] already were in the stellarator reactor-relevant ion-root plasma transport regime. Stable operation above the 2nd harmonic ECRH X-mode cutoff was demonstrated, which is instrumental for achieving high plasma densities in Wendelstein 7-X. Further important developments include the confirmation of low intrinsic error fields, the observation of current-drive induced instabilities, and first fast ion heating and confinement experiments. The efficacy of the magnetic island divertor was instrumental in achieving high performance in Wendelstein 7-X. Symmetrization of the heat loads between the ten divertor modules could be achieved by external resonant magnetic fields. Full divertor power detachment facilitated the extension of high power plasmas significantly beyond the energy limit of 80 MJ.

The modelling of a controlled tungsten dust injection experiment in TEXTOR by the dust dynamics code MIGRAINe is reported. The code, in addition to the standard dust–plasma interaction processes, also encompasses major mechanical aspects of dust–surface collisions. The use of analytical expressions for the restitution coefficients as functions of the dust radius and impact velocity allows us to account for the sticking and rebound phenomena that define which parts of the dust size distribution can migrate efficiently. The experiment provided unambiguous evidence of long-distance dust migration; artificially introduced tungsten dust particles were collected 120° toroidally away from the injection point, but also a selectivity in the permissible size of transported grains was observed. The main experimental results are reproduced by modelling.

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H  =  1 at βN ~ 1.8 and n/nGW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed.

Tungsten erosion in H-mode plasmas is quantified in the outer divertor of the JET ITER-Like Wall environment with optical emission spectroscopy on the 400.9 nm atomic neutral tungsten line. A novel crosscalibration procedure is developed to link slow, high spectral resolution spectroscopy and fast photomultiplier tube measurements in order to obtain ELM-resolved photon fluxes. Inter-ELM W erosion is exclusively impurity sputtering by beryllium because of the high sputter threshold for deuterons. Low beryllium concentrations resulted in low inter-ELM sputter yields of around 10 −4 with respect to the total flux. Intra-ELM W sources, which dominate the total W tungsten source, vary independently from the inter-ELM source. The amount of W erosion could only be partly explained by beryllium sputtering, indicating that during ELMs sputtering by fuel species is important. The total W outer divertor source is found to linearly increase with the power crossing the separatrix, whilst excessive divertor fueling can break this trend.The influence of the W source rate on the tungsten content of the core plasma is investigated using Soft X-Ray emission to determine the tungsten content. At low source rates the content is determined by the source, but at higher source rates, other phenomena determine the total tungsten content. Indications of impurity flushing by ELMs is seen at ELM frequencies above approximately 40 Hz. The inner/outer divertor asymmetry of the W source during ELMs is investigated, and the outer divertor W source is larger by a factor of 1.8 ± 0.7.

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: T, plasma current: MA, auxiliary power MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5– in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as...

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