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.

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.

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

The radiative cooling coefficient for molybdenum (Z=42) in a low density (ne ≲ 1015 cm-3) plasma is calculated. First, the molybdenum charge state distribution (CSD) is computed using the best available atomic physics data for ground state recombination and ionization, including the rates of excitation-autoionization for Mo6+ to Mo13+ and Mo23+ to Mo32+. The emissivities of Mo4+ to Mo41+ are then found using a collisional-radiative model such that the contributions from metastable levels to an ion's emissivity are taken into account. The CSD and the radiative emissivity for all molybdenum ions are combined to yield the total radiative cooling coefficient for molybdenum in a low density plasma. A radiative loss coefficient over 2 orders of magnitude smaller than that predicted by an `average ion' model for temperatures relevant to tokamak divertor and scrape-off layer plasmas (Te ≲ 50 eV) is found. The cooling coefficient of the present work varies from a factor of 2 smaller to a factor of 2 larger than that predicted by the `average ion' model for all other plasma temperatures. The coefficient calculated in the present work is benchmarked against the measured bolometric loss profile from a molybdenum dominated shot in the Frascati Tokamak Upgrade (FTU)

The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.

The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving β T and β N values of 17% and 5.7 (%m T MA −1 ), respectively. A newly available motional Stark effect diagnostic has enabled validation of currentdrive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable β in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.

The calculated radiative cooling rate coefficient for krypton as a trace impurity in low to moderate density plasmas is calculated. Collisional-radiative line emission, dielectronic recombination, radiative recombination and bremsstrahlung are considered as the principal radiative loss channels. Collisional-radiative models and the calculated charge state distribution for krypton have been benchmarked against measured ion brightness profiles in the FTU plasma. The calculated radiative loss rate is compared with two measurements of the radiative cooling coefficient for krypton in the FTU plasma. The measurements differ in how the krypton density profile is experimentally constrained. The krypton density profile is found (to be flat) from spectroscopic observations and (to be increasing radially outwards) from measurements of visible bremsstrahlung emission. The calculations show excellent agreement with the former set of measurements. Polynomial fits to the total radiative cooling coefficient are provided for ease of use in plasma modelling codes. Tables of ion emissivities are provided for use in modelling radiative losses from non-equilibrium plasmas.

Experimental results on impurity transport in tokamaks are based on various techniques. We study here how the choice of the injection technique and of the analysis method influences the results. We have used three different injection techniques available in Tore Supra: laser blow-off, gas puff and supersonic molecular beam injection. We show that the long time duration of the gas puff injection compared with particle confinement time provides very limited information. The laser blow-off technique and supersonic pulsed injections give satisfactory results for diffusion but low quality convection estimates, presumably because the fast source term quenches the role of convection in the continuity equation. The best method is shown to be the combined analysis of supersonic pulsed injections and continuous puffing of a gaseous species. We obtain convection velocity profiles to an uncertainty of about 0.5 m s−1. This method is applied to ohmic, weakly sawtoothing plasmas. The diffusion coefficient is independent of the impurity charge and the convection velocity is inward. Neoclassical calculations show that these plasmas are dominated by anomalous transport. Quasilinear gyrokinetic simulations are in qualitative agreement with the above experimental results. We deduce from the simulations that convection is dominated by the curvature term, which means that no charge dependence should be expected in this situation.