The tight aspect ratios (typically A≈1.4) and low magnetic field of spherical tokamak (ST) plasmas, when combined with densities approaching the Greenwald limit, provide a significant challenge for all currently available auxiliary heating and current drive schemes. NBI heating and current drive are difficult to interpret in sub-megampere machines, as in order to achieve suitable penetration into the plasma core, fast ions have to be highly suprathermal and, as a result of the low magnetic field, can be non-adiabatic (i.e. non-conserving of magnetic moment µ0). The physics of NBI heating in START is discussed. The neutral beam injector deployed on START was clearly successful, having been instrumental in producing a world record tokamak toroidal beta of ≈40%. A fast ion Monte Carlo code (LOCUST) is described that was developed to model non-adiabatic fast ion topologies together with a high level of charge exchange loss and cross-field transport (present in START due to an envelope of high density gas surrounding the plasma). Model predictions compare well with experimental data, collected using a scanning neutral particle analyser, bolometric instruments and equilibrium reconstruction using EFIT. In particular, beta calculations based upon reconstruction of the pressure profile (by combining measurements from Thomson scattering, charge exchange recombination spectroscopy and model predictions for the fast ion distribution function) agree well with beta values calculated using EFIT alone (the routine method for calculation of START beta). These results thus provide increased confidence in the ability of STs to sustain high beta high confinement H mode plasmas and in addition indicate that the injected fast ions in collisional START plasmas evolve mainly due to collisional and charge exchange processes, without driving any significant performance degrading fast particle MHD activity.

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.

A Fokker–Planck formalism has been developed to interpret experimental data from high power electron cyclotron heating of tokamak plasmas in which the electron distribution function can be substantially distorted. The Fokker–Planck equation is solved using a 2-D code which incorporates electron trapping, a steady Ohmic electric field and a bounce averaged electron cyclotron heating term. The quasilinear RF diffusion coefficient is calculated using a single particle model which includes the mildly relativistic resonance condition and takes into account localized RF power injection and tokamak rotational transform effects. Comparisons are presented of the calculated and observed soft X-ray spectra from 60 GHz second harmonic electron cyclotron heating experiments on the CLEO tokamak. Good agreement between theory and experiment is found. In addition, predicted RF current drive efficiencies for the COMPASS tokamak are presented.

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.