We present an ultrafast neural network (NN) model, QLKNN, which predicts core tokamak transport heat and particle fluxes. QLKNN is a surrogate model based on a database of 300 million flux calculations of the quasilinear gyrokinetic transport model QuaLiKiz. The database covers a wide range of realistic tokamak core parameters. Physical features such as the existence of a critical gradient for the onset of turbulent transport were integrated into the neural network training methodology. We have coupled QLKNN to the tokamak modelling framework JINTRAC and rapid control-oriented tokamak transport solver RAPTOR. The coupled frameworks are demonstrated and validated through application to three JET shots covering a representative spread of H-mode operating space, predicting turbulent transport of energy and particles in the plasma core. JINTRAC-QLKNN and RAPTOR-QLKNN are able to accurately reproduce JINTRAC-QuaLiKiz T i,e and n e profiles, but 3 to 5 orders of magnitude faster. Simulations which take hours are reduced down to only a few tens of seconds. The discrepancy in the final source-driven predicted profiles between QLKNN and QuaLiKiz is on the order 1%-15%. Also the dynamic behaviour was well captured by QLKNN, with differences of only 4%-10% compared to JINTRAC-QuaLiKiz observed at mid-radius, for a study of density buildup following the L-H transition. Deployment of neural network surrogate models in multi-physics integrated tokamak modelling is a promising route towards enabling accurate and fast tokamak scenario optimization, Uncertainty Quantification, and control applications.

The possibility of using extreme ultraviolet emission from low charge states of tungsten ions to diagnose the divertor plasmas of the ITER tokamak has been investigated. Spectral modelling of Lu-like W 3+ to Gd-like W 10+ has been performed by using the Flexible Atomic Code, and spectroscopic measurements have been conducted at the Sustained Spheromak Physics Experiment (SSPX) in Livermore. To simulate ITER divertor plasmas, tungsten was introduced into the SSPX spheromak by prefilling it with tungsten hexacarbonyl prior to the usual hydrogen gas injection and initiation of the plasma discharge. The tungsten emission was studied using a grazing-incidence spectrometer.

The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019-2020, and tested the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed Shattered Pellet Injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

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 series of controlled experiments has been carried out in DIII-D to induce a bulk ion flow in the SOL and evaluate its effect on the localization of impurities in the divertor. This induced SOL flow was created by simultaneous deuterium puffing and divertor exhaust using a divertor cryopump, and the impurity enrichment was measured directly. The experiments were designed to compare enrichment in discharges with and without induced flow having otherwise similiar divertor parameters. Significant increases in impurity compression and enrichment are observed when flow is induced, and the degree of impurity enrichment in the divertor is found to be dependent on the impurity of interest. Detailed particle measurements made possible by the direct measurement of impurity densities in several reservoirs indicate reasonable particle balance for helium throughout the duration of the discharge. Conversely, while the total input of neon is balanced by the total exhaust by the end of a discharge, particle balance is not observed during the course of the discharge. A significant wall inventory with a short release time (∼10 ms) is surmised.

Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: Ip=1.0-2.0 MA, q95 approximately= 2.4-6.0 and total input power (≲20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached' divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (≳4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (≲30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q95 equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

Alpha particles with energies on the order of megaelectronvolts will be the main source of plasma heating in future magnetic confinement fusion reactors. Instead of heating fuel ions, most of the energy of alpha particles is transferred to electrons in the plasma. Furthermore, alpha particles can also excite Alfvénic instabilities, which were previously considered to be detrimental to the performance of the fusion device. Here we report improved thermal ion confinement in the presence of megaelectronvolts ions and strong fast ion-driven Alfvénic instabilities in recent experiments on the Joint European Torus. Detailed transport analysis of these experiments reveals turbulence suppression through a complex multi-scale mechanism that generates large-scale zonal flows. This holds promise for more economical operation of fusion reactors with dominant alpha particle heating and ultimately cheaper fusion electricity.