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.

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.

The most recent experiments of disruption mitigation by massive gas injection in ASDEX Upgrade have concentrated on small-relatively to the past-quantities of noble gas injected, and on the search for the minimum amount of gas necessary for the mitigation of the thermal loads on the divertor and for a significant reduction of the vertical force during the current quench. A scenario for the generation of a long-lived runaway electron beam has been established; this allows the study of runaway current dissipation by moderate quantities of argon injected. This paper presents these recent results and discusses them in the more general context of physical models and extrapolation, and of the open questions, relevant for the realization of the ITER disruption mitigation system.

Routine reaction to approaching disruptions in tokamaks is currently largely limited to machine protection by mitigating an ongoing disruption, which remains a basic requirement for ITER and DEMO [1]. Nevertheless, a mitigated disruption still generates stress to device. Additionally, in future fusion devices, high-performance discharge time itself will be very valuable. Instead of reacting only on generic features, occurring shortly before the disruption, the ultimate goal is to actively avoid approaching disruptions at an early stage, sustain the discharges whenever possible and restrict mitigated disruptions to major failures. Knowledge of the most relevant root causes and the corresponding chain of events leading to disruption, the disruption path, is a prerequisite. For each disruption path, physics-based sensors and adequate actuators must be defined and their limitations considered. Early reaction facilitates the efficiency of the actuators and enhances the probability of a full recovery. Thus, sensors that detect potential disruptions in time are to be identified. Once the entrance into a disruption path is detected, we propose a hierarchy of actions consisting of (I) recovery of the discharge to full performance or at least continuation with a less disruption-prone backup scenario, (II) complete avoidance of the disruption to sustain the discharge or at least delay it for a controlled termination and, (III), only as last resort, a disruption mitigation. Based on the understanding of disruption paths, a hierarchical and path-specific handling strategy must be developed. Such schemes, testable in present devices, could serve as guideline for ITER and DEMO operation. For some disruption paths, experiments have been performed at ASDEX Upgrade and TCV. Disruptions were provoked in TCV by impurity injection into ELMy H-mode discharges and in ASDEX Upgrade by forcing a density limit in H-mode discharges. The new approach proposed in this paper is discussed for these cases. For the H-mode density limit sensors used so far react too late. Thus a plasma-state boundary is proposed, that can serve as an adequate early sensor for avoiding density limit disruptions in H-modes and for recovery to full performance.