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.

JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E in = 35 eV) and more than 100%, caused by Be self-sputtering (E in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

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.

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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.