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.

Theoretical investigations are performed in order to explain the plasma rotation increments induced by lower hybrid current drive (LHCD) in Tore Supra and the results are compared to the experimental observations. The intrinsic toroidal rotation is governed by several mechanisms in concert. The impact of the LHCD on each involved mechanism is analyzed. The neoclassical toroidal rotation is always in the counter-current direction. The toroidal diamagnetic velocity is of the order of the experimental toroidal velocity. At high plasma current the rotation evolution in the lower hybrid (LH) phase is controlled by the neoclassical friction force due to the trapped ions in banana trajectories through the toroidal diamagnetic velocity. This force results in the counter-current increment as observed in the experimental measurement of toroidal rotation. At low plasma current the rotation is dominated by momentum turbulent transport when the LH waves are applied. The Reynolds stress grows strongly compared to the high plasma current case and acts as a co-current force through its residual stress contribution. Momentum transport simulations are also performed with CRONOS (Artaud et al 2010) in order to assess the rotation increments induced by LHCD.

Please cite this article in press as: M. Irishkin, et al., Applications of Bayesian temperature profile reconstruction to automated comparison with heat transport models and uncertainty quantification of current diffusion, Fusion Eng. Des. (2015), http://dx.h i g h l i g h t s• We developed a method for automated comparison of experimental data with models.• A unique platform implements Bayesian analysis and integrated modelling tools.• The method is tokamak-generic and is applied to Tore Supra and JET pulses.• Validation of a heat transport model is carried out.• We quantified the uncertainties due to Te profiles in current diffusion simulations. a b s t r a c tIn the context of present and future long pulse tokamak experiments yielding a growing size of measured data per pulse, automating data consistency analysis and comparisons of measurements with models is a critical matter. To address these issues, the present work describes an expert system that carries out in an integrated and fully automated way (i) a reconstruction of plasma profiles from the measurements, using Bayesian analysis (ii) a prediction of the reconstructed quantities, according to some models and (iii) a comparison of the first two steps. The first application shown is devoted to the development of an automated comparison method between the experimental plasma profiles reconstructed using Bayesian methods and time dependent solutions of the transport equations. The method was applied to model validation of a simple heat transport model with three radial shape options. It has been tested on a database of 21 Tore Supra and 14 JET shots. The second application aims at quantifying uncertainties due to the electron temperature profile in current diffusion simulations. A systematic reconstruction of the Ne, Te, Ti profiles was first carried out for all time slices of the pulse. The Bayesian 95% highest probability intervals on the Te profile reconstruction were then used for (i) data consistency check of the flux consumption and (ii) defining a confidence interval for the current profile simulation. The method has been applied to one Tore Supra pulse and one JET pulse.

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.