The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, good confinement conditions can be maintained only for the first 2–3 s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two independent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhance W neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.

Abstract:JET underwent a transformation from a full carbon-dominated tokamak to a full metallic device with the ITER-like wall combination for the activated phase with Beryllium main chamber and Tungsten divertor. The ITER-Like Wall (ILW) experiment at JET provides an ideal test bed for ITER and shall demonstrate as primary goals the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments ( = 2.0 , = 2.0 − 2.4 , = 0.2 − 0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction of the long term retention rate by a factor ten with respect to carbon references. All experiments have been executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system has been reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges ( = 0.5 ), type III ( = 5.0 ), and type I ELMy H-mode plasmas ( = 12.0 ) a drop of the retention rate normalised to the operational time in divertor configuration has been measured from 1.27 × 10 has been obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and widens the operation without active cleaning in the DT phase in comparison to a full carbon device.

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In 2003, the performance of the ‘hybrid’ regime was successfully validated in JET experiments up to βN = 2.8 at low toroidal field (1.7 T), with plasma triangularity and normalized Larmor radius (ρ*) corresponding to identical ASDEX Upgrade discharges. Stationary conditions have been achieved with the fusion figure of merit ( ) reaching 0.42 at q95 = 3.9. The JET discharges show similar MHD, edge and current profile behaviour, when compared with the ASDEX Upgrade. In addition, the JET experiments have extended the hybrid scenario operation at higher toroidal field of 2.4 T and lower ρ* towards the projected ITER values. Using this database, transport and confinement properties are characterized with respect to the standard H-mode regime. Moreover, trace tritium has been injected to assess the diffusion and convective coefficients of the fusion fuel. The maximization of confinement and stability properties provides, to this scenario, a good probability of achieving a high fusion gain at reduced plasma current for durations of up to 2000 s in ITER.

This paper reports the progress made at JET-ILW on integrating the requirements of the reference ITER baseline scenario with normalised confinement factor of 1, at a normalised pressure of 1.8 together with partially detached divertor whilst maintaining these conditions over many energy confinement time. The 2.5MA high triangularity ELMy H-modes are studied with two different divertor configurations. The power load reduction with N seeding is reported. The relationship between an increase in energy confinement and pedestal pressure with triangularity is investigated. The operational space of both plasma configurations is studied together the ELM energy losses and stability of the pedestal of unseeded and seeded plasmas.