To improve our understanding of the dynamics and control of ITER terminations, a study has been carried out on data from existing tokamaks. The aim of this joint analysis is to compare the assumptions for ITER terminations with the present experience basis. The study examined the parameter ranges in which present day devices operated during their terminations, as well as the dynamics of these parameters. The analysis of a database, built using a selected set of experimental termination cases, showed that, the H-mode density decays slower than the plasma current ramp-down. The consequential increase inf _GW limits the duration of the H-mode phase or result in disruptions. The lower temperatures after the drop out of H-mode will allow the plasma internal inductance to increase. But vertical stability control remains manageable in ITER at high internal inductance when accompanied by a strong elongation reduction. This will result in ITER terminations remaining longer at lowq( q ₉₅ ~ 3) than most present-day devices during the current ramp-down. A fast power ramp-down leads to a larger change inβ _p at the H–L transition, but the experimental data showed that these are manageable for the ITER radial position control. The analysis of JET data shows that radiation and impurity levels significantly alter the H–L transition dynamics. Self-consistent calculations of the impurity content and resulting radiation should be taken into account when modelling ITER termination scenarios. The results from this analysis can be used to better prescribe the inputs for the detailed modelling and preparation of ITER termination scenarios.

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on W transport assumptions during the ELM, a maximum ELM frequency is also identified above which core tungsten accumulation takes place.

Magnetic triggering of edge localized modes (ELMs) in Ohmic H-mode plasmas was first reported in the TCV tokamak (Degelinget al2003Plasma Phys. Control. Fusion451637). This method, showing reliable locking of the ELM frequency to an imposed axisymmetric vertical plasma oscillation, was also demonstrated in the ITER-relevant type-I ELM regime in ASDEX Upgrade (Langet al2004Plasma Phys. Control. Fusion46L31) and JET (de la Lunaet al2015Nucl. Fusion56026001). However, the mechanisms of the ELM triggering due to a vertical motion has not been studied extensively. The non-linear reduced MHD code JOREK-STARWALL has been extended for 3D free-boundary computations (Hölzlet al2012J. Phys.: Conf. Ser .401012010), which has allowed us to simulate for the first time realistic vertical oscillations together with ELM simulations in a single consistent scheme. Our simulations demonstrate that stable plasmas can be destabilized by the application of a vertical oscillation for ITER. During the vertical motion, a toroidal current is induced in the pedestal. The origin of this current is analysed in detail with the use of simulations and a simple analytical model, revealing that it arises from the compression of the plasma cross section due to its motion through an inhomogeneous magnetic field. Lower pedestal currents between ELMs require bigger vertical displacements to destabilize ELMs, which directly points towards the increased edge current as the ELM driving mechanism. Finally the ELM triggering shows a very weak dependence on the plasma velocity for ITER in agreement with experiments.

The electron temperature and density pedestals tend to vary in their relative radial positions, as observed in DIII-D (Beurskenset al2011Phys. Plasmas18056120) and ASDEX Upgrade (Dunneet al2017Plasma Phys. Control. Fusion5914017). This so-called relative shift has an impact on the pedestal magnetohydrodynamic (MHD) stability and hence on the pedestal height (Osborneet al2015Nucl. Fusion55063018). The present work studies the effect of the relative shift on pedestal stability of JET ITER-like wall (JET-ILW) baseline low triangularity ( δ ) unseeded plasmas, and similar JET-C discharges. As shown in this paper, the increase of the pedestal relative shift is correlated with the reduction of the normalized pressure gradient, therefore playing a strong role in pedestal stability. Furthermore, JET-ILW tends to have a larger relative shift compared to JET carbon wall (JET-C), suggesting a possible role of the plasma facing materials in affecting the density profile location. Experimental results are then compared with stability analysis performed in terms of the peeling-ballooning model and with pedestal predictive model EUROPED (Saarelmaet al2017Plasma Phys. Control. Fusion ). Stability analysis is consistent with the experimental findings, showing an improvement of the pedestal stability, when the relative shift is reduced. This has been ascribed mainly to the increase of the edge bootstrap current, and to minor effects related to the increase of the pedestal pressure gradient and narrowing of the pedestal pressure width. Pedestal predictive model EUROPED shows a qualitative agreement with experiment, especially for low values of the relative shift.