We present the results from a new fuelling scan database consisting of 14 high triangularity ( ~ 0.41), Type I ELMy H-mode JET plasmas. As the fuelling level is increased from low, ( D ~ 0.2x10 22 el/s, n e,ped /n gw =0.7), to high dosing ( D ~ 2.6x10 22 el/s, n e,ped /n gw =1.0) the variation in ELM behaviour is consistent with a transition from 'pure Type I' to 'mixed Type I/II' ELMs [1]. However, the pulses in Scattering (HRTS) system. We continue by presenting, for the first time, the role of pedestal structure, as quantified by a least squares mtanh fit to the HRTS profiles, on the performance across the fuelling scan. A key result is that the pedestal width narrows and peak pressure gradient increases during the ELM cycle for low fuelling plasmas, whereas at high fuelling the pedestal width and peak pressure gradient saturates towards the latter half of the ELM cycle. An ideal MHD stability analysis shows that both low and high fuelling plasmas move from stable to unstable approaching the ideal ballooning limit of the finite peeling ballooning stability boundary. Comparison to EPED predictions show on average good agreement with experimental measurements for both pedestal height and width however when presented as a function of pedestal density, experiment and model show opposing trends. The measured pre-ELM pressure pedestal height increases by ~ 20% whereas EPED predicts a decrease of 25% from low to high fuelling. Similarly the measured pressure pedestal width widens by ~ 55%, in poloidal flux space, whereas EPED predicts a decrease of 20% from low to high fuelling. We give two possible explanations for the disagreement. First, it may be that EPED under-predicts the critical density, which marks the transition from kink-peeling to ballooning limited plasmas. Second, the stronger broadening of the experimental pedestal width than predicted by EPED is an indication that other transport related processes contribute to defining the pedestal width such as enhanced inter-ELM transport as observed at high fuelling, for mixed Type I/II ELMy pulses.M. Leyland

The baseline type I ELMy H-mode scenario has been re-established in JET with the new tungsten MKII-HD divertor and beryllium on the main wall (hereafter called ITER-like wall, JET-ILW). The first JET-ILW results show that the confinement is degraded by 20-30% in the baseline scenarios compared to the previous carbon wall JET (JET-C) plasmas. The degradation is mainly driven by the reduction in the pedestal temperature. Stored energies and pedestal temperature comparable to the JET-C have been obtained to date in JET-ILW baseline plasmas only in the high triangularity shape using N 2 seeding. This work compares the energy losses during ELMs and the corresponding time scales of the temperature and density collapse in JET-ILW baseline plasmas with and without N 2 seeding with similar JET-C baseline plasmas. ELMs in the JET-ILW differ from those with the carbon wall both in terms of time scales and energy losses. The ELM time scale, defined as the time to reach the minimum pedestal temperature soon after the ELM collapse, is ≈2ms in the JET-ILW and lower than 1ms in the JET-C. The energy losses are in the range ∆W ELM /W ped ≈7-12% in the JET-ILW and ∆W ELM /W ped ≈10-20% in JET-C, and fit relatively well with earlier multi-machine empirical scalings of ∆W ELM /W ped with collisionality. The time scale of the ELM collapse seems to be related to the pedestal collisionality. Most of the non-seeded JET-ILW ELMs are followed by a further energy drop characterized by a slower time scale ≈8-10ms (hereafter called slow transport events), that can lead to losses in the range ∆W slow /W ped ≈15-22%, slightly larger than the losses in JET-C. The N 2 seeding in JET-ILW significantly affects the ELMs. The JET-ILW plasmas with N 2 seeding are characterized by ELM energy losses and time scales similar to the JET-C and by the absence of the slow transport events.

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

DT discharges with 500 MW of fusion power on ITER will rely on partially detached divertor operations to keep target heat loads at manageable levels. Such divertor regime will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or a combination of diagnostics signals as feedback. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like Wall discharges by using saturation current (I sat ) measurements from divertor Langmuir probes as feedback signal to control the N seeding. The level of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such demonstration is a successful proof of principle in the context of ITER.