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The main limitation to the performance of JET optimized shear (OS) discharges is due to MHD instabilities, mostly in the form of a disruptive limit. The structure of the MHD mode observed as a precursor to the disruption, as measured from SXR and ECE diagnostics, shows a global ideal MHD mode. The measured mode structure is in good agreement with the calculated mode structure of the pressure driven kink mode. The disruptions occur at relatively low normalized beta (1 < β_N < 2), in good agreement with calculated ideal MHD stability limits for the n = 1 pressure driven kink mode. These low limits are mainly due to the extreme peaking factor of the pressure profiles. Other MHD instabilities observed in JET OS discharges include, usually benign, chirping modes. These modes, which occur in bursts during which the frequency changes, have the same mode structure as the disruption precursor but are driven unstable by fast particles.

The dependence of core plasma impurity transport on the Z number has been investigated for ASDEX Upgrade H mode discharges. For the elements Ne, Ar, Kr and Xe the diffusion coefficient in the centre is D ⩽ 6 × 10^-2m²/s and rises with the radial distance from the centre. With increasing Z number the transport becomes strongly convective with inward directed drift velocities that produce very peaked impurity densities for high Z. The inward drift decreases with decreasing deuterium density gradient. Neoclassical transport of the impurities has been calculated numerically. The calculated diffusion coefficient and drift velocity are close to the experimental values for the lower-Z elements Ne and Ar. However, for high Z, the calculated diffusion coefficient is smaller by a factor of up to 2.5 and the inward drift velocity is too small by a factor of 10. Toroidal rotation of the plasma that leads to an increased impurity density on the outboard side of the flux surfaces is not taken into account by the neoclassical calculations. Inboard/outboard asymmetries are not present for Ar and Ne with toroidal Mach number M_tor around 1. However, for heavier elements than Kr with M_tor ≈ 2 and an outboard/inboard ratio of ≈ 1.5, poloidal variation of the impurity density is important and might account for the discrepancy between the measured and calculated transport coefficients.

Ideal MHD limits to beta and bootstrap fraction are computed for pressure profiles similar to those of JET discharges with both internal and H mode transport barriers. Several non-monotonic current profiles are tested, and the pressure profile is constrained so that no negative current drive is required in steady state. Calculations are made both with and without an ideally conducting wall. The main result is that, for such peaked pressure profiles, the limits to beta and bootstrap fraction improve at low internal inductance, in particular when wall stabilization is taken into account. The highest limits to beta, and often also to normalized beta, occur for the maximum plasma current. There is also a weak dependence on q_min, for which three favourable regions have been identified. A highly advantageous region is found at q_min ≃ 1.6, where the limits to β* are 7.0% with, and 4.8% without, wall stabilization. The corresponding limits are 68 and 50%, respectively, for the bootstrap fraction and 4.0 and 2.9 for the normalized beta. These equilibria have low internal inductance, l_i = 0.62. For higher inductance, an optimum occurs when q_min ≃ 1.2, where the limit to β* is 5.3% with a wall and 4.7% without. The corresponding bootstrap fractions are about 46 and 38%, respectively. A third type of equilibrium that is interesting for steady state operation has q_min ≃ 2.1 and low inductance. Here the β* limits are lower, 4.9 and 3.4%, but the bootstrap fractions are higher, 77 and 60%.

Observations on ASDEX Upgrade show that fishbones can have a much stronger effect on the background plasma parameters than expected from an ideal, fast particle driven instability. They lead to a redistribution of electron temperature and impurity densities, similar in form and magnitude to sawteeth, albeit with the plasma parameter variation occurring on a significantly longer time-scale. Simulations of the current diffusion in discharges with favourable core confinement require fishbones to be furthermore associated also with magnetic reconnection to yield consistency between the conductivity and current profile information derived from ECE, bremsstrahlung, MSE and the localization of m = 1,n = 1 activity.

A resonance ω = k_||v_|| between energetic ions and a magnetic island with a half-width greater than the bulk ion Larmor radius is considered. As a result of such a resonant interaction, energetic ions with a relative density of a few per cent contribute significantly to the toroidal torque balance, leading to considerable modification of the equilibrium island rotation frequency ω. The possibility of a frequency pulsation scenario, instead of a stationary toroidal torque balance, is indicated.

A close coupled, distributed radiator heavy ion target is presented. Close coupled refers to a decrease in the distance between the hohlraum wall and the inertial confinement fusion capsule. In two dimensional, integrated, LASNEX calculations, this target produced 436 MJ of yield from 3.27 MJ of ion beam energy for a gain of 133. To achieve these results, the hohlraum dimensions were reduced by 27% from the previous distributed radiator, heavy ion target while driving the same capsule. This reduced the beam energy required from 5.9 to 3.27 MJ. Calculations of single mode Rayleigh-Taylor growth for this capsule show that this capsule is more stable than at least one of the NIF target designs (the PT design which uses a CH ablator doped with oxygen and bromine). This means that issues regarding the Rayleigh-Taylor instability for the heavy ion driven capsule can be settled on NIF. This close coupled target can also be scaled down in size for an Engineering Test Facility; LASNEX calculations predict that a gain of 94 can be achieved from 1.75 MJ of beam energy. In addition, gain curves for distributed radiator targets with the `conventional' case to capsule ratio and the close coupled case to capsule ratio are presented.

Magnetohydrodynamic stability calculations of the n = 1 kink mode for equilibria with increasing β and values of the axis safety factor q₀ > 1 and q₀ < 1 show that for q₀ < 1 there is a transition in both the unstable mode structure and the growth rate to a more virulent instability at approximately the β limit obtained with q₀ > 1. This provides justification for the usual computational procedure of optimizing β by taking q₀ > 1 and applying it to predict the β limit of sawtoothing discharges.

Discharge optimization for improving MHD stability of both core and edge was essential for the achievement of record fusion power discharges, in the ELM-free hot ion H mode regime, in the recent JET DT operation. The techniques used to increase edge stability are described. In particular the successful technique of current rampdown used to suppress the outer mode is reported. The increased stability of the outer mode by decreasing the edge current density confirms its identification as an n = 1 external kink. Decreasing the plasma current, however, decreases the ELM-free period, which is consistent with stability calculations that show an earlier onset of the ballooning limit. In order to increase external kink stability without causing a deterioration in the ELM-free period, a compromise was achieved by using plasma current rampdown, while working at the highest plasma current values possible. Results from a plasma current scan show that at the time of occurrence of the first giant ELM, the plasma stored energy, as well as the pressure measured at the top of the edge pedestal increase linearly with plasma current, for a given plasma configuration and power. This is consistent with models of the edge transport barrier, where the transport barrier width is proportional to the ion (or fast ion) poloidal Larmor radius. The MHD observations in DT and deuterium only discharges were found to be similar. Thus the experience gained on the control of MHD modes in deuterium plasmas could be fully exploited in the DT campaign.

An advanced heat removal scenario is required to handle the high input power of magnetic fusion reactors. The concepts of gas target and radiative divertors have been explored. A set of self-consistently coupled fluid equations for plasma and neutrals is suitable for this system because of the sophisticated equations for neutrals. The transfer of plasma momentum to neutrals, which is crucial in understanding the plasma-neutral interface, is described through the charge exchange process. Our code results show good agreement with gas target experiments on the PISCES experiment and demonstrate its superiority to the previous simulation. This benchmarked code and a set of equations have been applied to a slot divertor for DEMO reactor. The slot length for the divertor can be determined by producing enough frictional force to overcome the thermal force which creates a backflow of impurities. A small amount of radiating argon is enough to exhaust 40 MW of power from the core plasma in a 40 cm long slot for the inner divertor of DEMO.

High power combined NBI + ICRF heating experiments have been carried out in the JET Mark IIa divertor configuration in the hot ion ELM-free H mode regime, both in deuterium (DD) and in deuterium-tritium (DT) plasmas. Results are presented from a wide range of additional heating power levels, ICRF up to 9.5 MW tuned to the fundamental hydrogen minority, NBI up to 22 MW, and for plasma currents up to 4.2 MA and toroidal fields up to 3.6 T. Discharges with combined NBI + ICRF heating show a clear improvement in electron temperature, DD neutron yield and stored energy with respect to NBI only discharges. High energy neutral particle analyser data show that acceleration of the NBI deuterons takes place due to absorption of ICRF power at the second harmonic deuterium resonance. This is confirmed by numerical simulations with the PION code, indicating that up to 40% of the ICRF power is absorbed by bulk and NBI ions. ICRF heating has been an essential ingredient in the DT experiments in the ELM-free hot ion regime, contributing to the achievement of a record fusion power of 16.1 MW and a record stored energy of 17 MJ.

Report on the IAEA Technical Committee Meeting held at Hefei, China 13-15 October 1998.

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