After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint is made available with the understanding that it will not be cited or reproduced without the permission of the author.

Abstract. Recent DIII-D experiments show that ideal kink modes can be stabilized at high beta by a resistive wall, with sufficient plasma rotation. However, the resonant response by a marginally stable resistive wall mode to static magnetic field asymmetries can lead to strong damping of the rotation. Careful reduction of such asymmetries has allowed plasmas with beta well above the ideal MHD nowall limit, and approaching the ideal-wall limit, to be sustained for durations exceeding one second. Feedback control can improve plasma stability by direct stabilization of the resistive wall mode or by reducing magnetic field asymmetry. Assisted by plasma rotation, direct feedback control of resistive wall modes with growth rates more than 5 times faster than the characteristic wall time has been observed. These results open a new regime of tokamak operation above the free-boundary stability limit, accessible by a combination of plasma rotation and feedback control.

Experimental evidence is reported of an internal kink instability driven by a new mechanism: barely trapped suprathermal electrons produced by off-axis electron cyclotron heating on the DIII-D tokamak. It occurs in plasmas with an evolving safety factor profile q(r) when q(min) approaches 1. This instability is most active when ECCD is applied on the high field side of the flux surface. It has a bursting behavior with poloidal/toroidal mode number = m/n = 1/1. In positive magnetic shear plasmas, this mode becomes the fishbone instability. This observation can be qualitatively explained by the drift reversal of the barely trapped suprathermal electrons.

This is a preprint of a paper to be submitted for publication in Phys. Rev. Lett.

A transport code (TRANSP) is used to simulate future deuterium-tritium (DT) experiments in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modelling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modelling. Fusion energy yields and 01 particle parameters are calculated, including profiles of the 01 slowing down time, the 01 average energy, and the AlfvBn speed and frequency. Two types of simulation are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the Do in the neutral beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that 01 heating will enhance the plasma parameters or that confinement will increase with the addition of tritium. The maximum relative fusion yield calculated for these simulations is QDT-0.3, and the maximum a contribution to the central toroidal 0 is PJO)-0.5%. The stability of toroidicity induced Alfvkn eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed.

A TFTR supershot with a plasma current of 2.5 MA, neutral beam heating power of 33.7 MW, and a peak DT fusion power of 7.5 MW is studied using the TRANSP plasma analysis code. Simulations of alpha parameters such as the alpha heating, pressure, and distributions in energy and Vparallel/ v are given. The effects of toroidal ripple and mixing of the fast alpha particles during the sawteeth observed after the neutral beam injection phase are modeled. The distributions of alpha particles on the outer midplane are peaked near fotward and backward Vparalel /

Recent DIII-D experiments have shown that the n=1 resistive wall mode (RWM) can be controlled by an external magnetic field applied in closed loop feedback using the six element error field correction coil (C-coil). The RWM constitutes the primary limitation to normalized beta in recent DIII-D advanced tokamak plasma experiments. The toroidal rotation of DIII-D plasmas does not seem sufficient to completely suppress the RWM: a very slowly growing RWM (growth rate γ « 1/τ w ) is often observed at normalized beta above the no-wall limit and this small RWM slows the rotation. As the rotation decreases, there is a transition to more rapid growth (γ ~ 1/τ w ). The application of magnetic feedback is able to hold the RWM to a very small amplitude, prolonging the plasma duration above the no-wall limit for durations much longer than the growth time of the RWM. These initial experimental results are being used to compare control algorithms, to benchmark models of the feedback stabilization process and to guide the design of an upgraded coil-sensor system for stabilization of the RWM at normalized beta values closer to the ideal-wall limit.