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.

The experimental programme of Tore Supra, the largest superconducting tokamak in the world (a = 0.72 m, R = 2.4 m, Ip < 1.7 MA, BT < 4.5 T) was devoted in 2003 to studying the heat removal capability and particle exhaust in steady-state fully non-inductive current drive discharges simultaneously. This required both advanced technology integration and steady-state real-time plasma control. In particular, an improvement of the plasma position to within a range of few millimetres, and new real-time controls of radio frequency power and various actuators built around a shared memory network, have allowed Tore Supra to access a powerful steady-state regime with an improved safety level for the actively cooled plasma facing components. Feedback controlled fully non-inductive plasma discharges have been sustained in a steady-state regime for up to 6 min with a new world record for injected–extracted energy exceeding 1 GJ. Experimental results and an analysis of the physics involved in these discharges are reported and discussed.

Recent results on feedback control of the global shape of the current density profile in discharges with lower hybrid current drive (LHCD) on Tore Supra are presented. The global shape of the current density profile is characterized by its internal inductance li, and feedback controlled through the LH power or the launched LH wave spectrum. Feedback control of the flux on the plasma boundary has allowed for exploration of regimes with partial current drive (constant loop voltage) and full current drive (zero loop voltage). In stationary, steady state discharges at 0.8 MA (qa=7), the current profile shape is characterized by an li between 1.5 and 1.7 and a safety factor on-axis (q0) between 1 and 1.4. The energy confinement is 1.3 to 1.6 times higher than the value predicted by the Rebut-Lallia-Watkins scaling law