Highly peaked density and pressure profiles in a new operating regime have been observed on the Tokamak Fusion Test Reactor (TFTR). The qprofile has a region of reversed magnetic shear extending from the magnetic axis to r / u-0.3-0.4. The central electron density rises from 0.45 x lo2' m-3 to nearly 1.2 x lo2' m-' during neutral beam injection. The electron particle diffusivity drops precipitously in the plasma core with the onset of the improved confinement mode and can be reduced by a factor of N 50 to near the neoclassical particle diffusivity level.

Dissipation of plasma toroidal angular momentum is observed in the National Spherical Torus Experiment due to applied nonaxisymmetric magnetic fields and their plasma-induced increase by resonant field amplification and resistive wall mode destabilization. The measured decrease of the plasma toroidal angular momentum profile is compared to calculations of nonresonant drag torque based on the theory of neoclassical toroidal viscosity. Quantitative agreement between experiment and theory is found when the effect of toroidally trapped particles is included.

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The National Spherical Torus Experiment (NSTX) offers an operational space characterized by high-beta (β t = 39%, β N > 7, β N /β no-wall N > 1.5) and low aspect ratio (A > 1.27) to leverage the plasma parameter dependences of RWM stabilization and plasma rotation damping physics giving greater confidence for extrapolation to ITER. Significant new capability for RWM research has been added to the device with the commissioning of a set of six nonaxisymmetric magnetic field coils, allowing generation of fields with dominant toroidal mode number, n, of 1-3. These coils have been used to study the dependence of resonant field amplification on applied field frequency and RWM stabilization physics by reducing the toroidal rotation profile below its steady-state value through non-resonant magnetic braking. Modification of plasma rotation profiles shows that rotation outside q = 2.5 is not required for passive RWM stability and there is large variation in the RWM critical rotation at the q = 2 surface, both of which are consistent with distributed dissipation models.

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.

Progress in the development of integrated advanced ST plasma scenarios in NSTX (Ono et al 2000 Nucl. Fusion 40 557) is reported. Recent high-performance plasmas in NSTX following lithium coating of the plasma facing surfaces have achieved higher elongation and lower internal inductance than previously. Analysis of the thermal confinement in these lithiumized discharges shows a stronger plasma current and weaker toroidal field dependence than in previous ST confinement scaling studies; the ITER-98(y, 2) scaling expression describes these scenarios reasonably well. Analysis during periods free of MHD activity has shown that the reconstructed current profile can be understood as the sum of pressure driven, inductive and neutral beam driven currents, without requiring any anomalous fast-ion transport. Non-inductive fractions of 65–70%, and βP > 2, have been achieved at lower plasma current. Some of these low-inductance discharges have a significantly reduced no-wall βN limit, and often have βN at or near the with-wall limit. Coupled m/n = 1/1 + 2/1 kink/tearing modes can limit the sustained β values when rapidly growing ideal modes are avoided. A βN controller has been commissioned and utilized in sustaining high-performance plasmas. ‘Snowflake’ divertors compatible with high-performance plasmas have been developed. Scenarios with significantly larger aspect ratios have also been developed, in support of next-step ST devices. Overall, these NSTX plasmas have many characteristics required for next-step ST devices.

Wall conditioning in the Tokamak Fusion Test Reactor ͑TFTR͒ ͓K. M. McGuire et al., Phys. Plasmas 2, 2176 ͑1995͔͒ by injection of lithium pellets into the plasma has resulted in large improvements in deuterium-tritium fusion power production ͑up to 10.7 MW͒, the Lawson triple product ͑up to 10 21 m Ϫ3 s keV͒, and energy confinement time ͑up to 330 ms͒. The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through ͑1͒ distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and ͑2͒ injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun.

General plasma physics principles state that power flow Q(r) through a magnetic surface in a tokamak should scale as Q(r)= {32π2Rr3Te2c nea/[eB (a2−r2)2]} F(ρ*,β,ν*,r/a,q,s,r/R,...) where the arguments of F are local, nondimensional plasma parameters and nondimensional gradients. This paper reports an experimental determination of how F varies with normalized gyroradius ρ*≡(2TeMi)1/2c/eBa and collisionality ν*≡(R/r)3/2qRνe(me/ 2Te)1/2 for discharges prepared so that other nondimensional parameters remain close to constant. Tokamak Fusion Test Reactor (TFTR) [D. M. Meade et al., in Plasma Physics and Controlled Nuclear Fusion Research, 1990, Proceedings of the 13th International Conference, Washington (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] L-mode data show F to be independent of ρ* and numerically small, corresponding to Bohm scaling with a small multiplicative constant. By contrast, most theories predict gyro-Bohm scaling: F∝ρ*. Bohm scaling implies that the largest scale size for microinstability turbulence depends on machine size. Analysis of a collisionality scan finds Bohm-normalized power flow to be independent of collisionality. Implications for future theory, experiment, and reactor extrapolations are discussed.