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Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

Time-resolved high-resolution soft x-ray spectra from gas-puff injected Ar impurity ions have been investigated for neutral beam heated and ohmically heated discharges in the TEXTOR tokamak. The experimental spectra show systematic deviations from corona model calculations for the line intensities of the forbidden He-like lines x, y, z and the Li- and Be-like dielectronic satellite spectra: theoretical corona model calculations predict intensities significantly too low. High-intensity Li-like inner-shell excited satellites correlate with the neutral beam injection. The discrepancies could also be observed in the stationary phase of an inductively heated discharge. In the heating phase the discrepancies are even larger. We propose charge-exchange processes between the neutral atoms and the impurity ions as an explanation of the experimental findings. Good agreement with the experimental observations can then be obtained without the need for invoking large (anomalous) diffusion coefficients. A self-consistent coupling of the population kinetics of the neutrals and the impurity ions, also taking into account charge-exchange processes from excited states of hydrogen/deuterium permit the determination of the neutral fraction and of the electron lifetime on the sole basis of impurity spectra analysis. Independent Monte Carlo simulations of neutral gas transport also provides the ionization degree in the centre and the electron lifetime. These calculations are also in good agreement with the spectroscopic results.

Tritium and helium gas-puff experiments have been conducted on TFTR to investigate transport in reverse shear plasmas. Small amounts of tritium and helium gas are puffed into the equilibrium phase of these plasmas. Measurements of the 14 MeV neutron fusion product and the charge-exchange spectrum of helium reveal the transport details of tritium and helium, respectively. The inferred tritium and helium density profiles indicate the formation of a core transport barrier for each species in reverse shear plasmas with enhanced confinement.

Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit.

X-ray imaging crystal spectrometers with high spectral and spatial resolution are currently being used on magnetically confined fusion devices to infer the time history profiles of ion and electron temperatures as well as plasma flow velocities. The absolute measurement of flow velocities is important for optimizing various discharge scenarios and evaluating the radial electric field in tokamak and stellarator plasmas. Recent studies indicate that the crystal temperature must be kept constant to within a fraction of a degree to avoid changes of the interplanar 2d-spacing by thermal expansion that cause changes in the Bragg angle, which could be misinterpreted as Doppler shifts. For the instrumental parameters of the x-ray crystal spectrometer on Alcator C-Mod, where those thermal effects were investigated, a change of the crystal temperature by 1 °C causes a change of the lattice spacing of the order of Δd = 1 × 10−5 A introducing a fictitious velocity drift of the order of ~3 km s−1. This effect must be considered for x-ray imaging crystals spectrometers installed on LHD, KSTAR, EAST, J-TEXT, NSTX and, in the future, W7-X and ITER.

A high-resolution imaging x-ray crystal spectrometer is described for implementation on the EAST tokamak to provide spatially and temporally resolved data on the ion temperature, electron temperature and poloidal plasma rotation. These data are derived from observations of the satellite spectra of helium-like argon, Ar XVII, which is the dominant charge state for electron temperatures in the range from 0.4 to 3.0 keV and which is accessible to EAST. Employing a novel design, which is based on the imaging properties of spherically bent crystals, the spectrometers will provide spectrally and spatially resolved images of the plasma for all experimental conditions, which include ohmically heated discharges as well as plasmas with rf and neutral-beam heating. The experimental setup and initial experimental results are presented.

During the 1987 run, TFTR reached record values of QDD, neutron source strength, and Ti(0). Good confinement together with intense auxiliary heating has resulted in a plasma pressure greater than 3*105 Pascals on axis, which is at the ballooning stability boundary. At the same time improved diagnostics, especially ion temperature profile measurements, have led to increased understanding of tokamak confinement physics. Ion temperature profiles are much more peaked than previously thought, implying that ion thermal diffusivity, even in high ion temperature supershot plasmas, is greater than electron thermal diffusivity. Based on studies of the effect of beam orientation on plasma performance, one of the four neutral beamlines has been re-oriented from injecting co-parallel to counter parallel, which will increase the available balanced neutral injection power from 14 MW to 27 MW. With this increase in balanced beam power, and the addition of 7 MW of ICRF power it is planned to increase the present equivalant QDT of 0.25 to close to break-even conditions in the coming run.

TFTR pellet injection results reported by Schmidt (1985) have been extended to higher density and n tau in plasmas limited by a graphite inner-wall belt limiter. Increased pellet penetration and larger density increases were achieved by operation at reduced plasma current (1.6 MA) and minor and major radius (70 cm and 235 cm). Under these conditions, beam heating results have been extended to 7 MW.

The characteristics of plasma operation on the axisymmetric inner toroidal limiter in TFTR are described. After conditioning, plasmas with low metal content and low zeff are obtained with this limiter. There is no substantial increase in zeff with total input power during neutral beam injection. Compared to operation on the outer blade limiter, additional gas is required to fuel plasmas on the inner limiter. Injection of D pellets increased the plasma density substantially and produced energy confinement times up to 0.5 s in ohmically heated plasmas. The four neutral beam lines have injected up to 13.5 MW total power into the plasma for 0.5 s with up to 90 kV accelerating voltage. The scaling of the plasma stored energy was studied as a function of the input power, plasma current and plasma density. In the range 1.4 to 2.2 MA, the overall and incremental confinement times for both the total and thermal stored energies increase with plasma current at fixed density. There appears to be a weak negative scaling of the total stored energy with density at high injection power.

New long-pulse ion sources have been employed to extend the neutral beam pulse on TFTR from 0.5 sec to 2.0 sec. This made it possible to study the long-term evolution of supershots at constant current and to perform experiments in which the plasma current was ramped up during the heating pulse. Experiments were conducted with co and counter injection as well as with nearly balanced injection of deuterium beams up to a total power of 20 MW. The best results, i.e., central ion temperatures Tio > 25 keV and neo τE Tio values of 3 × 1020 keV sec m-3, were obtained with nearly balanced injection. The central toroidal plasma rotation velocity scales in a linear-offset fashion with beam power and density. The scaling of the inferred global momentum confinement time with plasma parameters is inconsistent with the predictions of the neoclassical theory of gyroviscous damping. An interesting plasma regime with properties similar to the H-mode has been observed for limiter plasmas with edge qa just above 3 and 2.5.