Low-aspect-ratio tokamaks with toroidal currents, Ip, up to 250 kA are formed and sustained in the Helicity Injected Tokamak experiment [Nelson et al., Phys. Rev. Lett. 72, 3666 (1994)] using coaxial helicity injection. These plasmas are produced without use of a current drive transformer. Average toroidal currents are sustained at high values, 〈Ip〉=225 kA for 2 ms, where electron thermal energies are measured up to 80 eV with spectroscopy data suggesting burnthrough to the higher ionization states of oxygen. Currents can also be sustained for longer periods at lower values, 〈Ip〉=138 kA for 7 ms. These tokamaks are characterized by a rotating, n=1 distortion, with poloidal distortions approximately following the field line pitch, which only occur on the outer bad-curvature region. Equilibrium reconstructions show these plasmas have a tokamak q profile (q0=5 – 8, q95=10 – 12, qcyl≂3.6), with a hollow toroidal current profile and up to 170 kA of closed field toroidal current in a low-aspect-ratio, A=1.68 configuration.

Coaxial helicity injection is used to produce low-aspect-ratio tokamaks with toroidal currents reaching 150 kA (highest value yet attained by helicity injection current drive) and sustained over 100 kA for many resistive diffusion times, without a current drive transformer. Current drive power efBciency, assuming no anomalous helicity dissipation, is 40% that of Ohmic. These tokamaks have a rotating n = 1 toroidal distortion, with poloidal distortions only on the outer bad-curvature region.Equilibrium reconstruction suggests these plasmas have up to 112 kA of closed-Geld toroidal current, an aspect ratio A = 1.69, a tokamak q pro61e, and a hollow toroidal current pro6le.

The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.

Abstract. Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current / injector current) up to 10 were obtained, in agreement with predictions. The discharges which lasted for up to 200ms, limited only by the programmed waveform are more than an order of magnitude longer in duration that any CHI discharges previously produced in a Spheromak or a Spherical Torus (ST).

The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving β T and β N values of 17% and 5.7 (%m T MA −1 ), respectively. A newly available motional Stark effect diagnostic has enabled validation of currentdrive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable β in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.