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In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βtwith nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

In inertial fusion energy (IFE) power plant designs, the fuel is a spherical layer of frozen DT contained in a target that is injected at high velocity into the reaction chamber. For direct drive, typically laser beams converge at the centre of the chamber (CC) to compress and heat the target to fusion conditions. To obtain the maximum energy yield from the fusion reaction, the frozen DT layer must be at about 18.5 K and the target must maintain a high degree of spherical symmetry and surface smoothness when it reaches the CC. During its transit in the chamber the cryogenic target is heated by radiation from the hot chamber wall. The target is also heated by convection as it passes through the rarefied fill-gas used to control chamber wall damage by x-rays and debris from the target explosion. This article addresses the temperature limits at the target surface beyond which target uniformity may be damaged. It concentrates on direct drive targets because fuel warm up during injection is not currently thought to be an issue for present indirect drive designs and chamber concepts. Detailed results of parametric radiative and convective heating calculations are presented for direct-drive targets during injection into a dry-wall reaction chamber. The baseline approach to target survival utilizes highly reflective targets along with a substantially lower chamber wall temperature and fill-gas pressure than previously assumed. Recently developed high-Z material coatings with high heat reflectivity are discussed and characterized. The article also presents alternate target protection methods that could be developed if targets with inherent survival features cannot be obtained within a reasonable time span.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

Current driven by fast Alfven waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma.

Progress in theory and in tokamak experiments leads to questions of the optimal development path for commercial tokamak power plants. The economic prospects of future designs are compared for several tokamak operating modes: (high poloidal beta) first stability, second stability and reverse shear. Using a simplified economic model and selecting uniform engineering performance parameters, this comparison emphasizes the different physics characteristics - stability and non-inductive current drive - of the various equilibria. The reverse shear mode of operation is shown to offer the lowest cost of electricity for future power plants.

The magnetic field of a low aspect ratio tokamak (LART) deviates rather significantly from the approximately 1/R dependence of conventional tokamaks. The parameterization of current drive efficiency on the basis of conventional large aspect ratio tokamaks becomes inaccurate. The physics of high harmonic ion cyclotron fast wave current drive (FWCD) and low frequency FWCD in LARTs is examined using exact evaluations of the current drive efficiencies from the adjoint technique. The results indicate that the high harmonic ion cyclotron current drive cannot penetrate to the plasma core at the high beta values typical of LART reactors, and hence this scheme may only be applicable for off-axis steady state current drive, while the low frequency current drive penetrates well into the plasma core and is appropriate for on-axis current profile control. The implications for advanced scenarios of LART are discussed

The combined theory of ion cyclotron wave coupling, mode conversion and absorption is developed for bounded inhomogeneous plasmas in slab geometry. The plasma response to the RF waves is described by the linearized Vlasov equation, with an equilibrium distribution function and a poloidal magnetic field profile which are selfconsistent solutions of the equilibrium Vlasov and Maxwell's equations. A general form of the wave equation, valid to arbitrary order in Larmor radius and including first order gradient terms, is derived. Expressions are also derived for the plasma intrinsic impedance. The wave equation, keeping second order terms in Larmor radius, is solved numerically for a plasma bounded by two conducting walls using the INTOR tokamak design parameters. It was found that the effect of the one-dimensional poloidal field is small. Also, a rough estimate of the error due to truncation to second order in Larmor radius was made and was found to be very small except in the high temperature regions. The effects of plasma boundedness and reflections from the mode conversion region on the coupling of waves from the RF source as a function of the plasma parameters and antenna spectra are studied. It is found that, at certain values of the plasma and antenna parameters, eigenmodes are formed, resulting in an increase in the loading impedance. It is verified that the impedance is not symmetric with respect to the spectrum in the poloidal direction, with highly peaked resonances. As the plasma temperature is increased, the resonances broaden and eventually disappear at very high temperatures. This is attributed to increased single pass absorption and decreasing reflection from the mode conversion region.

Supplementary wave heating is examined for two-ion-species plasmas from the fundamental tritium cyclotron resonance (ω = ωCT) to the first deuteron harmonic (ω = 2ωCD). At the first harmonic, it is found that a sufficiently large parallel wavelength greatly reduces the region of confluence between the fast wave and ion Bernstein wave. Power absorption considerations neglecting mode conversion processes show that first-harmonic heating efficiency is relatively insensitive to the species concentrations and that a dominant ion heating can be expected. For minority heating at the fundamental deuteron resonance, it is shown that inclusion of the two-ion hybrid resonance absorption considerably lowers the Q from the case where only minority species cyclotron damping is considered.

Analytical and numerical results of physical processes taking place around the second-harmonic resonance surface in ICRF heating are presented. It is shown that (1) symmetry of transmission coefficients follow from Onsager's reciprocity relation of the dielectric tensor, and (2) direct dissipation around the cyclotron harmonic layer is mostly due to the Bernstein branch and depends on k||, becoming low for low k||. The latter has the consequence that mode conversion and reflection are sensitively reduced by damping, but transmission is not.

Studies of radiofrequency (RF) coupling and heating to ignition of a conceptual tokamak fusion reactor by means of the fast magnetosonic wave at the first ion cyclotron harmonic (ω = 2ωci) arepresented. First, the current status of fast magnetosonic wave propagation and heating is briefly reviewed. Next, a spatially averaged, time-dependent start-up model is used to describe the role of RF heating for ignition of a tokamak reactor. The model shows that 240 MJ corresponding to an RF power level of 80 MW for a three-second period is quite sufficient to ignite a 700 MW(e) fusion power plant the size of NUWMAK. To couple the RF power to the torus, both external cavities coupled by apertures and poloidal coil systems are considered. Wave coupling from an external vacuum-filled cavity to fast magnetosonic waves via apertures provides matching through an impedance transformation from a co-axial feed to the plasma wave modes.

A simple two-dimensional model for evaluating the efficiency of ICRF coupling to a slab plasma is developed. The model takes advantage of the very low plasma surface impedance associated with the propagating magnetosonic wave to obtain lowest order analytical expressions for the surface RF magnetic field. The calculations are used to compare the loading characteristics of a simple loop antenna with a cavity resonator for high power ICRF heating.