Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

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.

High harmonic fast wave heating and current drive (CD) are being developed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 41, 1435 (2001)] for supporting startup and sustainment of the spherical torus plasma. Considerable enhancement of the core heating efficiency (η) from 44% to 65% has been obtained for CD phasing of the antenna (strap-to-strap ϕ=−90°, kϕ=−8m−1) by increasing the magnetic field from 4.5to5.5kG. This increase in efficiency is strongly correlated to moving the location of the onset density for perpendicular fast wave propagation (nonset∝B×k∥2∕ω) away from the antenna face and wall, and hence reducing the propagating surface wave fields. Radio frequency (RF) waves propagating close to the wall at lower B and k∥ can enhance power losses from both the parametric decay instability (PDI) and wave dissipation in sheaths and structures around the machine. The improved efficiency found here is attributed to a reduction in the latter, as PDI losses are little changed at the higher magnetic field. Under these conditions of higher coupling efficiency, initial measurements of localized CD effects have been made and compared with advanced RF code simulations.

Lower hybrid (LH) waves (Ωci ω Ωce, where Ωi,e ≡ Zi,eeB/mi,ec) have the attractive property of damping strongly via electron Landau resonance on relatively fast tail electrons and consequently are well-suited to driving current. Established modeling techniques use WKB expansions with self-consistent non-Maxwellian distributions. Higher order WKB expansions have shown some effects on the parallel wavenumber evolution and consequently on the damping due to diffraction [G. Pereverzev, Nucl. Fusion 32, 1091(1991

Full wave simulations of fusion plasmas show a direct correlation between the location of the fast-wave cut-off, radiofrequency (RF) field amplitude in the scrape-off layer (SOL) and the RF power losses in the SOL observed in the National Spherical Torus eXperiment (NSTX). In particular, the RF power losses in the SOL increase significantly when the launched waves transition from evanescent to propagating in that region. Subsequently, a large amplitude electric field occurs in the SOL, driving RF power losses when a proxy collisional loss term is added. A 3D reconstruction of absorbed power in the SOL is presented showing agreement with the RF experiments in NSTX. Loss predictions for the future experiment NSTX-Upgrade (NSTX-U) are also obtained and discussed.

Fast Wave (FW) studies of mode conversion (MC) processes at the ion-ion hybrid layer in toroidal plasmas must capture the disparate scales of the FW and mode converted ion Bernstein (IBW) and ion cyclotron waves (ICW). Correct modeling of the MC layer requires resolving wavelengths on the order of k ⊥ ρi ∼ 1 which leads to a scaling of the maximum poloidal mode number, Mmax, proportional to 1/ρ * (ρ * ≡ ρi/L). The computational resources needed a scale with the number of radial (Nr), poloidal (N θ ), and toroidal (N φ ) elements as Nr * N φ * N

A significant fraction of high-harmonic fast-wave (HHFW) power applied to NSTX can be lost to the scrape-off layer (SOL) and deposited in bright and hot spirals on the divertor rather than in the core plasma. We show that the HHFW power flows to these spirals along magnetic field lines passing through the SOL in front of the antenna, implying that the HHFW power couples across the entire width of the SOL rather than mostly at the antenna face. This result will help guide future efforts to understand and minimize these edge losses in order to maximize fast wave heating and current drive. * Electronic address: rperkins@pppl.gov

Abstract. Recent results obtained with high harmonic fast wave (HHFW) heating and current drive (CD) on NSTX strongly support the hypothesis that the onset of perpendicular fast wave propagation right at or very near the launcher is a primary cause for a reduction in core heating efficiency at long wavelengths that is also observed in ICRF heating experiments in numerous tokamaks. A dramatic increase in core heating efficiency was first achieved in NSTX L-mode helium majority plasmas when the onset for perpendicular wave propagation was moved away from the antenna and nearby vessel structures. Efficient core heating in deuterium majority L mode and H mode discharges, in which the edge density is typically higher than in comparable helium majority plasmas, was then accomplished by reducing the edge density in front of the launcher with lithium conditioning and avoiding operational points prone to instabilities. These results indicate that careful tailoring of the edge density profiles in ITER should be considered to limit rf power losses to the antenna and plasma facing materials. Finally, in plasmas with reduced rf power losses in the edge regions, the first direct measurements of high harmonic fast wave current drive were obtained with the motional Stark effect (MSE) diagnostic. The location and radial dependence of HHFW CD measured by MSE are in reasonable agreement with predictions from both full wave and ray tracing simulations.