Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the quiescent double barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent ELM-free H mode edge (termed QH mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH mode edge is characterized by ELM-free behaviour with continuous multiharmonic MHD activity in the pedestal region and has provided density and radiated power control for longer than 3.5 s (25τE) with divertor pumping. QDB plasmas are long pulse high performance candidates, having maintained a βN H89 product of 7 for five energy confinement times (Ti ≤ 16 keV, βN ≤ 2.9, H89 ≤ 2.4 τE ≤ 150 ms, DD neutron rate Sn ≤ 4 × 1015 s-1). The QDB regime has only been obtained in counter-NBI discharges (injection antiparallel to the plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation.

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5>E with qmin > 1.5. (The normalized performance is measured by the product βNH89, indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite qmin > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with βNH89 ≈ 7 for up to 6.3 s or ≈34τE. These discharges appear to have stationary current profiles with qmin ≈ 1.05 in agreement with the current profile relaxation time ≈1.8 s.

Fusion power has been increased by a factor of 3 in DIII-D by tailoring the pressure profile to avoid the kink instability in H-mode plasmas. The resulting plasmas are found to have neoclassical ion confinement. This reduction in transport losses in beam-heated plasmas with negative central shear is correlated with a dramatic reduction in density fluctuations. Improved magnetohydrodynamic stability is achieved by controlling the plasma pressure profile width. In deuterium plasmas the highest gain Q (the ratio of fusion power to input power), was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma. © 1996 The American Physical Society.

Strong shaping, favourable for stability and improved energy confinement, together with a significant expansion of the central region of improved confinement in negative central magnetic shear target plasmas, increased the maximum fusion power produced in DIII-D by a factor of 3. Using deuterium plasmas, the highest fusion power gain, the ratio of fusion power to input power, Q, was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma, which is similar to values achieved in tokamaks of larger size and magnetic field. A simple transformation relating Q to the stability parameters is presented.

Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-β advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper.