Utilizing an array of new diagnostics and simulation/modelling techniques, recent DIII-D experiments have elucidated a variety of energetic ion transport behaviour in the presence of instabilities ranging from large-scale sawteeth to fine spatial scale microturbulence. Important new insights include sawteeth, such as those of the ITER baseline scenario, causing major redistribution of the energetic ion population; high levels of transport induced by low-amplitude Alfvén eigenmodes can be caused by the integrated effect of a large number of simultaneous modes; and microturbulence can contribute to the removal of alpha ash while having little effect on fusion alphas. This paper provides an overview of recent and upcoming results from the DIII-D Energetic Particles research programme. © 2011 IAEA, Vienna.

In this paper we show that changing from an ion temperature gradient (ITG) to a trapped electron mode (TEM) dominant turbulence regime (based on linear gyrokinetic simulations) results experimentally in a strong density pump-out (defined as a reduction in line-averaged density) in low collisionality, low power H-mode plasmas. We vary the turbulence drive by changing the heating from predominantly ion heated using neutral beam injection to electron heated using electron cyclotron heating, which changes the ratio and the temperature gradients. Perturbed gas puff experiments show an increase in transport outside , through a strong increase in the perturbed diffusion coefficient and a decrease in the inward pinch. Linear gyrokinetic simulations with TGLF show an increase in the particle flux outside the mid-radius. In conjunction an increase in intermediate-scale length density fluctuations is observed, which indicates an increase in turbulence intensity at typical TEM wavelengths. However, although the experimental changes in particle transport agree with a change from ITG to TEM turbulence regimes, we do not observe a reduction in the core rotation at mid-radius, nor a rotation reversal.

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by 30% relative to earlier work (Politzer et al 2005 Nucl. Fusion 45 417). The advancement was enabled by improved understanding of the 'relaxation oscillations', previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the 'relaxation oscillations' are coupled core-edge modes amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced to classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ≥80%, βN ≤ 4, βP ≥ 3, and βT ≥ 2%. These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.

Energetic ion transport due to microturbulence is investigated in magnetohydrodynamic-quiescent plasmas by way of neutral beam injection in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. A range of on-axis and off-axis beam injection scenarios are employed to vary relevant parameters such as the character of the background microturbulence and the value of E b / T e, where Eb is the energetic ion energy and Te the electron temperature. In all cases, it is found that any transport enhancement due to microturbulence is too small to observe experimentally. These transport effects are modeled using numerical and analytic expectations that calculate the energetic ion diffusivity due to microturbulence. It is determined that energetic ion transport due to coherent fluctuations (e.g., Alfvén eigenmodes) is a considerably larger effect and should therefore be considered more important for ITER. © 2013 AIP Publishing LLC.

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.

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.

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, ΒT =3.6%, normalized beta, ΒN =3.5, and confinement factor, H89 =2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1 τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar, Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario. © 2006 American Institute of Physics.

Advanced Tokamak research in DIII-D seeks to develop a scientific basis for steady-state high-performance tokamak operation. Fully noninductive (fNI ≈ 100%) in-principle steady-state discharges have been maintained for several confinement times. These plasmas have weak negative central shear with qmin ≈ 1.5-2, βN ≈ 3.5, and large, well-aligned bootstrap current. The loop voltage is near zero across the entire profile. The remaining current is provided by neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD). Similar plasmas are stationary with fNI ≈ 90-95% and duration up to 2 s, limited only by hardware. In other experiments, βN ≈ 4 is maintained for 2 s with internal transport barriers, exceeding previously achieved performance under similar conditions. This is allowed by broadened profiles and active magnetohydrodynamic instability control. Modifications now underway on DIII-D are expected to allow extension of these results to higher performance and longer duration. A new pumped divertor will allow density control in high triangularity double-null divertor configurations, facilitating access to similar in-principle steady-state regimes with βN > 4. Additional current drive capabilities, both off-axis ECCD and on-axis fast wave current drive (FWCD), will increase the magnitude, duration, and flexibility of externally driven current. © 2006 Elsevier B.V. All rights reserved.

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.