Large sub-millisecond heat pulses due to Type-I edge localized modes (ELMs) have been eliminated reproducibly in DIII-D for periods approaching nine energy confinement times (τ E ) with small dc currents driven in a simple magnetic perturbation coil. The current required to eliminate all but a few isolated Type-I ELM impulses during a coil pulse is less than 0.4% of plasma current. Based on magnetic field line modelling, the perturbation fields resonate with plasma flux surfaces across most of the pedestal region (0.9 ψ N 1.0) when q 95 = 3.7 ± 0.2, creating small remnant magnetic islands surrounded by weakly stochastic field lines. The stored energy, β N , H-mode quality factor and global energy confinement time are unaltered by the magnetic perturbation. Although some isolated ELMs occur during the coil pulse, long periods free of large Type-I ELMs ( t > 4-6 τ E ) have been reproduced numerous times, on multiple experimental run days in high and intermediate triangularity plasmas, including cases matching the baseline ITER scenario 2 flux surface shape. In low triangularity, lower single null plasmas, with collisionalities near that expected in ITER, Type-I ELMs are replaced by small amplitude, high frequency Type-II-like ELMs and are often accompanied by one or more ELM-free periods approaching 1-2 τ E . Large Type-I ELM impulses represent a severe constraint on the survivability of the divertor target plates in future burning plasma devices. Results presented in this paper demonstrate that non-axisymmetric edge magnetic perturbations provide a very attractive development path for active ELM control in future tokamaks such as ITER.

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High temporal and spatial resolution measurements in the boundary of the DIII-D tokamak show that edge-localized modes ͑ELMs͒ are produced in the low field side, are poloidally localized and are composed of fast bursts ͑ϳ20 to 40 s long͒ of hot, dense plasma on a background of less dense, colder plasma ͑ϳ5 ϫ 10 18 m −3 , 50 eV͒ possibly created by the bursts themselves. The ELMs travel radially in the scrape-off layer ͑SOL͒, starting at the separatrix at ϳ450 m / s, and slow down to ϳ150 m / s near the wall, convecting particles and energy to the SOL and walls. The temperature and density in the ELM plasma initially correspond to those at the top of the density pedestal but quickly decay with radius in the SOL. The temperature decay length ͑ϳ1.2 to 1.5 cm͒ is much shorter than the density decay length ͑ϳ3 to 8 cm͒, and the latter decreases with increasing pedestal ͑and SOL͒ density. The local particle and energy flux ͑assuming T i = T e ͒ at the midplane wall during the bursts are 10% to 50% ͑ϳ1 to 2ϫ 10 21 m −2 s −1 ͒ and 1% to 2% ͑ϳ20 to 30 kW/ m 2 ͒, respectively, of the LCFS fluxes, indicating that particles are transported radially much more efficiently than heat. Evidence is presented suggesting toroidal rotation of the ELM plasma in the SOL. The ELM plasma density and temperature increase linearly with discharge/pedestal density up to a Greenwald fraction of ϳ0.6, and then decrease resulting in more benign ͑grassier͒ ELMs.

A plasma initiation and current ramp up scenario envisioned for ITER has been simulated in DIII-D experiments. These discharges were limited on the low field side (LFS) during the initial current ramp up, as specified for the ITER baseline startup scenario. Initial experiments produced internal inductance (ℓi), higher than the design value for the ITER shaping coils, often leading to vertical instabilities. A modified startup with larger volume was developed to reduce ℓi in the current ramp up. This large-bore scenario, also limiting on the LFS, produced a lower ℓi and avoided the vertical instabilities. Feedback control of ℓi, using the ohmic field coil power supply as the actuator, was successfully demonstrated. Such control may be useful in avoiding vertical instabilities and in providing access to sawtooth-free steady state and hybrid scenarios in ITER. Experiments at reduced inductive voltage and with electron cyclotron assist for breakdown and burnthrough have also been carried out. The Corsica equilibrium and transport code has modelled these data to provide validation of transport models used to simulate this phase of ITER discharges in order to yield more accurate extrapolation to ITER scenarios.