Progress towards high-performance steady-state operation on DIII-D

Abstract

Advanced Tokamak research in DIII-D seeks to develop a scientific basis for steady-state high-performance tokamak operation. Fully noninductive (f NI ≈ 100%) in-principle steady-state discharges have been maintained for several confinement times. These * Corresponding author. Greenfield et al. / Fusion Engineering and Design 81 (2006) [2807][2808][2809][2810][2811][2812][2813][2814][2815] plasmas have weak negative central shear with q min ≈ 1.5-2, β N ≈ 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 f NI ≈ 90-95% and duration up to 2 s, limited only by hardware. In other experiments, β N ≈ 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 β N > 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.

Conference summary: progress in experiments on innovative concepts, confinement and performance

“…In DIII-D the current profile is broadened by a combination of ECCD and a toroidal magnetic field ramp for improved coupling with the wall and active control coils, and β N ∼ 4 ∼ 6 l i (internal inductance) has been maintained for 2 s with ITBs, well above the no-wall stability limit [48]. Recent progress on DIII-D is summarized in figure 20 where the fusion ignition figure of merit G is plotted versus the bootstrap current fraction.…”

“…The necessity to sustain the current in confinement devices based on net toroidal current such as tokamaks represents a primary limitation in view of long pulse (continuous) and steady-state operation. To prepare for the next generation of steady-state superconducting tokamak experiments such as ITER several present experiments (Tore Supra [53], TRIAM-1M [54], HT-7 [55,56], DIII-D [32,48], JET [49,57], NSTX [58], JT-60U [23]) are therefore addressing key technologies and the physics of non-inductive current drive scenarios, particle control, actively cooled plasma facing components, diagnostics and real-time plasma control using various sets of control tools.…”

“…DIII-D being equipped with a set of control tools to allow precise control of key aspects of plasma stability, transport, and current drive, its approach to steady-state operation is to maximize fusion power and bootstrap current (bootstrap current fraction f BS = 50-75%) by operating at high normalized beta β N 3.5 and safety factor q min > 1.5, and provide the remaining current with NB (NBCD), ECCD and fast wave (FW) current drive. In this approach two scenarios have been investigated [48], one exhibiting high fusion performance, G 0.7, but not stationary and another reaching fully non-inductive conditions (non-inductive current fraction f NI ∼ 100%) at G ∼ 0.3 for several confinement times (or about τ R /2), as foreseen in the ITER Q = 5 steadystate scenario.…”