R. Prater

Heating and current drive by electron cyclotron waves

The physics model of electron cyclotron heating (ECH) and current drive (ECCD) is becoming well validated through systematic comparisons of theory and experiment. This work has shown that ECH and ECCD can be highly localized and robustly controlled in toroidal plasma confinement systems, leading to applications including stabilization of magnetohydrodynamic instabilities like neoclassical tearing modes, control and sustainment of desired profiles of current density and plasma pressure, and studies of localized transport in laboratory plasmas. The experimental work was supported by a broad base of theory based on first principles which is now well encapsulated in linear ray tracing codes describing wave propagation, absorption, and current drive and in fully relativistic quasilinear Fokker–Planck codes describing in detail the response of the electrons to the energy transferred from the wave. The subtle balance between wave-induced diffusion and Coulomb relaxation in velocity space provides an understandin...

Low voltage Ohmic and electron cyclotron heating assisted startup in DIII-D

There is considerable interest in the development of low voltage startup scenarios for large tokamaks since it is proposed that in ITER the electric field which will be applied for ionization and plasma current ramp-up will be limited to values of E ≤ 0.3 V/m. Studies of low voltage startup have been carried out in DIII-D with and without electron cyclotron preionization and preheating. Successful Ohmic startup has been achieved with E ~ 0.25 V/m by paying careful attention to error fields and prefill pressure, while electron cyclotron heating (ECH) assisted startup with E ~ 0.15 V/m has been demonstrated. ECH assisted startup gives improved reliability at such low electric fields and permits operation over an extended range of prefill pressures and error magnetic fields. Using ECH, startup at E = 0.3 V/m with |B⊥| > 50 G over most of the vessel cross-section has been demonstrated. Such an error field represents an increase by more than a factor of two over the highest value for which Ohmic startup was achieved at the same electric field. During low voltage Ohmic startup with extreme values of prefill pressure and/or error magnetic fields, excessive breakdown delays are observed. The experimental data agree well with theoretical predictions based on the Townsend avalanche theory. ECH assisted startup is always prompt. The primary effect of ECH during the plasma current ramp-up is a decrease of the resistive component of the loop voltage Vrcs. A significant reduction (~30%) in Vres is achieved for low ECH powers (PRF ~ 300-400 kW), but a further large increase in PRF results in only a modest additional decrease in Vres. ECH was not applied over the whole ramp-up phase in these experiments and produced a reduction in volt-second consumption up to the current flat-top (Ip ~ 1 MA) of 10%. These experiments confirm that the low electric fields specified in the ITER design are acceptable and demonstrate the substantial benefits which accrue from the use of ECH assisted startup.

L-mode validation studies of gyrokinetic turbulence simulations via multiscale and multifield turbulence measurements on the DIII-D tokamak

A series of carefully designed experiments on DIII-D have taken advantage of a broad set of turbulence and profile diagnostics to rigorously test gyrokinetic turbulence simulations. In this paper the goals, tools and experiments performed in these validation studies are reviewed and specific examples presented. It is found that predictions of transport and fluctuation levels in the mid-core region (0.4 < ρ < 0.75) are in better agreement with experiment than those in the outer region (ρ ≥ 0.75) where edge coupling effects may become increasingly important and multiscale simulations may also be necessary. Validation studies such as these are crucial in developing confidence in a first-principles based predictive capability for ITER.

Nondimensional transport scaling in DIII‐D: Bohm versus gyro‐Bohm resolved

The scaling of cross‐field heat transport with relative gyroradius ρ* was measured in low (L) and high (H) mode tokamak plasmas using the technique of dimensionally similar discharges. The relative gyroradius scalings of the electron and ion thermal diffusivities were determined separately using a two‐fluid transport analysis. For L‐mode plasmas, the electron diffusivity scaled as χe∝χBρ1.1±0.3* (gyro‐Bohm‐like) while the ion diffusivity scaled as χi∝χBρ−0.5±0.3* (worse than Bohm‐like). The results were independent of the method of auxiliary heating (radio frequency or neutral beam). Since the electron and ion fluids had different gyroradius scalings, the effective diffusivity and global confinement time scalings were found to vary from gyro‐Bohm‐like to Bohm‐like depending upon whether the electron or ion channel dominated the heat flux. This last property can explain the previously disparate results with dimensionally similar discharges on different fusion experiments that have been published. Experimen...

Advanced techniques for neoclassical tearing mode control in DIII-D

Two techniques were developed at DIII-D [J. L. Luxon, Nucl. Fusion 42, 64 (2002)] to tackle ITER-specific aspects of neoclassical tearing mode (NTM) control, namely, (1) the relatively small size of the rotating islands, smaller than the electron cyclotron current drive (ECCD) deposition region, and (2) the increased tendency of the islands, compared to present devices, to lock to the wall or to the residual error field, in a position not necessarily accessible to ECCD. Modulated ECCD is known to suppress small islands more efficiently, when “broad,” than continuous ECCD. At DIII-D, a NTM of poloidal/toroidal mode numbers m/n=3/2 was completely stabilized by a new technique where oblique electron cyclotron emission acted at the same time as an indicator of good alignment between ECCD and the island, and as a waveform generator, for modulation in synch and in phase with the island O-point. In another experiment, after locking in an unfavorable position, a 2/1 island was steered by externally generated magn...

Progress toward fully noninductive, high beta conditions in DIII-D

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 diagno...

ELECTRON CYCLOTRON CURRENT DRIVE EFFICIENCY IN GENERAL TOKAMAK GEOMETRY

Green’s-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasilinear rf diffusion operator describes wave–particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the noninductive current drive of electron cyclotron waves.

Active control for stabilization of neoclassical tearing modes

This work describes active control algorithms used by DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to stabilize and maintain suppression of 3/2 or 2/1 neoclassical tearing modes (NTMs) by application of electron cyclotron current drive (ECCD) at the rational q surface. The DIII-D NTM control system can determine the correct q-surface/ECCD alignment and stabilize existing modes within 100–500ms of activation, or prevent mode growth with preemptive application of ECCD, in both cases enabling stable operation at normalized beta values above 3.5. Because NTMs can limit performance or cause plasma-terminating disruptions in tokamaks, their stabilization is essential to the high performance operation of ITER [R. Aymar et al., ITER Joint Central Team, ITER Home Teams, Nucl. Fusion 41, 1301 (2001)]. The DIII-D NTM control system has demonstrated many elements of an eventual ITER solution, including general algorithms for robust detection of q-surface/ECCD alignment and for real-time maintenance of alignment ...

Advances in validating gyrokinetic turbulence models against L- and H-mode plasmas a)

Robust validation of predictive turbulent transport models requires quantitative comparisons to experimental measurements at multiple levels, over a range of physically relevant conditions. Toward this end, a series of carefully designed validation experiments has been performed on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to obtain comprehensive multifield, multipoint, multiwavenumber fluctuation measurements and their scalings with key dimensionless parameters. The results of two representative validation studies are presented: an elongation scaling study performed in beam heated L-mode discharges and an electron heating power scan performed in quiescent H-mode (QH-mode) discharges. A 50% increase in the elongation κ is observed to lead to a ∼50% increase in energy confinement time τe and accompanying decrease in fluctuation levels, qualitatively consistent with a priori theoretical predictions and nonlinear GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] simulation...

Off-axis neutral beam current drive for advanced scenario development in DIII-D

Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, BT, and the plasma current, Ip, point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40–45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the BT direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.