J. Lohr

An in-line power monitor for HE/sub 11/ low loss transmission lines

A power monitor has been developed for the DIII-D 110 GHz EC transmission line, which allows for the measurement of power flowing in the transmission line before it reaches the launcher. The power monitor uses a small break in the transmission line to radiate power, which is then measured.

Electron heat transport in improved confinement discharges in DIII-D

In DIII-D tokamak plasmas with an internal transport barrier (ITB), the comparison of gyrokinetic linear stability (GKS) predictions with experiments in both low and strong negative magnetic shear plasmas provide improved understanding for ion and electron thermal transport within much of the plasma. As previously reported, the region for improved ion transport seems well characterized by the condition OE~B>Y-, where SERB is the ExB flow shear, calculated from measured quantities, and y,, is the maximum linear growth rate for ion temperature gradient (ITG) modes in the absence of flow shear. Within a limited region just inside the ITB, the electron temperature gradient (ETG) modes appear to control the electron temperature gradient and, consequently, the electron thermal transport. The increase in electron temperature gradient with more strongly negative magnetic shear is consistent with the increase in the ETG mode marginal gradient. Closer to the magnetic axis the Te profile flattens and the ETG modes are predicted to be stable. With additional core electron heating, FIR scattering measurements near the axis show the presence of high k fluctuations (12 cm-l), rotating in the electron diamagnetic drift direction. This turbulence could impact electron transport and possibly also ion transport. Thermal diffusivities for electrons, and to a lesser degree ions, increase. The ETG mode can exist at this wavenumber, but it is computed to be robustly stable near the axis.

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...

Toroidal rotation in DIII-D in electron cyclotron heating and Ohmic H-mode discharges

Spatially and temporally resolved toroidal rotation measurements have been made in DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] discharges with no externally applied torque. The velocity measurements are made using the charge exchange recombination (CER) technique viewing emission from the intrinsic carbon impurity in deuterium discharges. Three cases have been studied: L mode and H mode with Ohmic heating and H mode with electron cyclotron heating (ECH). The ECH H mode has carbon counter-rotation in the center of the plasma, and co-rotation outside, where co- and counter- are relative to the direction of the toroidal plasma current. The Ohmic H mode has carbon rotation everywhere in the co-direction. Neoclassical theory is applied to compute the deuterium toroidal velocity and it is found that the counter-rotation measured for carbon in the core of the ECH H mode is also thus predicted for the bulk deuterium species. Short blips of neutral beams (NB) must be used for the CER technique and these blip...

INVESTIGATION OF THE FORMATION OF A FULLY PRESSURE-DRIVEN TOKAMAK

A noninductive current drive concept, based on internal pressure‐driven currents in a low‐aspect‐ratio toroidal geometry, has been demonstrated on the Current Drive Experiment Upgrade (CDX‐U) [Forest et al., Phys. Rev. Lett. 68, 3559 (1992)] and further tested on DIII‐D [in Plasma Physics and Controlled Nuclear Fusion Research, 1986, Proceedings of the 11th International Conference, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159]. For both experiments, electron cyclotron power provided the necessary heating to breakdown and maintain a plasma with high‐βp and low collisionality (eβp∼1, ν*≤1). A poloidal vacuum field similar to a simple magnetic mirror is superimposed on a much stronger toroidal field to provide the initial confinement for a hot, trapped electron species. With application of electron cyclotron heating (ECH), toroidal currents spontaneously flow within the plasma and increase with applied ECH power. The direction of the generated current is independent of the toroid...

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.

Detailed measurements of the electron cyclotron current drive efficiency on DIII-D

Electron cyclotron current drive (ECCD) experiments on the DIII-D tokamak are solidifying the physics basis for localized, off-axis current drive, the goal being to validate a predictive model for ECCD. The ECCD profiles are determined from the magnetic field pitch angles measured by motional Stark effect polarimetry. The measured ECCD switches from the co- to the counter-direction as the toroidal injection angle is varied with a profile width that is in accordance with ray tracing calculations. Tests of electron trapping in low beta plasmas show that the ECCD efficiency decreases rapidly as the deposition is moved off-axis and towards the outboard side of the plasma, but the detrimental effects of electron trapping on the current drive are greatly reduced in high beta plasmas. Overall, the measured ECCD is in good agreement with theoretical calculations using a quasilinear Fokker–Planck code over a wide range of injection angles and plasma parameters.

Discharge improvement through control of neoclassical tearing modes by localized ECCD in DIII-D

A271 DISCHARGE IMPROVEMENT THROUGH CONTROL OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D. Neoclassical tearing modes (NTMs) are MHD modes which can limit the performance of high beta discharges in tokamaks, in some cases leading to a major disruption. The destabilizing effect which results in NTM growth is a helical decrease in the bootstrap current caused by a local reduction of the plasma pressure gradient by seed magnetic islands. The NTM is particularly well suited to control since the mode is linearly stable although nonlinearly unstable, so if the island amplitude can be decreased below a threshold size the mode will decay and vanish. One means of shrinking the island is the replacement of the missing bootstrap current by a localized current generated by electron cyclotron current drive (ECCD). This method has been applied to the m=3/n=2 neoclassical tearing mode in DIII-D, in H-mode plasmas with ongoing ELMs and sawteeth, both of which generate seed islands periodically. In the case of the 3/2 mode, full suppression was obtained robustly by applying about 1.5 MW of ECCD very near the rational surface of the mode. When the mode first appears in the plasma the stored energy decreases by 20%, but after the mode is stabilized by the ECCD the beta may be raised above the initial threshold pressure by 20% by additional neutral beam heating, thereby generating an improvement in the limiting beta of nearly a factor 2. An innovative automated search algorithm was implemented to find and retain the optimum location for the ECCD in the presence of the mode.

Feedback control of the safety factor profile evolution during formation of an advanced tokamak discharge

Active feedback control for regulation of the safety factor (q) profile at the start of the high stored energy phase of an advanced tokamak discharge has been demonstrated in the DIII-D tokamak. The time evolution of the on-axis or minimum value of q is controlled during and just following the period of ramp-up of the plasma current using electron heating to modify the rate of relaxation of the current profile. In L-mode and H-mode discharges, feedback control of q is effective with the appropriate choice of either off-axis electron cyclotron heating or neutral beam heating as the actuator. The q profile is calculated in real time from a complete equilibrium reconstruction fitted to external magnetic field and flux measurements and internal poloidal field measurements from the motional Stark effect diagnostic.