The development of techniques for neoclassical tearing mode (NTM) suppression or avoidance is crucial for successful high beta/high confinement tokamaks. Neoclassical tearing modes are islands destabilized and maintained by a helically perturbed bootstrap current and represent a significant limit to performance at higher poloidal beta. The confinement-degrading islands can be reduced or completely suppressed by precisely replacing the "missing" bootstrap current in the island O-point or by interfering with the fundamental helical harmonic of the pressure. Implementation of such techniques is being studied in the DIII-D tokamak [J.L. Luxon, et al., Plasma Phys. and Control. Fusion Research, Vol. 1 (International Atomic Energy Agency, Vienna, 1987) p. 159] in the presence of periodic q = 1 sawtooth instabilities, a reactor relevant regime. Radially localized off-axis electron cyclotron current drive (ECCD) must be precisely located on the island. In DIII-D the plasma control system is put into a "search and suppress" mode to make either small rigid radial position shifts of the entire plasma (and thus the island) or small changes in toroidal field (and thus, ECCD location) to find and lock onto the optimum position for complete island suppression by ECCD. This is based on real-time measurements of an m n = 3 2 mode amplitude dB dt θ . The experiment represents the first use of active feedback control to provide continuous, precise positioning. An alternative to ECCD makes use of the six toroidal section "C-Coil" on DIII-D to provide a large non-resonant static m = 1, n = 3 helical field to interfere with the fundamental harmonic of an m n = 3 2 NTM. While experiments show success in inhibiting the NTM if a large enough n = 3 field is applied before the island onset, there is a considerable plasma rotation decrease due to n = 3 "ripple".
Error field optimization on DIII-D tokamak [Luxon, J.L., Nucl. Fusion 42 (2002) 8141 plasma discharges has routinely been done for the last ten years with the use of the external " n = 1 coil" or the "C-coil". The optimum level of correction coil current is determined by the ability to avoid the locked mode instability and access previously unstable parameter space at low densities. The locked mode typically has toroidal and poloidal mode numbers n = 1 and m = 2, respectively, and it is this component that initially determined the correction coil current and phase. Realization of the importance of nearby n = 1 mode components m = 1 and in = 3 has led to a revision of the error field correction algorithm. Viscous and toroidal mode coupling effects suggested the need for additional terms in the expression for the radial "penetration" field B that can induce a locked mode. To incorporate these effects, the low density locked mode threshold database was expanded. A database of discharges at various toroidal fields, plasma currents, and safety factors was supplemented with data from an experiment in which the fields of the n = 1 coil and C-coil were combined, allowing the poloidal mode spectrum of the error field to be varied. A multivariate regression analysis of this new low density locked mode database was done to determine the low density locked mode threshold scaling relationship ne B7°'01q,"~79B,,e, and the coefficients of the poloidal mode components in the expression for B,,,, . Improved plasma performance is achieved by optimizing B by Pen varying the applied correction coil currents.Pen GENERAL
A comprehensive set of L–H transition experiments has been performed on DIII-D to determine the requirements for access to H-mode plasmas in ITER's first (non-nuclear) operational phase with H and He plasmas and the second (activated) operational phase with D plasmas. The H-mode power threshold, P TH, was evaluated for different operational configurations and auxiliary heating methods for the different main ion species. Helium plasmas have significantly higher P TH than deuterium plasmas at low densities for all heating schemes, but similar P TH as deuterium plasmas at high densities except for H-neutral beam injection-heated discharges, which are still higher. Changes in P TH are observed when helium concentration levels in deuterium plasmas exceed 40%. There is a strong dependence of P TH on the magnetic geometry in the vicinity of the divertor. The trend of decreasing P TH with decreasing X-point height is observed for all of the main ion species irrespective of the heating method, which appears to indicate that there is a common physics process behind this effect for all of the ion species. Helium and deuterium plasmas exhibit a significant increase in P TH for strong resonant magnetic perturbations. The application of a local magnetic ripple of 3% from test blanket module mock-up coils did not change P TH in deuterium plasmas.
Abstract. Recent DIII-D experiments show that ideal kink modes can be stabilized at high beta by a resistive wall, with sufficient plasma rotation. However, the resonant response by a marginally stable resistive wall mode to static magnetic field asymmetries can lead to strong damping of the rotation. Careful reduction of such asymmetries has allowed plasmas with beta well above the ideal MHD nowall limit, and approaching the ideal-wall limit, to be sustained for durations exceeding one second. Feedback control can improve plasma stability by direct stabilization of the resistive wall mode or by reducing magnetic field asymmetry. Assisted by plasma rotation, direct feedback control of resistive wall modes with growth rates more than 5 times faster than the characteristic wall time has been observed. These results open a new regime of tokamak operation above the free-boundary stability limit, accessible by a combination of plasma rotation and feedback control.
Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardwarerelated fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments 1 and resistive wall mode experiments 2 done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported 3 . Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments 1,2 . These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.
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