1 - Nuclear Sciences and Applications - IAEA

528 PRATER et al.
1. INTRODUCTION
Electron cyclotron heating (ECH) in tokamaks is an effective means
of heating plasma to high temperature, and this has been demonstrated in
a number of devices [1,2]. On the basis of this work and of the apparent
advantages of ECH to the design and engineering of a tokamak reactor,
ECH has been assigned a role in heating plasma in CIT'to ignition. The
principal modes of heating have been studied in experiments on DIII-D,
including outside launch of the ordinary mode (O-mode) at the fundamental
frequency (the mode chosen for CIT) and the extraordinary mode
(X-mode) at the fundamental and second harmonic, and inside launch
of the fundamental X-mode. Central electron temperatures above 5 keV
have been achieved.
In addition to bulk heating, ECH has potential applications affecting
confinement and stability. In DIII-D experiments, central application
of ECH has been shown to generate the H—mode of improved confinement
[3,4]; heating near the q = 1 surface has suppressed sawteeth [5];
and applying ECH with the resonance near the edge has stabilized the
edge localized modes (ELMs), leading to improved energy confinement
for plasmas in which the H-mode was generated by neutral injection.
These ECH experiments at 60 GHz have used two antenna systems.
For the outside launch, a system of eight antennas is located on the outer
wall near the plasma midplane. The horns radiate with a Gaussian pattern
with an 11° e-folding half width. Each horn radiates typically 150 kW in
the desired polarization and 25 kW in the cross polarization. For inside
launch, ten antennas have been placed on the inboard wall of the vacuum
vessel near the plasma midplane. These antennas are mirrors which are
located behind the graphite armor that covers the wall, and the power
is radiated through holes in the graphite. The radiation pattern is a
Gaussian 12° in width, with the center of the pattern oriented 15° from
radial in the same toroidal direction as the plasma current.
In the experiments described here, the DIII-D tokamak was run
in the divertor configuration, with major radius 1.68 m, minor radius
0.62 m, and elongation 1.7 (volume 21 m 3 ). In most cases, the working
gas was deuterium. Plasma energy W was determined from a full MHD
equilibrium analysis using magnetic measurements as input, or from measurements
of plasma diamagnetism. Uncertainty is typically 0.05 in /?„. A
multichannel Thomson scattering system is available for which the laser
beam travels vertically at a major radius of 1.94 m; for this system, the
data point closest to the center of the plasma lies near rfa « 0.3. A CO2
interferometer is used to determine line-averaged density along a vertical
chord at the same major radius. Electron cyclotron emission power at
the second harmonic is measured by an absolutely calibrated ten channel
polychrometer for determination of the electron temperature profile.
CIT: Compact Ignition Tokamak.