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The Korea Superconducting Tokamak Advanced Research (KSTAR) project is the major effort of the national fusion programme of the Republic of Korea. Its aim is to develop a steady state capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. The major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, with a strongly shaped plasma cross-section and double null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beams, ion cyclotron waves, lower hybrid waves and electron cyclotron waves for flexible profile control in advanced tokamak operating modes. A comprehensive set of diagnostics is planned for plasma control, performance evaluation and physics understanding. The project has completed its conceptual design and moved to the engineering design and construction phase. The target date for the first plasma is 2002.

Electron cyclotron (EC) power has technological and physics advantages for heating and current drive (CD) in a tokamak reactor, and advances in source development make it credible for applications in the International Thermonuclear Experimental Reactor (ITER). Strong single pass absorption makes heating to ignition in ITER particularly simple. At densities to 3.6*1020 m-3, with ohmic temperatures and wave frequency 170 GHz, heating in the plasma core is readily obtained. For outside launch of an ordinary mode (O mode) near the fundamental EC frequency, the optimized EC current drive (ECCD) efficiency ((n)IR/P) shows a linear temperature scaling at temperatures up to ~15 keV. For temperatures above 30 keV, the efficiency saturates at approximately 0.3*1020 A/W.m2 for a frequency of 220 GHz in an ITER target plasma with a toroidal field of 6 T, due primarily to harmonic overlap and to a lesser extent due to limitations arising from relativistic effects. The same efficiency can also be obtained at 170 GHz for the same plasma equilibrium and q profile except that the magnetic field is reduced to (170/220)*6 T=4.6 T. The ECCD efficiencies are obtained with the comprehensive 3-D, bounce averaged Fokker-Planck codes CQL3D and BANDIT3D

Inertial electrostatic confinement (IEC) systems are predicated on a nonequilibrium ion distribution function. Coulomb collisions between ions cause this distribution to relax to a Maxwellian on the ion–ion collisional time scale. The power required to prevent this relaxation and maintain the IEC configuration for times beyond the ion–ion collisional time scale is shown to be greater than the fusion power produced. It is concluded that IEC systems show little promise as a basis for the development of commercial electric power plants.

A steady-state tokamak experiment (STE) is needed to develop the technology and physics database required for construction of a steady-state fusion power demonstration reactor in the early 21st century. The STE will provide an integrated facility for the development and demonstration of steady-state and particle handling, low-activation, high-heat-flux components and materials, efficient current drive, and continuous plasma and impurity control while maintaining high-quality plasma performance in steady-state, with reactor-like plasma conditions under severe conditions of heat and particle bombardment of the wall. The STE facility will also be used to develop operation and control scenarios for ITER (International Thermonuclear Experimental Reactor). >

The goal for advanced steady-state operation in ITER should be to demonstrate the operation of the plasma core for a steady-state fusion reactor. To accomplish this the authors must develop steady-state operating scenarios at high beta for high fusion power density, low auxiliary power requirements (Q{sub CD} {ge} 25, where Q{sub CD} {triple_bond}P fusion/P{sub CD} and P{sub DC} is the power required for sustaining the plasma current) for low recirculating power requirements, and at moderate safety factor (q{sub {psi}} {le} 4.5) to minimize the cost for the tokamak core of a steady-state demonstration power reactor based on the operating modes demonstrated in ITER. The key to achieving steady-state operation at high fusion power in ITER will be the development of operating scenarios with very high bootstrap current fractions (f{sub BS} {ge} 90%) in which the radial profile of the bootstrap current density is well aligned with that of the total plasma current density, and for which the MHD {beta}-limit exceeds {beta}{sub n}{sup *} = 0.05 T{minus}m/MA. They are in the process of developing such operating modes for ITER. In {section}1 they propose two advanced steady-state operating points; a preliminary operating point that was the basis for the MHD studies reportedmore » in {section}2, and a second operating point that has been optimized based on the authors studies to date. In {section}2 they present calculations indicating that the initial operating point is stable to MHD ballooning and low-n kink modes (with a conducting wall at r = 1.25a) up to {beta}{sub n}{sup *} {approx} 6 {times} 10{sup {minus}2} T {minus} m/MA. In {section}3 they present a free-boundary MHD equilibrium, and show that advanced steady-state operating modes are compatible with the ITER poloidal field system and diverter.« less

The Steady State Advanced Tokamak is designed to develop and demonstrate optimized steady state operating modes. It will explore improvements in energy confinement and beta limit scaling in high-aspectratio plasmas with a high bootstrap current fraction. It will demonstrate advanced divertor operation techniques for steady state energy and particle removal. Key physics features include a plasma with strong shaping, a doublenull poloidal divertor, fully non-inductive current drive, and current profile control capability. The plasma heating and current drive systems includes neutral beams, ICRF, and lower hybrid. The engineering design incorporates a superconducting magnet system, actively cooled divertor and first wall systems, and in-vessel remote maintenance.

Slab models are employed in the simulation of microinstabilities in inhomogeneous plasma.(AIP)

Electrostatic collisionless drift waves in a sheared magnetic field have been studied extensively with use of particle-code simulations. When the shear is strong, we have found that the plasma is dominated by spatially uncorrelated small-amplitude oscillations with frequencies close to those prescribed by the local dispersion relation. With a weaker shear, spatially correlated large-amplitude waves with a well-defined frequency have been observed. Theoretically predicted eigenmodes are not the dominant feature of the fluctuation spectrum.

The problem of driving quasi-steady-state currents in tokamak plasmas with pulsed RF sources is examined in the regime , where τRF is the length of the RF pulse, is the electron collision time, TRF is the period between RF pulses, and TL/R is the L over R time of the tokamak plasma. This regime is appropriate to the proposed Microwave Tokamak Experiment (MTX) in which a pulsed free electron laser will be used to drive current in Alcator C. It is found that δI/Ī ≈ TRF/TL/R 1, where δI/Ī is the fractional variation in the plasma current. The production of runaway electrons can be a problem in the collisional limit, νe TRF 1. This will be avoided in the proposed MTX experiment if the electron temperature exceeds 1.5 keV.

A general theory for the stability of collision-less drift-wave eigenmodes in sheared slab magnetic fields is developed using the S-matrix technique. The eigenmodes are described with the integral formulation which fully takes into account the non-local effects of finite ion gyro-orbits. The universal eigenmode is then shown to be absolutely stable for arbitrary radial wavelengths.

The paper studies cold plasma trapping and heating of hot electrons in mirror geometry using a time-dependent, bounce-averaged Fokker-Planck code with quasi-linear diffusion due to RF heating at fundamental and second harmonic frequencies. With the restriction k|| = 0, the code models the beam-controlled heating (spatially restricted electric fields) that will be used to create thermal barriers in the TMX-Upgrade tandem-mirror experiment. By spatially localizing the microwave beams, which are strongly absorbed in a single pass, the mean hot-electron energy may be controlled. Heating is away from the midplane to control anisotropy (P⊥/P|| ). For a given magnetic field geometry and cold-plasma source temperature Ts, the parameters of the hot electrons scale with the quantity χ ≡ 2/nsω, where is the electric field, ns is the cold-plasma density, and ω is the frequency.

A useful model for calculating the variation in the electron density, ion density, and ambipolar potential along magnetic field lines in the thermal barrier cell of a tandem mirror is presented.

In a recent GA report Ohkawa describes a novel electron cyclotron current-drive (ECCD) scheme. Ohkawa finds a net toroidal plasma current driven by ''up-down'' asymmetries in the electron pressure anisotropy, i.e., differences in p/sub perpendicular/ - p/sub parallel/ evaluated at the same absolute value of B between the top and bottom half of the poloidal cross-section. Such up-down asymmetries are associated with the s-dependent, first order in nu/..omega../sub b/ part of the electron distribution function (s is distance along a magnetic field line). The current-drive efficiency estimated by Ohkawa is competitive with other rf current-drive schemes. Ohkawa's scheme is attractive because it appears that this scheme can produce an rf-driven current even when the electron-cyclotron power is dissipated on trapped electrons; and the scheme requires no selectivity in the sign of the parallel velocity of the electrons heated by the electron cyclotron wave. In sections I through III of this memo we analyze the current-drive calculation presented by Ohkawa. We conclude that, in the limit of long mean-free-path appropriate to current tokamak experiments and tokamak reactors, the current Ohkawa attempts to compute from a fluid model is a neoclassical correction to the Fisch-Boozer current. Neoclassical effects that give rise tomore » corrections of this sort were first pointed out by Ohkawa in an earlier paper on ECH current drive. These neoclassical effects (which are associated with magnetic field variations and trapped particles) have been treated more accurately by using a kinetic rather than fluid theory in several recent papers. They are important because they can reduce the ECCD efficiency below the level computed by Fisch and Boozer.« less