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The n=0 axisymmetric vertical stability and vertical position control have been examined for the Tokamak Physics Experiment. The passive stabilization is accomplished by using stabilizer plates close to the plasma. The present configuration is found to provide robust stability over a wide range of plasma parameters. The active feedback control of the plasma vertical position is done using coils located inside the vacuum vessel. These are required to control random disturbances leading to /spl les/1.0 cm RMS displacements from the midplane, and acceptable coil currents and voltages are found.

Operation of the Burning Plasma Experiment (BPX) with a deuterium-tritium plasma and graphite divertor plates will result in significant retention of tritium in layers of eroded and subsequently redeposited carbon. This trapped tritium is at risk of release during a vacuum accident. An administrative limit of 2 g of releasable tritium has been established for assessment of the environmental impact of such an event. Analysis of the erosion and redeposition process indicates that the 2-g limit will be reached in about 200 full-parameter discharges. Helium-oxygen glow discharge cleaning will be used to remove the redeposited carbon and the trapped tritium. It is calculated that about 90 h of cleaning are required to remove the trapped tritium from 200 plasmas. The cleaning will be done during scheduled maintenance periods every third week. Studies indicate that the tritium inventory can be effectively managed with helium-oxygen glow discharge cleaning. >

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.

A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

A steepest descent algorithm is used to obtain a least-squares approximation for threedimensional, toroidal stellarator flux surfaces represented by a discrete set of Poincare puncture points. A stream function λ(ρ,θ,) is introduced as a renormalization parameter to improve the mode convergence properties of the double Fourier series for the inverse co-ordinates (R, Z) representing toroidally nested magnetic surfaces: R(ρ, θ,) = Σ Rmn(ρ) cos(mθ – n) and Z(ρ, θ,) = Σ Zmn(ρ) sin(mθ − n), where ρ is a scaled radial flux co-ordinate and (R, , Z) are standard cylindrical co-ordinates. The variable is the geometric toroidal angle, and θ is a parametric co-ordinate representing a poloidal angle. The stream function λ(ρ, θ,ρ) = Σ λmn() sin(mθ–n) and the rotational transform profile ι(ρ) are determined by solving a system of simultaneous linear equations obtained from the MHD equilibrium condition ∇ × ⋅∇ρ = 0, together with the stellarator condition for zero net toroidal current on each flux surface, ∇ × ⋅∇ = 0. Numerically computed Fourier representations are presented for vacuum configurations of the Advanced Toroidal Facility (ATF), Uragan-3 and TJ-II stellarator devices.

High speed pellet fueling experiments have been performed on the ISX-B device in a new regime characterized by large global density rise in both ohmic and neutral beam heated discharges. Hydrogen pellets of 1 mm in diameter were injected in the plasma midplane at velocities exceeding 1 km/s. In low temperature ohmic discharges, pellets penetrate beyond the magnetic axis, and in such cases a sharp decrease in ablation is observed as the pellet passes the plasma center. Density increases of approx. 300% have been observed without degrading plasma stability or confinement. Energy confinement time increases in agreement with the empirical scaling tau/sub E/ approx. n/sub e/ and central ion temperature increases as a result of improved ion-electron coupling. Laser-Thomson scattering and radiometer measurements indicate that the pellet interaction with the plasma is adiabatic. Penetration to r/a approx. 0.15 is optimal, in which case large amplitude sawtooth oscillations are observed and the density remains elevated. Gross plasma stability is dependent roughly on the amount of pellet penetration and can be correlated with the expected temporal evolution of the current density profile.

In some neutral-beam-heated ISX-B plasmas, the Thomson scattering profiles show that of constant temperature and surfaces of constant density do not coincide. This discrepancy is analysed in terms of the modification of the plasma equilibrium arising from toroidal rotation or other enhancements of the toroidal component of the stored energy density.

The beam from an electron gun was used to trace flux surfaces in the Advanced Toroidal Facility (ATF) torsatron. The ATF magnetic field was run steady state at 0.1 T, and the electron beam was detected optically with an image‐intensified, solid‐state camera when it impinged on a phosphor‐coated screen. Closed flux surfaces and islands at several low‐order resonances were observed. The largest island, located at the ι= 1/2 surface, was from 5 to 6 cm in width, and its presence implied the existence of magnetic field errors. To determine if these error fields could be traced to small misalignments of the magnetic coils, a device capable of accurately measuring the radial and vertical magnetic field components of individual coil sets was placed in the center of ATF. This device allowed for a determination of the precise location of each of the coils that make up the ATF coil set. No significant coil misalignments were found. A further investigation of the coil configuration led to the identification of dipole fields in the helical field coil leads as the source of the field errors. The techniques developed in making these measurements are described in the text.

Target plasmas for neutral beam injection (NBI) in the Advanced Toroidal Facility (ATF) are formed using approximately 200 kW of 53 GHz electron cyclotron heating (ECH). During the early operational period of this device, field errors existed which produced large islands at the ι = 1/2 surface and several smaller, but non-negligible closely spaced islands at larger radii. The evolution of typical plasma parameters during NBI into plasmas produced in this perturbed magnetic configuration is discussed. The ion temperatures near the centre of the ECH target plasmas, as measured from Doppler broadening of impurity lines, are approximately 100 eV, and the central electron temperatures are close to 800 eV. NBI at powers up to 1.4 MW characteristically raises Tj(O) by factors of two to three for 50−80 ms, but thereafter the ion temperature decays. The electron temperature is not raised by NBI above its level in the ECH plasmas; it typically drops monotonically from the start of injection. Both the Ti and the Te profiles indicate the existence of a low temperature region from the ι = 1/2 surface (ρ = 0.6) to the plasma edge, apparently because of the island structure, and modelling suggests that the plasma collapse may be caused by the peculiar radiation profile which develops under these conditions. However, whether the characteristics of the temperature profiles and the collapse can truly be ascribed to the existence of the islands can be decided only after repair of the field errors.

Substantial advances have been made in the design of stellarator configurations to satisfy physics properties and fabrication feasibility requirements for experimental devices. However, reactors will require further advances in configuration design, in particular with regard to maintenance and operational characteristics, in order to have high availability. The diamagnetic properties of bulk high temperature superconductor (HTS) material can be used to provide simple mechanisms for magnetic field-shaping by arranging them appropriately in an ambient field produced by relatively simple coils.A stellarator configuration has been developed based on this concept. A small number of toroidal field coils carrying appropriate current would be sufficient to create a background toroidal field. Discrete HTS monoliths (“pucks” or “tiles”) are placed on a shaped structure that can be split in the poloidal direction at arbitrary locations. This allows modular stellarators to be designed with large openings that...

High-beta, low-aspect-ratio (compact) stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady-state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has a substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.

The KSTAR team is carrying out the design and research and development for a steady-state-capable advanced superconducting tokamak to establish the basis for an attractive fusion reactor as a future energy source. The physics requirements are driven by the plasma control and exhaust capabilities needed to extend the performance and pulse length of tokamak plasmas. The tokamak has major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, a strongly shaped plasma cross-section shaping (elongation 2.0 and triangularity 0.8), and a double-null poloidal divertor. The initial pulse length is 20 s, long enough to study physics on confinement timescales, but short enough to permit economical plasma-facing component technology. The pulse length can be increased to 300 s through upgrades. The machine will be operable in either hydrogen or deuterium, but neutron yields will be constrained to avoid the cost and inconvenience of remote maintenance and low-activation materials. The magnet system provides an inductively driven 20 s pulse with full current, beta, and shaping. With non-inductive current drive, steady-state plasmas can be sustained over a wide range of profile shapes and plasma pressures, Passive structures are provided to stabilize the vertical instability and high-beta modes and internal coils are provided for fast position control. The divertor structures are designed for particle removal, recycling control, impurity control, and flexibility for advanced divertor operation. The plasma heating system is designed for heating, current-drive, profile control, and flexibility. It will deliver power via neutral beams (8 MW), ion-cyclotron waves (6 MW), and lower-hybrid waves (1.5 MW), each of which can be expanded through upgrades. A comprehensive set of diagnostics is planned for plasma control, performance evaluation, and physics understanding.

Strategies for the improvement of quasiaxisymmetric stellarator configurations are explored. Calculations of equilibrium flux surfaces for candidate configurations are also presented. One optimization strategy is found to generate configurations with improved neoclassical confinement, simpler coils with lower current density, and improved flux surface quality relative to previous designs. The flux surface calculations find significant differences in the extent of islands and stochastic regions between candidate configurations. (These calculations do not incorporate the predicted beneficial effects of perturbed bootstrap currents.) A method is demonstrated for removing low order islands from candidate configurations by relatively small modifications of the configuration. One configuration is identified as having particularly desirable properties for a proposed experiment. (C) 2001 American Institute of Physics.

High-beta, low-aspect-ratio ("compact") stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has a substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described. (C) 2000 American Institute of Physics. [S1070-664X(00)94905-X].

A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the comme...