Toroidal Fusion Devices Research Articles

Resonant plasma heating below the cyclotron frequency

Resonant heating of a magnetized plasma by low frequency waves of large amplitude is considered. It is shown that the magnetic moment can be changed nonadiabatically by a single large amplitude wave, even at frequencies normally considered nonresonant. Two examples clearly demonstrate the existence of the resonances leading to chaos and the generic nature of heating below the cyclotron frequency. First the classical case of an electrostatic wave of large amplitude propagating across a confining uniform magnetic field, and second a large amplitude Alfvén wave, propagating obliquely across the magnetic field. Waves with frequencies a small fraction of the cyclotron frequency are shown to produce significant heating; bringing, in the case of Alfvén waves, particles to speeds comparable to the Alfvén velocity in a few hundred cyclotron periods. Stochastic threshold for heating occurs at significantly lower amplitude with a perturbation spectrum consisting of a number of modes. This phenomenon may have relevance for the heating of ions in the solar corona as well as for ion heating in some toroidal confinement fusion devices.

Control of tearing modes in toroidal fusion experiments using “designer” error fields

It is demonstrated that a magnetic island chain formed by a saturated tearing instability in a toroidal magnetic fusion device can lock to a special class of externally generated magnetic perturbation in a stabilizing phase. The theoretical apparatus needed to design such perturbations is outlined. These special perturbations—which are termed “designer” error fields—could be used to control the amplitudes of tearing modes in toroidal magnetic fusion experiments without the requirement of fast phase modulation. This type of control would be far more feasible in a reactor environment than conventional active feedback control via external magnetic perturbations.

Overview of ECRH experimental results

A review of the present status of electron cyclotron heating and current drive experiments in toroidal fusion devices is presented. In addition to basic heating and current drive studies the review also addresses advances in wave physics and the application of electron cyclotron waves for instability control, transport studies, pre-ionization/start-up assist, etc. A comprehensive overview is given with particular emphasis on recent advances since the major review of Erckmann and Gasparino (1994) ( 36 1869), including results from the latest generation of high-power, high-frequency experiments.

Neutron diagnostics in compact helical system

An efficient neutron diagnostic system has been established and applied to deuterium experiments on compact helical system. A neutron counter was installed at the center of the torus, and the detection efficiency was about two orders of magnitude higher than the values of most toroidal fusion devices. The effect of the stray magnetic field on the operation of a proportional counter was experimentally confirmed to be negligibly small. The ion temperature, as low as 400 eV, was obtained from neutron measurement with a time resolution of 1 ms. The decay time of neutron emission rate after the termination of a 1% deuterium-doped beam was measured and the confinement time of neutral beam-injected fast ions was estimated. The fast ion confinement was clearly improved by strengthening the toroidal magnetic field.

Control of magnetic islands with horizontal field trim coils in a low aspect ratio torsatron

The presence of large magnetic islands associated with rational magnetic surfaces in toroidal fusion devices can lead to a reduction of the confinement time of plasmas in such devices. In stellarators and torsatrons, the island size is usually minimized by careful design and coil construction within close tolerances in order to reduce the effect of error fields. In addition, magnetic islands resulting from unavoidable error fields can be reduced or eliminated through the use of auxiliary trim coils. The Compact Auburn Torsatron has been equipped with two pairs of large Helmholtz coils producting mutually orthogonal magnetic fields in the horizontal plane. These trim coils are used to control the size and phase of the t=1/2 magnetic island. A reduction of the inherent t=1/2 magnetic island size by a factor of three is achieved through a systematic variation of both horizontal trim field components. The minimization of the island size occurs simultaneously with the largest poloidal rotation rate of the island as a function of the correction coil current

Theory of free-electron-laser heating and current drive in magnetized plasmas

The introduction of a powerful new microwave source, the free-electron laser, provides new opportunities for novel heating and current-drive schemes to be used in toroidal fusion devices. This high-power, pulsed source has a number of technical advantages for these applications, and its use is predicted to lead to improved current-drive efficiencies and opacities in reactor-grade fusion plasmas in specific cases. The Microwave Tokamak Experiment at the Lawrence Livermore National Laboratory will provide a test for some of these new heating and current-drive schemes. Although the motivation for much of this research has derived from the application of a free-electron laser to the heating of a tokamak plasma at a frequency near the electron cyclotron frequency, the underlying physics, i.e., the highly nonlinear interaction of an intense, pulsed, coherent electromagnetic wave with an electron in a magnetized plasma including relativistic effects, is of general interest. Other relevant applications include ionospheric modification by radio-frequency waves, high-energy electron accelerators, and the propagation of intense, pulsed electromagnetic waves in space and astrophysical plasmas. This review reports recent theoretical progress in the analysis and computer simulation of the absorption and current drive produced by intense pulses, and of the possible complications that may arise, e.g., parametric instabilities, nonlinear self-focusing, trapped-particle sideband instability, and instabilities of the heated plasma.

RF Plasma heating in Toroidal Fusion Devices, by V. E. Golant and V. I. Fedorov. Consultants Bureau, 1989, $85, 194 pages.

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Status of local transport measurements and analysis in toroidal devices

This paper reviews the status of methods used to analyze local transport of particles, energy, and angular momentum in plasmas contained in toroidal fusion devices. The standard technique at present is based on determination of particle, energy, and momentum fluxes through analysis based on measured profiles of density, temperature and angular rotation speed and on calculation of all sources of particles, energy, and momentum. Recent experiments have also been done using perturbations of local density, temperature, or angular rotation speed caused by modulating the input sources. The local transport information is then extracted from the time history of the perturbations. These two techniques are contrasted and the strengths of each are discussed in this paper. Recommendations for further work needed to make progress in understanding transport are also given.

Toroidal ion-pressure-gradient-driven drift instabilities and transport revisited

A unified theory of ion-pressure-gradient-driven drift wave instabilities and transport is presented, which ties the long-wavelength trapped-ion mode to the moderate-wavelength hydrodynamic mode in toroidal geometry. An analytic dispersion relation that retains ion drift resonances, and keeps the leading-order contribution from finite Larmor radius effects and parallel compressibility, is derived. Results indicate that the slab and toroidal branches of these instabilities are of comparable importance, and are both strong candidates to explain the observed anomalous ion loss in toroidal fusion devices. However, it is concluded that in the limit of flat-density profiles characteristic of H-mode discharges, the stabilizing influence of perpendicular compressibility is insufficient to corroborate an improvement, if any, in ion confinement quality. Mixing-length expressions for the fluctuation amplitudes and both electron and ion transport coefficients are derived. Results also indicate that the heretofore experimentally observed favorable current scaling of the energy confinement time may saturate in low ion-collisionality discharges. Finally, it is shown that a population of energetic trapped particles, such as those that may be produced during radio frequency or perpendicular neutral beam heating, can significantly exacerbate the instability. Several suggestions for experiments are made to help in differentiating among various anomalous transport scenarios.

Tritium Assay of Li2O in the LBM/LOTUS Experiments

The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM. PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results form the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system.

Particle and impurity control in toroidal fusion devices

A review of working particle and impurity control techniques used in and proposed for magnetic fusion devices is presented. The requirements of both present-day machines and envisaged fusion reactors are considered. The various techniques which have been proposed are characterized by whether they affect sources, sinks, or fluxes; in many cases a particular method or device can appear in more than one category. Examples are drawn from published results. The solutions proposed for the large devices which will be operating during the next 5 years are discussed.

Garching shows stellarators may be good after all

Stellarators appear to be back in business. After a series of consistently disappointing results in the 1960's, Lyman Spitzer's scheme for plasma confinement in a toroidal fusion device was generally abandoned in the last decade by American magnetic-confinement researchers, in favor of the more promising tokamak concept.

Toroidal fusion devices ignited by ohmic heating.

A toroidal configuration where the fusion ignition may be reached by ohmic heating alone is proposed. It is a hybrid between a tokamak and a diffused pinch. The reversal of a toroidal field in a diffused pinch is replaced by a doublet-like separatrix. Experimental results on mhd stability, especially limiting q-value of the doublet configuration will determine the validity of the scheme.

Program for development of toroidal superconducting magnets for fusion research

Toroidal fusion devices will require superconducting magnet systems of greater size and complexity than heretofore constructed. A program employing the talents of a number of organizations for the development of superconducting magnets for toroidal fusion devices has been inaugurated. The objective of this program is to demonstrate the suitability and reliability of toroidal superconducting magnets for fusion research devices and experimental power reactors (EPR) in a time compatible with the AEC Controlled Thermonuclear Research (CTR) program goal for EPR construction during early 1980's. Since much of the superconductivity expertise resides now in AEC laboratories, the focus of the program is in these laboratories, but industrial and consultative participation is an integral part of the plan.

Difficulties with injection of relativistic electron beams into toroidal fusion devices

It is shown that neither E × B drift nor diamagnetic beam configurations allow injection of externally generated strong relativistic electron beams into fusion devices for the purpose of plasma heating.