Search

A summary of experimental results pertaining to plasma energy and particle transport in the Impurity Study Experiment (ISX-B) tokamak is presented. Emphasis is placed on results with neutral beam heating, usually in the co- direction (aligned with the plasma current), and the relative roles of various energy loss mechanisms are discussed. The derived electron thermal diffusivity and the predictions of various models have been compared. The measured values are within a factor of 2 of the values expected to result from resistive pressure-driven modes. Evidence for the presence of these modes is discussed. Values of ion thermal diffusivity are compared with the predictions of neoclassical theory. Anomaly factors (the ratio of experimental value to theoretical value) between 1 and 5 are found. Comparisons between experimental results and theoretical predictions are also made for cases with increased toroidal field ripple, produced by using only 9 of the 18 toroidal field coils. Convection is shown to play a small role in energy transport, except at the plasma periphery. The behavior of both plasma and impurity particles is discussed and shown to be strongly dependent on the direction of the plasma current relative to the neutral beam (coinjection and counterinjection). Beta limits aremore » discussed. The maximum values obtained for both ..beta../sub I/ (approx. 2) and (approx. 2.5%) are not thought to be limited except by the restricted power available for heating.« less

Pulsed torsatrons and heliotrons are susceptible to runaway electron formation and confinement resulting from the inherent good containment in the vacuum fields and the high loop voltages during the initiation and termination of the helical and vertical fields (''field ramping''). Because runaway electrons can cause an unacceptable level of hard X rays near the machine, a runaway suppression system was designed and included in the initial operation of the Advanced Toroidal Facility (ATF). The main component of the system is a rotating paddle that is normally left in the vacuum chamber during the field ramps. This device proved to be very effective in reducing the runaway population. Measurements of hard X rays from ATF have shown that the runaways are produced primarily during the field ramping but that usually a small steady-state runaway component is also present during the ''flat-top'' portion of the fields. The paddle is the main source of the hard X rays (thick-target bremsstrahlung), although other objects in the vacuum chamber also serve as targets for the runaways at various times. The maximum X-ray energy found by pulse height analysis is /approximately/12--15 MeV; the mean energy appears to be a few mega-electron-volts. A noticeable forward peaking ofmore » the bremsstrahlung from the paddle is evident. The limiters do not appear to be major sources of bremsstrahlung. 17 refs., 14 figs.« less

A phased program of plasma heating at the electron cyclotron frequency is proposed for the Oak Ridge tokamaks ISX and ORMAK Upgrade. The past history of the program of electron cyclotron heating (ECH) at ORNL on mirrors and in the ELMO Bumpy Torus has been successful. Future technological developments in the production of high power high frequency microwave tubes look promising at this time. The physics of wave propagation and particle heating are fairly well understood and indicate the viability of this technique. Studies on breakdown and on runaway electron reduction will provide useful information for larger machines. Recent experiments in the USSR on small tokamaks have shown that ECH is a viable heating technique. Providing that the microwave tubes become available, the engineering considerations suggest that the technique is practical and workable, based on present day technology.

Electron cyclotron heating (ECH) experiments in the Advanced Toroidal Facility (ATF) torsatron exploit unique capabilities for external control of the magnetic configuration and long‐pulse operation. The ECH power deposition profile is determined from the change of the electron temperature profiles during ECH power turn‐off and from power modulation. The measured dependence of absorbed power on the magnetic field agrees well with that obtained from ray tracing calculations for first‐pass and multiple‐bounce absorption. In ATF, parameters of basic physics interest (magnetic shear, magnetic well/hill, and contained trapped particle fraction) have been varied dynamically during a single long‐pulse (up to 20‐s) discharge (dynamic configuration control). Varying a single physics parameter while keeping others fixed elucidates the parameter’s influence on plasma behavior. Modulation of the magnetic well significantly affects the plasma stored energy. This may be related to the observed electron density fluctuations, which are consistent with theoretical predictions of pressure‐gradient‐driven resistive interchange turbulence. Measurements of the electric potential with a heavy ion beam probe show positive potential in the core of typical ECH discharges and the presence of a velocity shear layer near the edge.

The new edge control scenario for the large helical device (LHD) with a local island divertor (LID) has been proposed, and technical and design studies of the LID have been done in detail. Although the LHD edge plasma control will primarily be done with a closed full helical divertor that utilizes a natural separatrix in the edge region, the LID that utilizes an m / n =1/1 island will be used in the early stage of the LHD experiment, prior to installing the closed full helical divertor. The advantage of the LID over the closed full helical divertor is the technical ease of hydrogen pumping because of the toroidally localized hydrogen recycling. This localized recycling, however, leads to high heat and particle fluxes onto divertor plates of an LID divertor head, and an average heat flux of 5 MW m − 2 has been designed for 3 MW steady state discharges. This paper will summarize the new edge control scenario proved experimentally, as well as the technical and design studies of the LID for LHD.

A high power ion Bernstein wave heating system has been introduced on PBX‐M for heating and for controlling the plasma pressure profile in an effort to achieve the stable high beta ‘‘second stability’’ regime. The pressure profile can be controlled through local bulk ion heating as well as density profile control. In bean‐shaped plasmas with plasma currents range from 180 kA to 250 kA, good ion heating up to the highest applied rf power, (≊700 kW,) has been observed. The observed broadening of the ion temperature profile is consistent with localized off‐axis bulk ion heating as predicted by IBW ray tracing calculations. Application of IBW also resulted in a greatly modified density profile. The ability for IBW to change the density profile appears to be particularly attractive for controlling the bootstrap current profile for advanced tokamaks. Many important IBWH‐related edge physics results were also obtained, including ponderomotive edge plasma modification and parametric instability onset conditions. ...

Recent experiments in the Advanced Toroidal Facility (ATF) torsatron [Plasma Physics and Controlled Nuclear Fusion Research 1990 (IAEA, Vienna, in press)] have emphasized the role of magnetic configuration control in transport studies. Long‐pulse plasma operation up to 20 sec has been achieved with electron cyclotron heating (ECH). With neutral beam injection (NBI) power of ≥1 MW, global energy confinement times of 30 msec have been obtained with line‐average densities up to 1.3×1020 m−3. The energy confinement and the operational space in ATF are roughly the same as those in tokamaks of similar size and field. The empirical scaling observed is similar to gyro‐reduced Bohm scaling with favorable dependences on density and field offsetting an unfavorable power dependence. The toroidal current measured during ECH is identified as the bootstrap current. The observed currents agree well with predictions of neoclassical theory in magnitude and in parametric dependence. Variations of the magnetic configuration ...

The Advanced Toroidal Facility (ATF) [Fusion Technol. 10, 179 (1986)] is the world’s largest stellarator. It was designed and built to demonstrate high beta, steady‐state operation in a toroidal confinement system. During its final operating period ATF achieved pulse lengths of over one hour (4667 s). The objectives of these experiments were (1) investigation of plasma performance at times that are long compared to the plasma/wall equilibrium time; (2) determination of plasma control and wall conditioning techniques; and (3) adaptation of plasma diagnostic and data acquisition systems to long‐pulse operation. Other experiments have also extended earlier studies of dimensionless‐parameter plasma confinement scaling. By employing two discrete electron cyclotron heating (ECH) frequencies (28 and 35 GHz), and by simultaneously modulating the ECH power, magnetic field, and plasma density, it has been possible to maintain fixed plasma beta and collisionality while modulating the normalized gyroradius.