Increasing Plasma Current Research Articles

Confinement and power balance in the S-1 spheromak

The confinement and scaling features of the S-1 spheromak have been investigated using magnetic, spectroscopic and Thomson scattering data in conjunction with numerical modelling. Results from the multi-point Thomson scattering diagnostic show that the central beta remains constant (βt0 ∼ 5%) as the plasma current density increases from 0.68 to 2.1 MA·m−2. The density is observed to increase slowly over this range, while the central electron temperature increases much more rapidly. Analysis of the global plasma parameters shows a decrease in the volume average beta and the energy confinement as the total current is increased. The power balance has been modelled numerically with a zero-dimensional non-equilibrium time dependent coronal model and is consistent with the experimental observations. In addition to energy losses from radiation, the model includes convection and conduction. The conduction includes magnetic field fluctuation induced energy loss.

Impurity profiles for H-mode discharges in DIII-D

Impurity concentration profiles have been determined for H-mode discharges in the DIII-D tokamak from measured ne, Te, Zeff and radiated emissivity profiles. The central impurity levels in DIII-D high current H-modes, as modelled using this technique, remain below those seen in L-modes (fractional nickel concentrations ≤0.02%) throughout the neutral beam heating pulse. In contrast to some other experiments (ASDEX [1], JET [2], JFT-2M [3]), the H-mode does not terminate because of excessive radiation in DIII-D discharges heated with co-injected neutral beams. For increasing plasma current, the global impurity concentrations decrease and the profiles become more dominated by edge radiation. H-modes as obtained with electron cyclotron heating and co-injected neutral beams at similar heating powers also have low impurity levels, but the impurity distribution is significantly more hollow in the case of neutral beam heating.

Results of counterneutral beam injection on the DIII-D tokamak

Limiter, single-null, and double-null divertor plasmas have been heated by hydrogen neutral beam injection in the direction opposite to that of the plasma current. The H-mode power threshold is reduced by approximately 25% for counterinjection. Sawtooth suppression was possible only for high neutral beam power and for high values of the safety factor. No confinement improvement over similar coinjection discharges was observed in H-mode, L-mode, or limiter discharges. In plasmas without sawteeth, metallic impurity buildup in the central region of the plasma eventually leads to a reduction in confinement. The parametric dependence of energy confinement for both L-and H-mode plasmas was found to behave in the same manner during similar coinjection experiments. H-mode confinement increased linearly with increasing plasma current (65 msec/MA at 9 MW). Both L-mode and H-mode energy confinement decreased with increasing neutral beam power. At 1.25 MA, the L-mode plasmas had a confinement time that was less than that predicted by Kaye–Goldston scaling.

Confinement physics of H-mode discharges in DIII-D

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

Recent results of ultra low q experiments in TORIUT-6 AND REPUTE-1

In REPUTE-1, the radiation barrier has been overcome and ion and electron temperatures of ⪅ 0.6 keV (Ti) and ⪅0.3 keV (Te) have been achieved at an electron density of (0.2–0.7) × 1020 m−3 and toroidal beta values of 12-18%. A resistivity anomaly due to m = l kink activity has been observed which may be connected with the peaked temperature profile and the rather broad current density profile. The plasma resistance is lower than that of reversed field pinch discharges in REPUTE-1 and decreases with increasing plasma current. In TORIUT-6, the effects of current ramp-up and carbonization of the first wall have been studied. A hollow current density profile, typical of ultra low q equilibria, is preserved which may be due to the skin current effect. The difference between current ramp-up discharges and standard discharges is the extended duration of the quiescent phase and the preservation of the position of the pitch minimum away from the axis in the ramp-up discharges. After carbonization, the duration of the quiescent phase increases and the plasma resistivity is reduced by 50%.

DIII-D Latest Results and Implications

The DIII-D tokamak has demonstrated the feasibility of the divertor configuration for future reactor relevant devices. H-mode (good) confinement has been demonstrated for deuterium discharges in both single-null and double-null divertor configurations and under a variety of heating scenarios including neutral beam injection, electron cyclotron heating, and ohmic heating. Plasma beta values (β = 6.8%) near expected theoretical limits (3.5 I/aB) have been demonstrated. Plasma currents are limited by safety factors q > 2 near the surface of the plasma. Energy confinement improves with increasing plasma current until q decreases below 3. Divertor heat flow has been documented and initial boundary physics studies carried out. Non-inductive current drive has been demonstrated using neutral beam injection.

Photoemission optogalvanic spectroscopy: An i n s i t u method for plasma electrode surface characterization

Photoemission optogalvanic spectroscopy (POGS) is shown to be useful for plasma electrode surface characterization. A pulsed ultraviolet laser is used to induce photoemission from the electrode surface in a radio frequency plasma reactor and the increase in plasma current is detected. The photoemission process is first characterized in vacuum and then compared to that in several plasma gases using Al and Si electrodes. In vacuum, the laser-induced photoemission signal is generally consistent with space-charge-limited current. When below the space-charge limit, the magnitude of the photoemission signal depends upon laser wavelength and power, surface composition, and film thickness. The removal of SiO2 from Si and the contamination of Al in fluorine-containing plasmas is monitored using this technique. A large increase in the POGS signal is observed as the oxide is removed from Si or when a fluorinated Al surface is exposed to an O2 plasma. The POGS signal decreases with fluorine exposure with both Al and Si. We are continuing to explore the utility of this technique as an endpoint detector and in situ contamination monitor.

Ultra low-q discharge and high temperature experiments in REPUTE-1.

The outline and necessity of research in the ultra low-q regime are discussed from the view point of the total understanding of the current carrying systems. The ultra low-q state lies between tokamak and reversed field pinch and has a safety factor profile of 1/ (n+1) <q (r) <1/n, n=1, 2, 3, 4…. In REPUTE-1, the radiation barrier has been overcome and achieved plasma temperatures, Ti and Te, exceed 0.5 keV with electron density of 0.7×1020m-3. Resistivity anomaly due to m=1 kink activity is observed and considered to be related to a peaked temperature profile with a rather broad current density profile. The plasma resistivity in the ultra low-q regime is proved to be much lower than predicted values and decreases with increasing plasma current.

Initial results of RFP experiments in REPUTE-1

Experiments of a reversed field pinch device REPUTE-1 constructed in University of Tokyo started in 1984. The conductivity electron temperature and the line-averaged electron density increase approximately linearly with the increase of plasma current in the range up to 240 kA. The study of the effect of the bias toroidal field on the discharge characteristics shows that a higher bias field gives lower resistivity and a higher flat top plasma current. The toroidal flux at the current flat top tends to be a selective value independent of the initial bias field.

Dependence of Plasma Properties on Elongation Ratio in Non-Circular Tokamak TNT-A

The relation between the plasma properties and elongation ratio is studied in a non-circular tokamak in the case of the safety factor q a ∼2. With the increase of elongation ratio κ (1.3→1.5), the plasma current increases from 19.5 to 25 kA, the loop voltage decreases from 4 to 3 V, the mean current density and the electron temperature increase slightly (∼10%), while the relative amplitude and the poloidal dependence of Mirnov oscillations ( m =2 mode) change little.

Nonlinear kink instabilities in tokamaks surrounded by force-free fields

The theory of Pao for the nonlinear behavior of linearly unstable kink modes in a sharp boundary plasma surrounded by a vacuum is extended to include force-free and distributed plasma currents. Nonlinear stabilization of the external kink (m=1) instability by force-free fields is enhanced by increasing the force-free current up to an optimum value. For fixed force-free currents, nonlinear kink stability is decreased by increasing the distributed plasma current. In general, the optimum force-free current increases as the plasma current increases. A condition to determine the nonlinear kink stability is derived by a perturbational method for long wavelength perturbations.

Equilibria of high pressure elliptic flux-conserving tokamak

An analytical calculation is carried out to determine the plasma current and poloidal beta for a toroidal plasma with an elliptic cross section under the constraint of flux conservation. It is shown that the pressure buildup during heating occurs in two stages: An outward shift in the magnetic axis and an elongation in the flux surfaces. The total plasma current increases with the pressure variable more rapidly for an elliptic cross section than it does for a circular cross section. This current rise produces a rather slow rise of the poloidal beta with pressure, particularly for large elongations. The ensuing class of flux conserving equilibria is characterized by large beta but modest poloidal beta.

High-Pressure Flux-Conserving Tokamak Equilibria

An analytic theory is developed to calculate poloidal beta ..beta../sub I/ and the diamagnetic parameter ..mu../sub I/ for axisymmetric toroidal magnetohydrodynamic equilibria confining high-pressure plasmas (..beta.. approx. O (a/R)) under the constraint of flux conservation. To satisfy the equilibrium equations, the plasma current increases with pressure as p/sup 1///sup 3/. Previously calculated equilibrium limits on poloidal ..beta.. are avoided. (AIP)

Measurement and modification of first-wall surface composition in the Oak ridge tokamak (ormak.)

Impurities coming into the plasma from the walls of present‐day toroidal plasma confinement devices modify plasma behavior substantially. Small fractions of high‐Z ions in the plasma greatly decrease plasma temperatures and increase plasma energy losses. We are attempting to identify and control the sources of impurities from the ’’first‐wall’’ in ORMAK. Auger electron spectroscopy, soft x‐ray appearance‐potential spectroscopy, and other surface‐sensitive techniques have been used to characterize the surface composition of the first wall and to develop methods to remove carbon and oxygen. Oxygen glow‐discharge cleaning has been shown, in the laboratory, to be an effective way of removing carbon from gold films (simulated ORMAK liner material) and the use of oxygen‐discharge cleaning in ORMAK has resulted in a decrease in plasma contamination, a 50% increase in plasma current, and an accompanying increase in plasma temperature. In spite of these improvements the walls of ORMAK are far from clean. Substanti...

Measurement and modification of first-wall surface composition in the Oak Ridge tokamak (ORMAK)

Impurities coming into the plasma from the walls of present-day toroidal plasma confinement devices modify plasma behavior substantially. Small fractions of high-Z ions in the plasma greatly decrease plasma temperatures and increase plasma energy losses. We are attempting to identify and control the sources of impurities from the ’’first-wall’’ in ORMAK. Auger electron spectroscopy, soft x-ray appearance-potential spectroscopy, and other surface-sensitive techniques have been used to characterize the surface composition of the first wall and to develop methods to remove carbon and oxygen. Oxygen glow-discharge cleaning has been shown, in the laboratory, to be an effective way of removing carbon from gold films (simulated ORMAK liner material) and the use of oxygen-discharge cleaning in ORMAK has resulted in a decrease in plasma contamination, a 50% increase in plasma current, and an accompanying increase in plasma temperature. In spite of these improvements the walls of ORMAK are far from clean. Substantial amounts of carbon, oxygen, iron, and other elements remain.