V.K. Paré

Heat pulse propagation studies in TFTR

The time-scales for sawtooth repetition and heat pulse propagation are much longer (tens of milliseconds) in the large tokamak TFTR than in previous, smaller tokamaks. This extended time-scale, coupled with more detailed diagnostics, has led us to revisit the analysis of heat pulse propagation as a method to determine the electron heat diffusivity χe in the plasma. A combination of analytic and computer solutions of the electron heat diffusion equation is used to clarify previous work and to develop new methods for determining χe. Direct comparison of the predicted heat pulses with soft-X-ray and ECE data indicates that the space-time evolution is diffusive. However, the χe determined from heat pulse propagation usually exceeds that determined from background plasma power balance considerations by a factor ranging from two to ten. Some hypotheses for resolving this discrepancy are discussed.

Impurity transport and plasma rotation in the ISX-B tokamak

Recent calculations have shown that when external momentum sources and plasma rotation are included in the neoclassical theory, the standard results for impurity transport can be strongly altered. Under appropriate conditions, inward convection is reduced by co-injection and enhanced by counter-injection. In order to examine the theoretical predictions, several observations of impurity transport have been made in the ISX-B tokamak during neutral-beam injection for comparison with the transport seen with Ohmic heating alone. Both intrinsic contaminants and deliberately introduced test impurities display a behaviour that is in qualitative agreement with the predicted beam-driven effects. These correlations are particularly noticeable when the comparisons are made for deuterium where the impurity transport in the Ohmically heated discharges exhibits neoclassical-like characteristics, i.e. accumulation and long confinement times. Similar but smaller effects are observed in beam-heated hydrogen discharges; neoclassical-like behaviour is not seen in Ohmically heated hydrogen sequences. Emphasis has been placed on measuring toroidal plasma rotation, and semi-quantitative comparisons with the theories of beam-induced impurity transport have been made. It is possible that radial electric fields other than those associated with momentum transfer and increased anomalous processes during injection could also play a role.

Confinement in beam-heated plasmas: the effects of low-Z impurities

Confinement studies on the Impurity Study Experiment (ISX-B) in beam-heated plasmas contaminated with small quantities of low-Z impurities are reported. Experimental results on the correlation of particle and energy confinement are presented. A linear relationship of energy confinement and plasma density is observed. As density is increased further, this effect saturates and energy confinement becomes independent of electron density. The experiments have been extended to higher beam power, resulting in an expansion of the ISX-B operating space. Impurities other than neon (carbon and silicon) have been tried and do not produce an enhancement in confinement. Edge cooling by the introduction of impurities has been demonstrated. The change in confinement has been shown to be correlated with changes in the normalized poloidal field fluctuation level (θ/Bθ) but not with the density fluctuation level (ne/ne). The experimental results are compared with models of drift-wave and resistive ballooning turbulence and an explanation is offered for the difference between the results with recycling and non-recycling impurities.

Confinement improvement in beam heated ISX-B discharges with low-z impurity injection

Abstract Results are reported on improved confinement in the Impurity Study Experiment (ISX-B) neutral beam heated plasmas when a small amount of neon is injected shortly after the start of beam heating. The scaling of energy confinement is modified by the introduction of a dependence on line-averaged density. Calculations show the improvement is primarily caused by a reduction in electron heat conduction.

Particle removal with pump limiters in ISX-B

Abstract The first pump limiter experiments were performed on ISX-B. Two pump limiter modules were installed in the top and bottom of one toroidal sector of the tokamak. The modules consist of inertia cooled, TiC-coated graphite heads and ZrAl getter pumps each with a pumping speed of 1000–2000 l/s. The objective of the initial experiments was the demonstration of plasma particle control with pump limiters. The first set of experiments were performed in ohmic discharges (OH) in which the effect of the pump limiters on the plasma density was clearly demonstrated. In discharges characterized by Ip = 110 kA, B T = 15 kG , n e = 1−5 × 10 13 cm −3 and t = 0.3 s, the pressure rise in the pump limiters was typically 2 mTorr with the pumps off and 0.7 mTorr after activating the pumps. When the pumps were activated, the line-average plasma density decreased by up to a factor 2 at identical gas flow rates. The second set of measurements were performed in neutral beam heated discharges (NBI) with injected powers between 0.6 MW and 1.0 MW. Due to a cooling problem on one of the ZrAl pumps, the NBI experiments were carried out with one limiter only. The maximum pressure observed in NBI-discharges was 5 mTorr without activating the pumps, i.e., approximately twice as high as in OH-discharges. The exhaust efficiency, which is defined as the removed particle flux divided by the total particle flux in the scrape-off layer, is estimated to be 5%.

Second stability in the ATF torsatron—Experiment and theory

Access to the magnetohydrodynamic (MHD) second stability regime has been achieved in the Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)]. Operation with a field error that reduced the plasma radius and edge rotational transform resulted in peaked pressure profiles and increased Shafranov shift that lowered the theoretical transition to ideal MHD second stability to β0≊1.3%; the experimental β values (β0≤3%) are well above this transition. The measured magnetic fluctuations decrease with increasing β, and the pressure profile broadens, consistent with the theoretical expectations for self‐stabilization of resistive interchange modes. Initial results from experiments with the field error removed show that the pressure profile is now broader. These later discharges are characterized by a transition to improved (×2–3) confinement and a marked change in the edge density fluctuation spectrum, but the causal relationship of these changes is not yet clear.

Measurements of periodic ripple transport in the ISX-B tokamak

The effect of periodic toroidal field (TF) ripple on ion confinement has been studied in the ISX-B tokamak by comparing neutral-beam-heated plasma performance with 9 and 18 TF coils. Three ripple physics issues were treated by these experiments: (1) enhanced ion thermal conductivity, (2) enhanced loss of energetic ions, and (3) ripple damping of beam-induced toroidal plasma rotation, which may affect the plasma losses. Under a wide variety of plasma conditions, ripple reduced the central-ion temperature by a factor of approximately two (600 eV → 300 eV). Ion temperature was found to be nearly independent of applied neutral-beam power in the large ripple configuration (9 TF coils). These results are shown to be in reasonable agreement with theoretical models of ripple transport. Charge-exchange measurements of the fast-neutral flux indicated no loss of fast passing ions due to ripple, but a large depletion of the fast ions trapped in local ripple wells, as expected theoretically. The central toroidal rotation velocity was reduced by a factor of six by ripple, yielding a momentum confinement time substantially less (factor of about seven) than that expected from standard theoretical expressions for ripple-enhanced ion viscosity.

Beta and confinement scaling studies with neutral-beam heating in the ISX-B tokamak

Experiments to investigate the scaling of volume-averaged beta and a global energy confinement time for neutral-beam-heated (Pb ≤ 2.5 MW) discharges in the ISX-B tokamak are described. The results are condensed into a set of empirical scaling formulas which can be used as a guide for other theoretical and experimental studies of confinement in high-beta, neutral-beam-heated plasmas. The dependence on toroidal field BT, plasma current Ip, and line-averaged electron density e was determined by varying each of these while keeping other external variables fixed. Magnetic diagnostics were used to obtain global properties, and Thomson-scattering-based profile analysis was carried out to permit more detailed investigation of selected cases. The poloidal beta, βp, is found to be independent of BT and e at fixed beam power Pb; confinement is found to deteriorate with increasing Pb but to improve with IP, consistent with previous results. The mechanisms which govern this confinement scaling have not been discerned, but it apparently does not depend on β, BT, or the (m = 1; n = 1) MHD activity, which typically dominates the MHD diagnostic signals. Losses are primarily through the electron channel, and the scaling of electron energy confinement time is similar to that of .

Measurement of χe in ISX-B beam-heated discharges by heat pulse propagation

New measurements of the electron thermal diffusivity coefficient ?e for beam-heated (0?1.6 MW) discharges in the Impurity Study Experiment (ISX-B) tokamak confirm the numerical values and trends, with beam power inferred earlier from power balance considerations, and thus support the conclusion that the confinement deterioration with increasing beam power is due to enhanced electron heat conduction losses.

Effects of neutral-beam co- and counter-injection on impurity radiation from ISX-B plasmas

In addition to heating, neutral beam injection modifies the impurity transport in ISX-B plasmas. Radiometric data show the effects of co- and counter-injection on impurity radiation profiles to be significantly different. Data from an array of twelve collimated detectors and an uncollimated monitor (all pyroelectric) show the radial distribution of volume impurity emission to be peaked at the centre of the discharge with counter-injection and at the plasma edge with co-injection, especially for Pinj > 1 MW. At high co-injection power, only 20% of the total input power is lost as impurity radiation. Also, co-injection-heated discharges reach an equilibrium shortly after beam turn-on, whereas counter-injection-heated discharges do not reach equilibrium and usually terminate disruptively. It is conjectured that the disruption is due to modification of the plasma current distribution resulting from the extreme cooling at the plasma core.