M. Ali Mahdavi

Divertor heat flux reduction by D2 injection in DIII-D

D{sub 2} gas injected into ELMing H-mode discharges in DIII-D reduced total integrated heat flux to the divertor by {approximately}2{times} and peak heat flux by {approximately}5{times}, with only modest degradation to plasma stored energy. Steady gas injection without particle pumping results in eventual degradation in stored energy. The initial reduction in peak heat flux at the divertor tiles may be primarily due to the increase in radiated power from the X-point/divertor region. The eventual formation of a high density region near the X-point appears to play a role in momentum (and energy) transfer from the flux surfaces near the outboard strike point to flux surfaces farther out into the scrapeoff. This may also contribute to further reduction in peak heat flux.

D III-D divertor target heat flux measurements during ohmic and neutral beam heating

Time resolved power deposition profiles on the D III-D divertor target plates have been measured for Ohmic and neutral beam injection heated plasmas using fast response infrared thermography (τ ≤ 150 μs). Giant Edge Localized Modes have been observed which punctuate quiescent periods of good H-mode confinement and deposit more than 5% of the stored energy of the core plasma on the divertor armour tiles on millisecond timescales. The heat pulse associated with these events arrives approximately 0.5 ms earlier on the outer leg of the divertor relative to the inner leg. The measured power deposition profiles are displaced relative to the separatrix intercepts on the target plates, and the peak heat fluxes are a function of core plasma density.

Measurement and modeling of the DIII-D divertor plasma

Abstract We report here on a study of the DIII-D divertor plasma which combines experimental measurements and 2D plasma modeling using the Braams 82 code. New data from a set of Langmuir probes recently installed in the divertor are presented. These data complement divertor heat flux measurements and help to better characterize divertor operation in neutral-beam heated discharges. In L-mode plasmas we find that the electron temperature rises with beam power (up to 45 eV with Pbeam = 5 MW) and is a strong function of n¯e. The peak divertor heat flux rises linearly with power and develops a strong inboard/outboard asymmetry which depends upon the toroidal field direction. Such behavior is not seen in the B2 simulation results. Following the H-mode transition the divertor heat flux, plasma density and temperature remain low even as the main plasma density rises.

Plasma boundary experiments on DIII-D tokamak

Abstract A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, Ip ≲ 3 MA, βT ≲ 10.7%, P(auxiliary)≲ 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions.

Recycling and neutral transport in the DIII-D tokamak

Abstract Several diagnostics have been used to characterize the edge plasma in the DIII-D divertor region and thereby to understand the particle recycling and neutral particle transport in an open divertor. An array of Langmuir probes mounted on the carbon divertor plates determines that generally the electron temperature is low ( T e ⋍ 10–20 eV ), and the electron density is high ( n e ⋍ 5 × 10 19 m −3 ). The edge plasma temperature and density are also measured by a moveable Langmuir probe mounted at the midplane of the plasma; the data from these shot-by-shot edge radial scans are then connected with the data for the core plasma obtained by Thomson scattering. The heat flux to the divertor plates is measured by an absolutely calibrated infrared camera; a plasma model is used to compare the heat flux with the measured T e and n e . The molecular neutral pressure at the edge of the plasma at several poloidal and toroidal locations is obtained from absolutely-calibrated pressure gauges; a gas puff enables in-situ gauge calibrations before each plasma shot. Typically, the divertor pressure is 10–50 times larger than that at the midplane. Time-resolved Hα brightness measurements are obtained from an absolutely calibrated television camera viewing the divertor region from above; both strike points are viewed simultaneously. The emission from the inner and outer strike points are usually equal after the H-mode transition, but are often unequal after the first ELM in H-mode and during some phases of L-mode discharges. A tangential camera measures the falloff in emission from the X-point to the plasma midplane. These measurements have been modeled with the DEGAS neutral transport code specifically modified for DIII-D. The wall geometry, wall composition, measured magnetic geometry, and measured plasma parameters are inputs to the model. The code calculates the gauge pressure as a function of position and the Hα television picture directly. During H-mode, we find a factor of two agreement in both the measured pressures and the absolute Hα brightnesses with one exception: the measured divertor pressure is higher than the code prediction. However, we find that a local gas source located below the X-point equal to only 10% of the divertor recycling source will bring the divertor pressure into agreement with the data. A difference in the electron temperature and density at the two strike points has been used to model the asymmetric Hα emissions. Comparisons have been made between the model and the data during H-mode periods of the discharge. The model has also been used to predict the influence of baffles to form a more closed divertor configuration.

Impurity transport during the H-mode in DIII-D

In H-mode plasmas in DIII-D, large modulations in spectroscopically measured impurity densities have been observed during shots with giant edge localized modes (ELMs). These spectral modulations have been analysed with the MIST impurity transport code. This analysis indicates that impurities are alternately flowing towards the plasma centre and then away from it. This alternating flow is correlated with ELM produced changes in the electron density. The electron density oscillations are extreme, causing the density profile to switch from hollow (just before an ELM) to centrally peaked (just after an ELM). Neoclassical convection, dependent on ion density gradients, causes impurities to concentrate most heavily where the electron density is largest and can explain the modulating impurity behaviour. Anomalous diffusion, D 1.0 ? 104 cm2/s, reduces the degree of impurity peaking. As the plasma current increases, the increase in hollowness of electron density profiles can account for the observed decrease in central impurity accumulation. Transport of cobalt, injected by laser ablation, has also been studied; cobalt transport variations are consistent with the ELM induced changes seen in intrinsic impurity transport. The transport results may be consistent with neoclassical impurity convective fluxes and suggest that impurity accumulation in tokamaks will occur unless the electron density profile is flat or particle confinement is low.

EFFECT OF DIVERTOR GEOMETRY ON IMPURITY EXHAUST

The effect of poloidal divertor geometry on the exhaust of impurities is studied experimentally. In the expanded-boundary geometry, strong accumulation of argon in the divertor region and minimal reflux into the main plasma are observed. A simple model provides an explanation for the observations and leads to the conclusion that the two key geometrical factors for effective impurity control are an expansion of the divertor channel and a narrow gap between the divertor channel and the wall.

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.

Angular dependence of power in the DIII-D divertor

Irregularities in the DIII-D divertor tile angles are used to test the dependence of deposited power on the angle of the magnetic field in the range 0–3°. Results are presented which show that the sin θ law, which is assumed in thermal design calculations for CIT and ITER divertor plates, is accurate down to 0.3°.

Increased divertor exhaust by electrical bias in DIII-D

The particle exhaust fate of the DIII-D tokamak divertor cryopump system was investigated as a function of gas-puff-regulated core plasma line-averaged density, divertor electrical bias potential and position of the divertor strike point with respect to the pump entrance aperture. The experiment was done with rapidly ELMing H mode plasmas. Divertor bias approximately doubled the maximum achievable exhaust rate and also reduced its sensitivity to the strike point position. The achievable exhaust rate varied roughly as the third power of the core plasma density