M. E. Rensink

Simulation of experimentally achieved DIII‐D detached plasmas using the UEDGE code

The introduction of a divertor Thomson scattering system in DIII‐D [J. Luxon et al., International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1986), p. 159] has enabled accurate determination of the plasma properties in the divertor region. Two plasma regimes are identified: detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5–10 eV determined earlier. Fluid models of the DIII‐D scrape‐off layer plasma successfully reproduce many of the features of these two regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low‐temperature region associated with the detached plasma mode. Low‐temperature atomic physics processes that are not included in the present models may account for this discrepancy.

Stability of the detachment front in a tokamak divertor

Abstract Experimental observations show that when a divertor plasma detaches, radiation regions (fronts) move towards the X-point. This indicates that the localization of the fronts in the divertor is unstable. We study theoretically the stability of detached divertor operation and compare our results with experimental data. We show that the impurity radiation and plasma recombination effects play a crucial role in the evolution of detached plasma parameters. In the tokamak SOL these effects result in the saturation of upstream plasma density and temperature as the total number of particles (ions and neutrals) in the SOL is increased, and cause both impurity radiation and ionization–recombination fronts propagation towards the X-point. These conclusions are supported by experimental data from Alcator C-Mod.

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 in edge plasmas with application to liquid walls

The transport of impurity ions in a magnetically confined plasma is studied in the region between their origin at a material surface and the core plasma as defined by closed magnetic flux surfaces. The focus is physics understanding of the results of two-dimensional (2-D) transport modeling of the plasma and neutrals. A simple one-dimensional model is introduced to identify key processes and illustrate how such processes affect the core-edge impurity level. The 2-D simulation gives detailed results of scaling of the impurity level with parameters such as anomalous radial diffusion, hydrogen–plasma recycling, core power flux, and core-edge density. The results are obtained for a slab model of a tokamak, but by changing the magnetic connection length, scaling to other types of devices can be inferred. Lithium and fluorine impurities are considered explicitly as examples with low and moderate charge-state number, Z, for liquid wall materials; trends found for these cases provide guidance to the behavior of o...

Divertor design for the tokamak physics experiment

In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

Divertor target profiles and recycling studies in TCV single null lower standard discharges

A standard, single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of T-e and n(e) before and after He glow.

Attainment of a stable, fully detached plasma state in innovative divertor configurations

A computational study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE. Several divertor configurations are considered, with radially or vertically extended, tightly baffled, outer divertor legs and with or without a secondary X-point in the divertor leg volume. For otherwise identical conditions, a scan of the input power from the core plasma is performed. As the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state, which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remains stable. At low power, the detachment front eventually moves all the way to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, for long-legged divertors, the detached operation window is quit...

Particle control in the DIII-D advanced divertor

A novel, electrically biasable, semiclosed divertor was installed and operated in the DIII-D lower outside divertor location. The semiclosed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. Electrical bias controls the distribution of particle recycle between the inner and outer divertors by E*B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduces the sensitivity of plenum pressure to separatrix position. In particular, E*B drifts in the DIII-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density.<<ETX>>