Verification of doppler coherence imaging for 2D ion velocity measurements on DIII-D

Coherence Imaging Spectroscopy (CIS) has emerged as a powerful tool for investigating complex ion phenomena in the boundary of magnetically confined plasma devices. The combination of Fourier-transform interferometry and high-resolution fast-framing cameras has made it possible to make sensitive velocity measurements that are also spatially resolved. However, this sensitivity makes the diagnostic vulnerable to environmental effects including thermal drifts, vibration, and magnetic fields that can influence the velocity measurement. Additionally, the ability to provide an absolute calibration for these geometries can be impacted by differences in the light-collection geometry between the plasma and reference light source, spectral impurities, and the presence of thin-films on in-vessel optics. This paper discusses the mitigation of these effects and demonstration that environmental effects result in less than 0.5 km/s error on the DIII-D CIS systems. A diagnostic comparison is used to demonstrate agreement between CIS and traditional spectroscopy once tomographic artifacts are accounted for.This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Award Nos. DEFC02-04ER54698 (DIII-D), DE-AC52-07NA27344 (LLNL), and DE-AC05-00OR22725 (ORNL). DIII-D data shown in this paper can be obtained in digital format by following the links at https://fusion.gat.com/global/D3D DMP

Measurements and modeling of type-I and type-II ELMs heat flux to the DIII-D divertor

Type-I and type-II edge-localized-modes (ELMs) heat flux profiles measured at the DIII-D divertor feature a peak in the vicinity of the strike-point and a plateau in the scrape-off-layer (SOL), which extends to the first wall. The plateau is present in attached and detached divertors and it is found to originate with plasma bursts upstream in the SOL. The integrated ELM heat flux is distributed at ∼65% in the peak and ∼35% in this plateau. The parallel loss model, currently used at ITER to predict power loads to the walls, is benchmarked using these results in the primary and secondary divertors with unprecedented constraints using experimental input data for ELM size, radial velocity, energy, electron temperature and density, heat flux footprints and number of filaments. The model can reproduce the experimental near-SOL peak within ∼20%, but cannot match the SOL plateau. Employing a two-component approach for the ELM radial velocity, as guided by intermittent data, the full radial heat flux profile can be well matched. The ELM-averaged radial velocity at the separatrix, which explains profile widening, increases from ∼0.2 km s−1 in attached to ∼0.8 km s−1 in detached scenarios, as the ELM filaments’ path becomes electrically disconnected from the sheath at the target. The results presented here indicate filaments fragmentation as a possible mechanism for ELM transport to the far-SOL and provide evidence on the beneficial role of detachment to mitigate ELM flux in the divertor far-SOL. However, these findings imply that wall regions far from the strike points in future machines should be designed to withstand significant heat flux, even for small-ELM regimes

Analysis of edge stability for models of heat flux width

Detailed measurements of the ne, Te, and Ti profiles in the vicinity of the separatrix of ELMing H-mode discharges have been used to examine plasma stability at the extreme edge of the plasma and assess stability dependent models of the heat flux width. The results are strongly contrary to the critical gradient model, which posits that a ballooning instability determines a gradient scale length related to the heat flux width. The results of this analysis are not sensitive to the choice of location to evaluate stability. Significantly, it is also found that the results are completely consistent with the heuristic drift model for the heat flux width. Here the edge pressure gradient scales with plasma density and is proportional to the pressure gradient inferred from the equilibrium in accordance with the predictions of that theory

DIAGNOSIS OF EDGE LOCALIZED MODE EVOLUTION IN DIII-D USING FAST-GATED CID AND INFRARED CAMERAS

The tangentially viewing visible and vertically viewing infrared cameras systems on DIII-D were upgraded to permit emission measurements during edge localized modes (ELMs) with integration times as short as 1 and 100 {micro}s respectively. The visible system was used to obtain 2-D poloidal profiles of CIII (465 nm) and D{sub {alpha}} (656.3 nm) emission with 20 {micro}s integration during various stages of ELM events in the lower DIII-D divertor. The infrared (IR) system was used to measure the heat flux to the divertor targets at 10 kHz with 100 {micro}s exposure. Upgrades to the data processing and storage systems permitted efficient comparison of the temporal evolution of these measurements

Fueling efficiency of pellet injection on DIII-D

Pellet injection has been used on the DIII-D tokamak to study density limits and particle transport in H-mode and inner wall limited L-mode plasmas. These experiments have provided a variety of conditions in which to examine the fueling efficiency of pellets injected into DIIID plasmas. The fueling efficiency defined as the total increase in number of plasma electrons divided by the number of pellet fuel atoms, is determined by measurements of density profiles before and just after pellet injection. We have found that there is a decrease in the pellet fueling efficiency with increased neutral beam injection power. The pellet penetration depth also decreases with increased neutral beam injection power so that, in general, fueling efficiency increases with penetration depth. The fueling efficiency is generally 25% lower in ELMing H-mode discharges than in L-mode due to an expulsion of particles with a pellet triggered ELM. A comparison with fueling efficiency data from other tokamaks shows similar behavior

Study of Z scaling of runaway electron plateau final loss energy deposition into wall of DIII-D

Controlled runaway electron (RE) plateau-wall strikes with different initial impurity levels are used to study the effect of background plasma ion charge Z (resistivity) on RE-wall loss dynamics. It is found that Joule heating (magnetic to kinetic energy conversion) during the final loss does not go up monotonically with increasing Z but peaks at intermediate Z similar to 6. Joule heating and overall time scales of the RE final loss are found to be reasonably well-described by a basic 0D coupled-circuit model, with only the loss time as a free parameter. This loss time is found to be fairly well correlated with the avalanche time, possibly suggesting that the RE final loss rate is limited by the avalanche rate. First attempts at measuring total energy deposition to the vessel walls by REs during the final loss are made. At higher plasma impurity levels Z > 5, energy deposition to the wall appears to be consistent with modeling, at least within the large uncertainties of the measurement. At low impurity levels Z < 5, however, local energy deposition appears around 5-20x less than expected, suggesting that the RE energy dissipation at low Z is not fully understood. Published by AIP Publishing.This work was supported in part by the U.S. Department of Energy under Nos. DE-FG02-07ER54917, DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC52-07NA27344, and DE-AC05-06OR23100 and in part by the Spanish Direccion General de Investigacion Cientifica y Tecnica under Projects ENE2012–31753 and ENE2015-66444R (MINECO/FEDERE, UE). DIII-D data shown in this paper can be obtained in digital format by following the links at https://fusion.gat.com/global/D3D_DMPPublicad

Edge Recycling and Heat Fluxes in L- and H-mode NSTX Plasmas

Introduction Edge characterization experiments have been conducted in NSTX to provide an initial survey of the edge particle and heat fluxes and their scaling with input power and electron density. The experiments also provided a database of conditions for the analyses of the NSTX global particle sources, core fueling, and divertor operating regimes

Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D

Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs) during ELMy H-mode discharges. Samples with pre-adhered, pre-characterized dust have been exposed at the outer strike point (OSP) in a series of discharges with varied intra-(inter-) ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem “adhesive tape” sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER. Keywords: Dust, Fusion, Tokamak, Remobilization, Adhesio