New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Abstract Carbon-based codeposits formed in carbon-containing fusion devices have the potential to dominate tritium retention in the torus. One of the tritium removal techniques currently being studied is thermo-oxidation, which is unique in its ability to remove tritium from codeposits without mechanical intervention in the torus and in its ability to remove tritium from codeposits in tile gaps and shaded areas. In preparation for an oxidation experiment planned to be performed in DIII-D, we have investigated the potential collateral effects of thermo-oxidation on DIII-D in-vessel components. Laboratory oxidation experiments were performed at 2 Torr ([approximately]270 Pa) and 15 Torr ([approximately]2 kPa) O2 pressure and temperatures in the range 100 to 350°C (373 to 623 K) for 2 to 8 h. After oxidation, components were examined for visual or mechanical change, and when appropriate, mass changes were also obtained. In some cases, optical diagnostics were also performed. The specimens were mostly spare/surplus components and spanned a wide variety of materials and functions, e.g., cryopump components; structural, mechanical, and diagnostic components; and fast-wave antennas. The effect of oxidation was found to be negligible for nearly all DIII-D components and materials tested.

We report new results from recent particle control and transport experiments in the DIII-D tokamak. We find that dynamic particle balance calculations (particle sources and sinks calculated vs time) yield similar results to a shot-averaged “static” (calculated by pressure rise) particle balance. The dynamic particle balance measurements show very low wall retention in both NBIand ECheated H-modes, compared with large retention during plasma startup. Two oxygen bakes at 1.3 kPa oxygen pressure for 2 hours at 350°C were completed and advanced inductive plasma operation was quickly recovered. No damage to tokamak internal components was observed. The removal of the hydrogenic species in DIII-D by thermo-oxidation compares well with laboratory experiments and we obtained data on thermo-oxidation of fresh C layers in DIII-D. Particle balance in discharges with resonant magnetic perturbation (RMP) ELM suppression show that the wall retention rate and inventory is dependent on pedestal density and divertor conditions. C injection experiments have shown that most of the carbon is deposited at the inner strike point of a SN divertor in L-mode, additional deposition in the private flux zone is present in H-mode. With an unbalanced DN plasma shape, there is more localized C deposition near the injection point in the non-active divertor.

E.A. Unterberg, S.L. Allen, A.W. Leonard, N.H. Brooks, J.W. Davis, B.W.N. Fitzpatrick, A.A. Haasz, G.L. Jackson, M.A. Mahdavi, C.K. Tsui, K.R. Umstadter, P.C. Stangeby, W.P. West, and the DIII-D Team. Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA 2 Lawrence Livermore National Laboratory, Livermore, California, USA General Atomics, PO Box 85608, San Diego, California 92186-5608, USA University of Toronto, Toronto, Ontario, CANADA University of California-San Diego, La Jolla, California, USA

Second-harmonic X-mode (X2) electron cyclotron (EC) heating (ECH) has been used in DIII-D in conjunction with plasma initiation and current ramp-up. Although the toroidal inductive electric field E-phi in DIII-D is high enough (0.9 to 1.0 V/m) to allow robust start-up without EC assist, start-up infusion devices such as ITER will have lower fields (E-phi = 0.3 V/m), and EC assist can provide a reproducible breakdown and an increased margin for burn-through of low-Z impurities. ECH, applied before the inductive electric, field, is used to separate the various phases of plasma breakdown and start-up and is defined as preionization. Preionization first occurs near the X2 resonance location and then expands in the vessel volume. Perpendicular launch (k(parallel to) = 0) is, found to produce the strongest preionization. The power threshold for preionization can be reduced by optimizing the prefill and the vertical, field, although the lowest power threshold is not at the optimum, value, for ohmic start-up alone. An orbit-following code confirms that cold electrons (0.03 eV) can be sufficiently heated by ECH to energies above the threshold of ionization of hydrogen. This code predicts heating in new tokamaks such as KSTAR and ITER to energies where preionization can occur. The ITER start-up scenario has been simulated in DIII-D experiments, and X2 ECH assist has been applied at reduced toroidal loop voltage to assist burn-through and plasma current ramp-up.

Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, B T, and the plasma current, I p, point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40–45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the B T direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.

Flow velocities of the order 10–20km/s in the direction of the high-field side divertor have been measured for deuterons and low charge-state carbon ions in the scrape-off layer at the crown of low-density L-mode plasmas, suggesting that these carbon ions at the crown move with the background plasma flow. Simulations with the multi-fluid edge code UEDGE including cross-field drifts due to E×B and B×∇B yield calculated divertor conditions which are more consistent with the measurements, but flows at the crown that are stagnant or in the opposite direction than observed. The simulations indicate that both the ion temperature gradient force and deuteron frictional drag play a role in determining the flow direction and magnitude of low charge-state carbon ions. The effect of the assumed radial transport model, toroidal core rotation, and neutral pumping at the divertor plates on the flow at the crown is investigated.

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicrometre sized dust is routinely observed using Mie scattering from a Nd : Yag laser. The source is strongly correlated with the presence of type I edge localized modes (ELMs). Larger size (0.005–1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust; on the other hand, large flakes or debris falling into the plasma may induce a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micrometre-size particles into plasma discharges. In DIII-D, a sample holder filled with 30–40 mg of dust is inserted in the lower divertor and exposed, via sweeping of the strike points, to the diverted plasma flux of high-power ELMing H-mode discharges. After a brief dwell (∼0.1 s) of the outer strike point on the sample holder, part of the dust penetrates into the core plasma, raising the core carbon density by a factor of 2–3 and resulting in a twofold increase in the radiated power. In TEXTOR, instrumented dust holders with 1–45 mg of dust are exposed in the scrape-off-layer 0–2 cm radially outside of the last closed flux surface in discharges heated with 1.4 MW of NBI. Launched in this configuration, the dust perturbed the edge plasma, as evidenced by a moderate increase in the edge carbon content, but did not penetrate into the core plasma.

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the porous plug injector (PPI), has been employed for in situ study of chemical erosion in the tokamak divertor environment by injection of CH4 [A.G. McLean et al., these Proceedings]. A new interpretive code, 3D-DIVIMP-HC, has been developed and applied to the interpretation of the CH, CI, and CII emissions. Particular emphasis is placed on the interpretation of 2D filtered-camera (TV) pictures in CH, CI and CII light taken from a view essentially straight down on the PPI. The code replicates sufficient measurements to conclude that most of the basic elements of the controlling physics and chemistry have been identified and incorporated in the code-model.

Naturally occurring and injected dust particles are observed in the DIII-D tokamak in the outer midplane scrape-off-layer (SOL) using a visible fast-framing camera, and the size of dust particles is estimated using the observed particle lifetime and theoretical ablation rate of a carbon sphere. Using this method, the lower limit of detected dust radius is ∼3μm and particles with inferred radius as large as ∼1mm are observed. Dust particle 2D velocities range from approximately 10 to 300m/s with velocities inversely correlated with dust size. Pre-characterized 2–4μm diameter diamond dust particles are introduced at the lower divertor in an ELMing H-mode discharge using the divertor materials evaluation system (DiMES), and these particles are found to be at the lower size limit of detection using the camera with resolution of ∼0.2cm2 per pixel and exposure time of 330μs.

An improved, self-contained gas injection system for the divertor material evaluation system (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH4 are found to be 58±12 in an attached divertor (ne∼1.5×1013/cm3, Te∼25eV, Tsurf∼450K), and 94±20 in a semi-detached cold divertor (ne∼6.0×1013/cm3, Te∼2–3eV, Tsurf∼350K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C+ at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH4 from the PPI, then compared with that seen in intrinsic sputtering.

In this paper the manipulation of power deposition on divertor targets at DIII-D by the application of resonant magnetic perturbations (RMPs) for suppression of large type-I edge localized modes (ELMs) is analysed. We discuss the modification of the ELM characteristics by the RMP applied. It is shown that the width of the deposition pattern in ELMy H-mode depends linearly on the ELM deposited energy, whereas in the RMP phase of the discharge those patterns are controlled by the externally induced magnetic perturbation. It was also found that the manipulation of heat transport due to the application of small, edge RMP depends on the plasma pedestal electron collisionality . We compare in this analysis RMP and no RMP phases with and without complete ELM suppression. At high , the heat flux during the ELM suppressed phase is of the same order as the inter-ELM and the no-RMP phase. However, below this collisionality value, a slight increase in the total power flux to the divertor is observed during the RMP phase. This is most likely caused by a more negative potential at the divertor surface due to hot electrons reaching the divertor surface from the pedestal area along perturbed, open field lines.

We have made two significant discoveries in our recent studies of quiescent H-mode (QH-mode) plasmas in DIII-D. First, we have found that we can control the edge pedestal density and pressure by altering the edge particle transport through changes in the edge toroidal rotation. This allows us to adjust the edge operating point to be close to, but below the ELM stability boundary, maintaining the ELM-free state while allowing up to a factor of two increase in edge pressure. The ELM boundary is significantly higher in more strongly shaped plasmas, which broadens the operating space available for QH-mode and leads to improved core performance. Second, for the first time on any tokamak, we have created QH-mode plasmas with strong edge co-rotation; previous QH-modes in all tokamaks had edge counter-rotation. This result demonstrates that counter-NBI and edge-counter rotation are not essential conditions for QH-mode. Both these investigations benefited from the edge stability predictions based on peeling–ballooning mode theory. The broadening of the ELM-stable region with plasma shaping is predicted by that theory. The theory has also been extended to provide a model for the edge harmonic oscillation that enhances edge transport in the QH-mode. Many of the features of that theory agree with the experimental results reported either previously or in this paper. One notable example is the prediction that co-rotating QH-mode is possible provided sufficient shear in the edge rotation can be created.

The effectiveness of the puff-and-pump technique to enrich a seeded impurity in the divertor relative to the core and, thereby, to maximize radiation in the divertor depends sensitively on both the magnetic geometry and the ion B × ∇B drift direction. In the puff-and-pump scenario used here, argon impurities injected into the private flux region are inhibited from accumulation in the core plasma by enhanced plasma flows to the divertor created by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping near the divertor targets. Modelling of single-null, H-mode plasmas with the UEDGE fluid transport code indicates that particle drifts in the scrape-off layer and divertor strongly affect the locations where the argon seed impurity accumulates. It is also found in double-null cases that argon always shows a larger accumulation in the divertor out of which the ion B × ∇B drift is directed, regardless of the divertor into which the argon is injected. Experiments have shown that the degree to which the deuterium gas-puffing rate inhibits the escape of the seed impurity from the divertor(s) depends critically on the direction of the ion B × ∇B drift and on whether the plasma is single-null or double-null. The transition in behaviour from double-null to single-null character during puff-and-pump occurs for |dR sep| = 0.4 cm when the ion B × ∇B drift was pointing away from the dominant divertor. The lowest argon density buildup in the main plasma of any of the configurations studied during puff-and-pump was achieved in single-null plasmas with the ion B × ∇B drift direction away from the divertor.