W.P. West

Radiative divertor experiments in DIII-D with D2 injection

Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: Ip=1.0-2.0 MA, q95 approximately= 2.4-6.0 and total input power (20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q95 equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

Physics of confinement improvement of plasmas with impurity injection in DIII-D

External impurity injection into L mode edge discharges in DIII-D has produced clear confinement improvement (a factor of 2 in energy confinement and neutron emission), reduction in all transport channels (particularly ion thermal diffusivity to the neoclassical level), and simultaneous reduction of long wavelength turbulence. Suppression of the long wavelength turbulence and transport reduction are attributed to synergistic effects of impurity induced enhancement of E × B shearing rate and reduction of toroidal drift wave turbulence growth rate. A prompt reduction of density fluctuations and local transport at the beginning of impurity injection appears to result from an increased gradient of toroidal rotation enhancing the E × B shearing. Transport simulations carried out using the National Transport Code Collaboration demonstration code with a gyro-Landau fluid model, GLF23, indicate that E × B shearing suppression is the dominant transport suppression mechanism.

Impurity enrichment studies with induced scrape-off layer flow on DIII-D

A series of controlled experiments has been carried out in DIII-D to induce a bulk ion flow in the SOL and evaluate its effect on the localization of impurities in the divertor. This induced SOL flow was created by simultaneous deuterium puffing and divertor exhaust using a divertor cryopump, and the impurity enrichment was measured directly. The experiments were designed to compare enrichment in discharges with and without induced flow having otherwise similiar divertor parameters. Significant increases in impurity compression and enrichment are observed when flow is induced, and the degree of impurity enrichment in the divertor is found to be dependent on the impurity of interest. Detailed particle measurements made possible by the direct measurement of impurity densities in several reservoirs indicate reasonable particle balance for helium throughout the duration of the discharge. Conversely, while the total input of neon is balanced by the total exhaust by the end of a discharge, particle balance is not observed during the course of the discharge. A significant wall inventory with a short release time (~10 ms) is surmised.

Divertor heat and particle control experiments on the DIII-D tokamak

Abstract In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D 2 gas puffing and impurity injection were studied separately and in both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

Erosion and deposition of metals and carbon in the DIII-D divertor

Net erosion rates at the outer strike point of the DIII-D divertor plasma were measured for several materials during quiescent H-mode operation with deuterium plasmas. Materials examined include graphite, beryllium, tungsten, vanadium and molybdenum. For graphite, net erosion rates up to 4 nm/sec were found. Erosion rates for the metals were much smaller than for carbon. Ion fluxes from Langmuir probe measurements were used to predict gross erosion by sputtering. Measured net erosion was much smaller than predicted gross erosion. Transport of metal atoms by the plasma across the divertor surface was also examined. Light atoms were transported farther than heavy atoms as predicted by impurity transport models.

Long pulse high performance discharges in the DIII-D tokamak

Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance {approx}10 has been sustained for >5 {tau}{sub E} with q{sub min} >1.5. (The normalized performance is measured by the product {beta}{sub N} H{sub 89} indicating the proximity to the conventional {beta} limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and {beta} {approx}{le} 5%. The limit to increasing {beta} is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate {approx}75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and {beta} control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with {beta}{sub N}H{sub 89} {approx} 7 for up to 6.3 s or {approx} 34 {tau}{sub E}. These discharges appear to be in resistive equilibrium with q{sub min} {approx} 1.05, in agreement with the current profile relaxation time of 1.8 s.

Measurement and verification of zeff radial profiles using charge exchange recombination spectroscopy on DIII-D

Charge exchange recombination (CER) spectroscopy in the visible spectrum is used to measure the radial ion density distribution of impurities in the core plasma of DIII-D. The radial profile of the effective ionic charge, Zeff(r), is subsequently calculated from the impurity densities and the electron density of the plasma. The CER measured radial distributions rely on a calculated neutral beam attenuation radial profile, which is confirmed by independent measurement. This technique, which determines the deuterium density of the neutral beam by coupling measured beam Dα emissions with a time dependent collisional radiative calculation, will be described. The CER derived absolute density/concentrations of carbon are verified by comparisons with the spectrometrically measured visible bremsstrahlung emission of the core plasma, which is proportional to Zeff. Conversely, the seeded neon concentration is overestimated by a factor of 1.7 by CER. This correction is shown to be caused by the enhanced direct capture into the upper level of the measured visible neon transition (Ne X n = 11 to 10, 5249 A) from excited (n = 2) beam atoms. Owing to several problems, including line radiation contamination of the spectral region of the diagnostic, the standard Zeff(r) derived from inversion of line integrated visible bremsstrahlung emissions does not provide reliable profiles, but rather a measure of the average impurity content. The Zeff profiles are found to vary considerably in shape and magnitude over different operational regimes, confirming the need for accurate profiles.

Comprehensive 2D measurements of radiative divertor plasmas in DIII-D

Abstract This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (partially detached divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2D ( R, Z ) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. D α emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D 2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. L α and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.

Radiative divertor plasmas with convection in DIII-D

The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features.

Divertor heat and particle flux due to ELMs in DIII-D and ASDEX-upgrade

The authors characterize the divertor target plate heat and particle fluxes that occur due to Edge-Localized-Modes (ELMs) during H-mode in DIII-D and ASDEX-Upgrade. During steady-state ELMing H-mode the fraction of main plasma stored energy lost with each ELM varies from 6% to 2% as input power increases above the H-mode power threshold. The ELM energy is deposited near the strikepoints on the divertor target plates in a fast time scale of {le} 1 ms. The spatial profile of the ELM heat pulse is flatter and broader, up to about a factor of 2, than that of the heat flux between ELMs. On ASDEX-Upgrade the inboard strike-point receives the greatest fraction, {ge} 75%, of ELM divertor heat flux, while on DIII-D the in/out split is nearly equal. The toroidal asymmetry of the heat pulse has produced a peaking factor on DIII-D of no more than 1.5. The particle flux, as measured by Langmuir probes, has also been found to be localized near the divertor strike-points. The increased particle flux during ELMs is a significant fraction of the total time-integrated divertor plate particle flux.