G.L. Jackson

Initial Boronization of JT-60U Tokamak Using Decaborane

A decaborane-based boronization system has been installed in the JT-60U tokamak in order to reduce the influx of impurities during plasma discharges. Boronization has been performed under a glow discharge using a helium-decaborane gas mixture. The properties of the boron films deposited through boronization and the effects of boronization on the tokamak discharges were investigated. It was found that the deposition of a boron layer with high purity was achieved with few impurities other than hydrogen through boronization, and that the present boronization deposited toroidally nonuniform boron film. It was also found that the decaborane-based boronization resulted in good plasma performance similar to that of conventional boronization.

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.

Higher fusion power gain with profile control in DIII-D tokamak plasmas

Strong shaping, favourable for stability and improved energy confinement, together with a significant expansion of the central region of improved confinement in negative central magnetic shear target plasmas, increased the maximum fusion power produced in DIII-D by a factor of 3. Using deuterium plasmas, the highest fusion power gain, the ratio of fusion power to input power, Q, was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma, which is similar to values achieved in tokamaks of larger size and magnetic field. A simple transformation relating Q to the stability parameters is presented

Reduction of divertor carbon sources in DIII-D

Abstract The evolution of carbon release from the DIII-D lower divertor tiles is studied using atomic and molecular spectroscopy. Newly installed virgin graphite tiles in 1992 are found to have had a chemical erosion yield, Ychem⩽ 3–5%, consistent with both laboratory results and similar experiments in other totamaks. The average Ychem measured in the DIII-D lower divertor decreased approximately a factor of ten between 1992 and 2000. The presumed cause of this reduction is the cumulative effect of >30 wall-conditioning boronizations and 105 s of plasma exposure, although the relative importance of these two mechanisms is unknown. This result indicates that a substantial reduction in carbon chemical erosion, and its relative importance as a source of carbon, can be obtained by long-term in situ wall conditioning techniques. The total carbon source sputtered into the DIII-D lower divertor has also apparently decreased over the same period. However, there has been no significant decrease in the average core carbon contamination with the decreasing lower divertor carbon source.

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.

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.

Comparing simulation of plasma turbulence with experiment

The direct quantitative correspondence between theoretical predictions and the measured plasma fluctuations and transport is tested by performing nonlinear gyro-Landau-fluid simulations with the GRYFFIN (or ITG) code [W. Dorland and G. W. Hammett, Phys. Fluids B 5, 812 (1993); M. A. Beer and G. W. Hammett, Phys. Plasmas 3, 4046 (1996)]. In an L-mode reference discharge in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)], which has relatively large fluctuations and transport, the turbulence is dominated by ion temperature gradient (ITG) modes. Trapped electron modes and impurity drift waves also play a role. Density fluctuations are measured by beam emission spectroscopy [R. J. Fonck, P. A. Duperrex, and S. F. Paul, Rev. Sci. Instrum. 61, 3487 (1990)]. Experimental fluxes and corresponding diffusivities are analyzed by the TRANSP code [R. J. Hawryluk, in Physics of Plasmas Close to Thermonuclear Conditions, edited by B. Coppi, G. G. Leotta, D. Pfirsch, R. Pozzoli, and E. Sind...

Impurity feedback control for enhanced divertor and edge radiation in DIII-D discharges

Long pulse and steady state fusion ignition devices will require a significant radiated power fraction to minimize heat flux to, and sputtering of, the first wall. While impurity gases have been proposed to enhance radiation, precise control of impurity gas injection is essential to achieve an adequate radiative power fraction while maintaining good energy confinement and low central impurity concentration. We report here the first experiments in the DIII-D tokamak using feedback control of the rate of impurity gas injection. These experiments were carried out with active divertor pumping using the in-situ DIII-D cryopump. The radiated power fraction was controlled by sensing either UN edge line radiation (Ne{sup +7}) or mantle radiation from selected bolometer channels and using the DIII-D digital plasma control system to calculate radiated power real-time and generate an error signal to control an impurity gas injector valve.

Divertor plasma studies on DIII-D: experiment and modelling

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.