R. D. Stambaugh

Physics of the L-mode to H-mode transition in tokamaks

Combined theoretical and experimental work has resulted in the creation of a paradigm which has allowed semi-quantitative understanding of the edge confinement improvement that occurs in the H-mode. Shear in the E*B flow of the fluctuations in the plasma edge can lead to decorrelation of the fluctuations, decreased radial correlation lengths and reduced turbulent transport. Changes in the radial electric field, the density fluctuations and the edge transport consistent with shear stabilization of turbulence have been seen in several tokamaks. The purpose of this paper is to discuss the most recent data in the light of the basic paradigm of electric field shear stabilization and to critically compare the experimental results with various theories.

Confinement physics of H-mode discharges in DIII-D

The authors data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

Investigation of physical processes limiting plasma density in high confinement mode discharges on DIII-D

A series of experiments was conducted on the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit [M. Greenwald et al., Nucl. Fusion 28, 2199 (1988)] was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased nonlinearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was ⩽3 MW. Formation of MARFEs on closed field lines was avoided with lo...

Physics of the detached radiative divertor regime in DIII-D

This paper summarizes results from a two-dimensional (2D) physics analysis of the transition to and stable operation of the partially detached divertor (PDD) regime induced by deuterium injection in DIII-D. The analysis [1] shows that PDD operation is characterized by a radiation zone near the X-point at -15 eV which reduces the energy flux into the divertor and thereby also reduces the target plate heat flux, an ionization zone below the X-point which provides a deuterium ion source to fuel parallel flow down the outer divertor leg, an ion-neutral interaction zone in the outer leg which removes momentum and energy from the flow and finally a volume recombination zone above the target plate which reduces the particle flux to the low levels measured on the plates and thereby also contributes to reduction in target plate heat flux.

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.

A design study for an advanced divertor for DIII-D and ITER: the radiative slot divertor

Reduction of the divertor heat load is an important issue for future tokamaks, particularly during the technology phase of ITER. We discuss a conceptual design for one type of advanced divertor: the radiative slot divertor. The goal of this divertor configuration is to enhance the radiation in the divertor region and thereby reduce the heat load at the strike points. At the same time, any effects on the core plasma must be minimized. Proof-of-principle experiments to enhance the radiation in the DIII-D divertor have been performed both with deuterium and impurity injection. We compare several computer models with results from these experiments to predict performance and thereby guide designs of radiative divertors for future machines. We have estimated impurity radiation using calculations of the background plasma with a two-dimensional fluid code (B2 or LEDGE) coupled with models of impurity radiation. The DEGAS code has been used to estimate hydrogenic transport, charge exchange and radiation losses. Estimates of impurity transport are provided by 11/1-dimensional models and calculations of impurity frictional-force terms. These model, results are in qualitative agreement with the ∼1 MW reduction of measured divertor power in DIII-D during divertor impurity puffing experiments. Specific designs, including engineering details, for applications to DIII-D and ITER will be discussed.

Two‐stage turbulence suppression and E×B velocity shear measured at the L–H transition

At the onset of the L–H transition in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p.159], a fast (≊100 μsec) suppression of microturbulence is observed as the edge transport barrier is formed. This fast edge suppression is followed by a much slower (tens of msec), but substantial (≥50%) reduction in the relative density fluctuation level. This second turbulence suppression phase, which is observed to correlate with growing E×B velocity shear, has been localized to the plasma interior, and may explain why the observed transport reduction in the H mode has been observed to extend deep into the plasma, well beyond the edge transport barrier.

Recent results from DIII-D and their implications for next generation tokamaks

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

E×B transport in the DIII-D boundary plasma

We have measured the electrostatic turbulence and associated particle transport in the DIII-D boundary plasma using a fast reciprocating Langmuir probe array located on the outboard midplane. Both the normalized rms fluctuation levels (density and floating potential) and the fluctuation-driven particle transport are altered by the L-H transition in the SOL. At the separatrix, the density fluctuation level is reduced a factor of 2, consistent with reflectometry results. There is a corresponding decrease in the turbulent particle flux. Deeper in the SOL, the turbulent particle transport in H-mode exceeds the L-mode value. The perpendicular diffusion coefficient D ⊥ and particle confinement time τ p have been estimated, assuming that the transport is purely turbulent and uniform on a flux surface. We find D ⊥ =0.7 D B (L) and 0.04 D B (ELM-free H), and τ p =54 ms (L) and 480 ms (ELM-free H).

Scrape-off layer measurements in DIII-D

In this paper, scrape-off layer measurements in DIII-D are presented as a function of the main discharge plasma parameters. A systematic study is under way to understand and predict the behavior of the edge and divertor plasma in DIII-D and this scaling behavior will be crucial for the design of ITER. To facilitate the studies, a fast reciprocating Langmuir probe incorporating five graphite tips was installed at the midplane of DIII-D which has the capability of performing multiple plunges 1 cm inside the separatrix during 5 MW of NBI. Density and temperature profiles in the midplane (reciprocating probe), near the top (Thomson scattering) and at the lower divertor plate (fixed Langmuir probe array) are compared by mapping the measurements into magnetic flux coordinates. Local pressure measurements are compared on different parts of a flux surface. The three different local measurements also indicate the spatial evolution of plasma conditions as plasma approaches the divertor plate. Ohmic and L-mode discharges exhibit similar (exponential) density and temperature decay in the scrape-off layer. H-mode discharges, however, display a faster spatial decay reflecting at least a factor of 3 decrease in the perpendicular diffusion coefficient. Consistency of the magnitude and scaling behavior of the edge profile parameters with models of the scrape-off layer is examined.