An overview of the DIII-D program

The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q{sub DT} = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors

Recent results from the DIII-D tokamak and implications for future devices

Improvements to the DIII-D tokamak have led to significant new research results and enhanced performance. These results provide important inputs to the design of next generation divertor systems including the upgrade of the DIII-D divertor. The use of graphite for the plasma facing components and careful wall preparation has enabled the routine achievement of regimes of enhanced energy confinement. In elongated discharges, triangularity has been found to be important in attaining good discharge performance as measured by the product of the normalized plasma pressure and the energy confinement time, {beta}{tau}{sub E} This constrains the design of the divertor configuration (X-point location). Active pumping of the divertor region using an in-situ toroidal cryogenic pump has demonstrated control of the plasma density in H-mode discharges and allowed the dependence of confinement on plasma density and current to be separately determined. Helium removal from the plasma edge sufficient to achieve effective ash removal in reactor discharges has also been demonstrated using this pumping configuration. The reduction of the heat flux to the divertor plates has been demonstrated using two different techniques to increase the radiation in the boundary regions of the plasma and thus reduce the heat flux to the divertor plates; deuterium gas injection has been used to create a strongly radiating localized zone near the X-point, and impurity (neon) injection to enhance the radiation from the plasma mantle. Precise shaping of the plasma current profile has been found to be important in achieving enhanced tokamak performance. Transiently shaped current profiles have been used to demonstrate regimes of plasmas with high beta and good confinement. Control of the current profile also is important to sustaining the plasma in the Very High (VH)-mode of energy confinement

Scrape-off layer ion acceleration during fast wave injection in the DIII-D tokamak

Fast wave injection is employed on the DIII-D tokamak as a current drive and electron heating method. Bursts of energetic ions with energy E o>20keV are observed immediately following fast wave injection in experiments featuring the 8th ion cyclotron harmonic near the antenna. Using the energy and pitch angle of the energetic ion burst as measured by a fast-ion loss detector, it is possible to trace the origin of these ions to a particular antenna. The ion trajectories exist entirely within the scrape-off layer. These observations are consistent with the presence of parametric decay instabilities near the antenna strap. It is suggested that the phase space capabilities of the loss detector diagnostic can improve studies of wave injection coupling and efficiency in tokamaks by directly measuring the effects of parametric decay thresholds. © 2012 IAEA, Vienna

DIII-D experimental plan for FY-1989

This document summarizes the Experimental Plan for the DIII-D tokamak facility for the fiscal year 1989. The long-range DIII-D 5 yr plan is directed ultimately at the goal of achieving good confinement at high beta in a plasma with non-inductively driven current. This is important to the design of a steady-state reactor. This program may be thought of as occurring in two phases. In the first phase of the program we axe separately investigating high beta plasma confinement in inductively-driven plasmas, and non-inductive current drive. In the second phase we will combine these two elements to investigate high beta plasma confinement with non-inductive current drive. The FY 89 plan continues the first phase of the DIII-D experimental effort that contains a strong focus on beta and confinement in non-circular plasma configurations and in the divertor configuration in particular. Important work also continues in the development of rf heating systems for heating, profile control, and current drive. This research is coupled to theoretical efforts at General Atomics. The FY 89 research program outlined herein is diverse and multifaceted. However, it is also characterized by a greater synthesis of techniques toward a common goal. An example is the application of ECH for sawtooth suppression that would improve the low q confinement and allow higher {beta} to be obtained. We believe this research program will provide a solid foundation for the continued development of the tokamak toward high beta steady-state reactor application. The DIII-D FY 89 research program will provide results that will help resolve many CIT and ITER Physics R&D issues. In addition, DIII-D confinement studies will be an important input to the newly formed National Transport Task Force

Updated DIII-D experimental plan for FY-1989

The program proposed here is designed to support and build toward the long-term plan put forward during 1987 for the DIII-D facility. This plan has as its ultimate goal developing sufficient understanding and predictive capability to enable the demonstration of a high beta plasma with non-inductively driven toroidal current. The early stages of this plan call for the optimization of the plasma configuration for good confinement at high beta while simultaneously developing the need rf power systems for current drive, profile control, and heating

Lower hybrid wave electron heating experiments in Doublet IIA

Experiments designed to heat electrons by Landau damping of waves at approximately twice the lower hybrid frequency have been carried out on Doublet IIA. This objective is in contrast to other lower hybrid experiments which are designed to heat ions using frequencies corresponding to the lower hybrid resonance frequency. Up to 500 kW of rf power was applied to discharge with approximately 100 kW ohmic input using parallel wavelengths chosen to optimize the spatial distribution of the power deposition based on linear or quasi-linear Landau damping

Anomalies in the Applied Magnetic Fields on Diii-D and Their Implications for the Understanding of Stability Experiments

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks

Conceptual design summary for modifying Doublet III to a large dee-shaped configuration

The Doublet III tokamak is to be reconfigured by replacing its indented (doublet) vacuum vessel with a larger one of a dee-shaped cross section. This change will permit significantly larger elongated plasmas than is presently possible and will allow higher plasma current (up to 5 MA) and anticipated longer confinement time. Reactor relevant values of stable beta and plasma pressure are predicted. This modification, while resulting in a significant change in capability, utilizes most of the existing coils, structure, systems and facility