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 of the tokamak systems. These results provide important inputs to the design of next-generation divertor systems including the upgrade of the DIII-D divertor. Key results have been obtained in understanding the divertor configuration and developing effective configurations for new devices. 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 B tau(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 to 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. Ongoing and planned enhancements to the DIII-D tokamak systems will allow us to exploit these gains further. The recent upgrade of the ICRF system to 6 MW should allow sufficient current drive to begin independent control of the plasma current. Work is beginning on a radiative divertor modification to the vessel interior which will combine the desirable features of highly triangular plasma shape with strong pumping and baffling to minimize the back-streaming of gas and impurities.