Recent results and planned upgrades for the DIII-D Tokamak

A major focus of the experimental program on the DIII-D tokamak is the need to better define the optimal design requirements and operating scenarios of future devices such as ITER. Using the flexible DIII-D plasma control system, several of the proposed ITER operating scenarios have been investigated and experiments have been carried out to demonstrate control of a variety of plasma instabilities and mitigate the effects of plasma disruptions. Several major upgrades to the DIII-D device are being planned for the near future to allow research to continue to address critical issues that arise as tokamak research advances into the regime of longer pulse lengths, fully non-inductive current drive, and higher performance operation. The auxiliary heating systems of the tokamak will be upgraded by increasing the ECH power to 12 MW and extending the neutral beam power pulse length to 10 s. Important plasma current profile effects will be addressed by modifying two of the neutral beam injectors to allow up to 10 MW of off-axis heating. Also, a flexible array of feedback control coils will be installed inside the vacuum vessel to continue the investigation of the control of ELMs and other MHD phenomena. A description of these upgrades and a discussion of some of the important recent results will be presented.