User Facilities

DIII-D National Fusion Facility (DIII-D)

San Diego, California
Start of Operations
Number of Users
618 (FY 2016)

DIII-D, the largest magnetic fusion user facility in the U.S., is a tokamak confinement device with significant engineering flexibility to explore the optimization of the advanced tokamak approach to fusion energy production.


The DIII-D National Fusion Facility (DIII-D), at General AtomicsExternal link, is the largest magnetic fusion research experiment in the U.S., with a program mission to establish the scientific basis for the optimization of the tokamak approach to fusion energy production. Capabilities of this facility include a flexible field-shaping coil system to produce a variety of plasma shapes, a diverse mix of auxiliary heating and current drive systems, coil sets both inside and outside the vacuum vessel that are used to correct error fields and study the plasma response to perturbing magnetic fields, all-carbon plasma-facing surfaces, over 50 diagnostic systems to examine plasma parameters, and an advanced digital control system for feedback control of the plasma.


Research at this facility contributes to world-leading science by 1) providing solutions to physics and operational issues critical to the success of ITER; 2) developing the physics basis for steady-state tokamak operation required for efficient power production; 3) contributing substantially to the technical basis for a Fusion Nuclear Science Facility (FNSF); and 4) advancing the fundamental understanding and predictive capability of fusion science.  Research is conducted to:

  • Resolve the disruption problem for the tokamak through advanced stability control (3D coils and electron cyclotron heating), plasma quench mitigation systems, and innovative diagnostic measurements.
  •  Provide the basis for solutions that resolve the divertor challenge for FNSF and a future demonstration power plant (DEMO) through advanced divertor and plasma material interaction studies.
  • Explore the physics of the burning plasma state using electron cyclotron heating for dominant electron heating at low injected torque (with variable co/counter neutral beam injection).
  • Investigate and understand the conditions required for steady-state tokamak operation through broadened current distributions using off axis neutral beam injection capability.
  • Develop the three-dimensional (3D) optimization of the tokamak concept to improve edge stability, rotation and core mode control, through the use of 3D perturbation coils.
Last modified: 10/24/2017 9:02:00 AM