The exhaust system of a magnetic confinement fusion device, responsible for removing waste heat, helium ash, and plasma impurities while protecting the vessel walls from direct plasma contact.
In a magnetic confinement fusion device such as a tokamak, the plasma must be kept away from the surrounding vessel walls. The divertor accomplishes this by using shaped magnetic fields to channel the outermost layer of plasma—the scrape-off layer (SOL)—away from the confined core and onto dedicated target plates.[1]
Most modern tokamaks employ an axisymmetric single-null divertor, in which the poloidal magnetic field is shaped to create an X-point—a magnetic null where field lines diverge.[2]
In a reactor-scale device, the power crossing the separatrix can exceed 100 MW, yet the wetted area may be only a few square meters. Two primary strategies reduce the heat load:
The second strategy is the detached divertor regime, where the plasma temperature near the target drops below ~5 eV, forming a radiating cushion that dramatically reduces heat flux.[1]
ITER will use tungsten for its divertor targets, selected for its high melting point (3422 °C), low sputtering yield, and low tritium retention. The ITER divertor is designed to be replaced remotely multiple times.[4]
These include the Super-X divertor (tested on MAST Upgrade), snowflake divertor, and liquid-metal divertors. No single solution has yet been validated at reactor scale.[2]