In magnetic confinement fusion, a divertor is a magnetic field configuration which diverts the heat and particles escaped from the magnetically confined plasma to dedicated plasma-facing components, thus spatially separating the region plasma-surface interactions from the confined core (in contrast to the limited configuration). This requires establishing a separatrix-bounded magnetic configuration, typically achieved by creating poloidal field nulls (X-points) using external coils.
The divertor is a critical part of magnetic confinement fusion devices, first introduced by Lyman Spitzer in the 1950s for the stellarator concept. [1][2] It extracts heat and ash produced by the fusion reaction while protecting the main chamber from thermal loads, and reduces the level of plasma contamination due to sputtered impurities. In tokamaks, high confinement modes are more readily achieved in diverted configurations.
At present, it is expected that future fusion power plants will generate divertor heat loads greatly exceeding the engineering limits of the plasma-facing components. The search for mitigation strategies to the divertor power exhaust challenge is a major topic in nuclear fusion research.
- ^ Spitzer, Lyman (1958). "The Stellarator Concept". The Physics of Fluids. 1 (4): 253–264. Bibcode:1958PhFl....1..253S. doi:10.1063/1.1705883. Retrieved 2024-10-23.
- ^ Burnett, C. R.; Grove, D. J.; Palladino, R. W.; Stix, T. H.; Wakefield, K. E. (1958). "The Divertor, a Device for Reducing the Impurity Level in a Stellarator". The Physics of Fluids. 1 (5): 438–445. Bibcode:1958PhFl....1..438B. doi:10.1063/1.1724361. Retrieved 2024-10-23.