Magnetic confinement fusion device component
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. 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.
Tokamak divertors.
A tokamak featuring a divertor is known as a divertor tokamak or divertor configuration tokamak. In this configuration, the particles escape through a magnetic "gap" (separatrix), which allows the energy absorbing part of the divertor to be placed outside the plasma.
The divertor configuration also makes it easier to obtain a more stable H-mode of operation. The plasma facing material in the divertor faces significantly different stresses compared to the majority of the first wall.
Stellarator divertors.
In stellarators, low-order magnetic islands can be used to form a divertor volume, the island divertor, for managing power and particle exhaust. The island divertor has shown success in accessing and stabilizing detached scenarios and has demonstrated reliable heat flux and detachment control with hydrogen gas injection, and impurity seeding in the W7-X stellarator. The magnetic island chain in the plasma edge can control plasma fueling. Despite some challenges, the island divertor concept has demonstrated great potential for managing power and particle exhaust in fusion reactors, and further research could lead to more efficient and reliable operation in the future.
The helical divertor, as employed in the Large Helical Device (LHD), utilizes large helical coils to create a diverting field. This design permits adjustment of the stochastic layer size, situated between the confined plasma volume and the field lines ending on the divertor plate. However, the compatibility of the Helical Divertor with stellarators optimized for neoclassical transport remains uncertain.
The non-resonant divertor provides an alternative design for optimized stellarators with significant bootstrap currents. This approach leverages sharp "ridges" on the plasma boundary to channel flux. The bootstrap currents modify the shape, not the location, of these ridges, providing an effective channeling mechanism. This design, although promising, has not been experimentally tested yet.
Given the complexity of the design of stellarator divertors, compared to their two-dimensional tokamak counterparts, a thorough understanding of their performance is crucial in stellarator optimization. The experiments with divertors in the W7-X and LHD have shown promising results and provide valuable insights for future improvements in shape and performance. Furthermore, the advent of non-resonant divertors offers an exciting path forward for quasi-symmetric stellarators and other configurations not optimized for minimizing plasma currents.
References.
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