The confinement in a tokamak reactor is the result of magnetic field lines forming a set of closed magnetic surfaces. At the edge of the plasma a thin (~ 1 cm) region of open field lines is created – the scrape-off layer (SOL) – through which charged particles and heat flowing out of the core plasma are guided into a so-called divertor where the plasma impinges on a material surface (the divertor target plates). The parallel heat flux in the SOL region of ITER and DEMO is expected to be even higher than sun’s surface.

 

The current strategy is to optimise the operation with a conventional divertor based on detached conditions to be tested on the ITER device currently under construction in Cadarache. This strategy is based on different factors:

  • development of plasma facing components to cope with very large power fluxes (>5 MW/m2)
  • selection of divertor geometry and magnetic flux expansion to reduce the normal heat flux on the target, i.e., distributing the heat over a larger surface
  • removal of plasma energy before it reaches the target via impurity radiation by increasing edge plasma density and injecting impurities in the SOL region, so as to decrease the fraction of the heating power that leaves the vessel via the SOL channel.
  • recycling and increase of density lowering the temperature close to the target, with consequent detachment (the temperature drops below ionization, therefore the particles are neutralized and there is no direct plasma flux nor power to the divertor target)

 

However the successful operation of conventional divertor solution in ITER is not guaranteed:

  • the present experiment operate with SOL conditions that are very different from those expected in ITER and DEMO
  • the simulations with the present SOL models and codes are not fully reliable when extrapolating to ITER and DEMO conditions
  • the stability of the detachment front has to be assessed for ITER and DEMO conditions
  • there might be problems are related to integration with the core plasma, e.g.:
    • impurity contamination of the core with consequent reduction of fusion performance
    • compatibility of plasma radiation fraction

 

Therefore a specific project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a Divertor Tokamak Test facility. This tokamak should be capable of hosting scaled experiments integrating most of the possible aspects of the DEMO power and particle exhaust.

 

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