The meeting was a unique occasion to discuss divertor physics in various magnetic configurations, novel ideas for next step devices and, most importantly, the ITER divertor – the largest and most complex divertor ever constructed. Although this design reflects current understanding and technology, requirements to remove access power and impurities generated during the reaction from the plasma needs further development for future fusion power plants.

Inside a tokamak – the donut-shaped fusion device, a plasma is heated to hundreds of millions of degrees and confined by magnetic fields that keep it from the wall of the machine. The divertor is the ‘exhaust pipe’ target chamber located at the very bottom of the reactor, where impurities such as Helium ‘ash’ are diverted through outer magnetic field lines to a location far from the plasma.

This configuration produces ‘purer’ plasmas with better energy confinement – a critical parameter for the performance of a fusion device, thus maintaining the plasma hot enough for long enough, for fusion reactions to take place on a substantial scale, and at the same time, protecting the surrounding walls from thermal and neutronic loads.

In ITER – the international fusion reactor scale experiment being assembled in France – the divertor will be made up of 54 ten tons cassettes. These will need to be replaced by remote handling at least once during the machine's lifetime. Watch this video to see how remote handling tasks are carried out at JET, currently the world’s largest experimental fusion device.

Divertor geometries

The ITER divertor will face heat fluxes of 10-20 MW per square metre, ten times higher than the heat load on a spacecraft re-entering Earth's atmosphere! Its most exposed surface will face a maximum expected temperature between 1000 °C and 2000 °C. For this reason, and for its low tritium (fusion fuel) retention, Tungsten, a refractory metal that melts only at a very high temperature (3400 °C), has been chosen as the armour material for the components facing the plasma.

Future ITER operation experience will provide much of the technological basis for the design of future fusion power plants capable of generating electricity, including tests of divertor systems. Finding more resistant materials and the optimal magnetic configuration to better handle the heat fluxes at the divertor is a main focus of ongoing fusion research and development.

Designing divertor systems with pumps able to distinguish between Helium generated in fusion of Deuterium and Tritium (two Hydrogen isotopes) and unburned fusion fuel is another focus of ongoing research.

"The core plasma operational scenario imposes a number of constraints on the divertor design systems for reactor-scale tokamaks," said Anthony Leonard, scientist at the US General Atomics, and chair of the meeting.

"In the core we want high plasma density values to achieve high levels of fusion power and these need to be compatible with the values determined by the divertor geometry at the plasma edge," said Saskia Mordijck, an Assistant Professor at the US College William and Mary.