Vacuum vessel of the Tokamak à Configuration Variable (TCV).

To create plasma in a tokamak, a controlled amount of working gas is introduced into a vacuum vessel — a hollow steel chamber in the shape of a torus. The plasma is created inside the vessel and kept away from its walls by magnetic fields. Although the plasma temperature can exceed 100 million kelvin, the surrounding components are maintained at temperatures of only a few hundred degrees by magnetic confinement and active cooling.

Material

The vessel is typically made of stainless steel, with cooling channels incorporated into the walls of larger tokamaks. Currents induced in the conducting vessel can help stabilise certain rapid plasma motions and magnetic instabilities.

First Wall

First wall panels are mounted on the inner side of the vacuum vessel and directly face the plasma. They must withstand high heat fluxes during normal operation as well as much larger transient loads during events such as edge-localised modes (ELMs) and plasma disruptions.

The first wall is covered by replaceable plasma-facing components made from refractory materials chosen to withstand high heat loads and minimise contamination of the plasma. Carbon, beryllium and tungsten have all been used in fusion devices; ITER now plans to use tungsten armour for its first wall.

The material must have a high melting point, resist erosion and sputtering, and release as few impurity atoms as possible into the plasma. High-Z impurities can cause substantial radiative energy losses. The material should also retain as little hydrogen isotopes and other gases as possible, because trapped species can later be released back into the plasma. Gas release from the walls can be reduced by baking the vacuum vessel before operation, allowing adsorbed gases to desorb and be removed by the vacuum pumps.

Tritium Retention

The ability of plasma-facing materials to retain particles can become a serious issue in a fusion power plant, especially when tritium is used as fuel. Tritium trapped in wall materials is unavailable for the fuel cycle and increases the radioactive inventory of the machine. Plasma-facing materials are therefore selected and conditioned to minimise tritium retention.

3D rendering of the interior of a tokamak vacuum vessel.

Neutron Flux

Cross-section of the ITER vacuum vessel.

In a fusion power plant, the vacuum vessel and surrounding structures will be exposed to an intense flux of 14.1 MeV neutrons from the deuterium–tritium reaction. These neutrons can activate materials and cause radiation damage and embrittlement, so radiation-resistant structural materials and effective shielding are required. Activated in-vessel components will require extensive remote maintenance.

Shielding

In a future fusion power plant, the kinetic energy of fusion neutrons will be absorbed in the blanket and converted into heat, which can then be removed by a coolant and used for electricity generation.

Future D–T fusion power plants are expected to use lithium-containing breeding blankets behind the first wall. Fusion neutrons will react with lithium to produce tritium, which can then be extracted and recycled as fuel. ITER will not operate with a full breeding blanket, but it will test four representative Test Blanket Module concepts during its deuterium-tritium phase.

Together, the first wall, breeding blanket and neutron shielding form a substantial layer between the plasma and the superconducting magnets, which is one reason fusion tokamaks must be very large machines.

Fusion reactor blanket and neutron shielding.

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Neutron shielding and tritium-breeding blanket in stellarator chamber

Vacuum

ITER tokamak vacuum vessel.

A high vacuum is maintained by a combination of mechanical pumps and powerful cryopumps. In ITER, the vessel will be evacuated to about one millionth of normal atmospheric pressure before plasma operation. The cryopumps use cryopanels cooled with supercritical helium and coated with activated charcoal, which acts as a sorbent for helium and hydrogen isotopes such as tritium.

Geometry

The geometry of a tokamak is often characterised by its aspect ratio, A = R/a, where R is the major radius — the distance from the central axis of the torus to the centre of the plasma — and a is the minor radius of the plasma cross-section. A low aspect ratio produces a compact, nearly spherical plasma shape, while a high aspect ratio corresponds to a thinner torus with a larger central opening. Tokamaks with particularly low aspect ratios are known as spherical tokamaks.

4 pictures

4 pictures