Structure of a tritium atom.

ITER will be the first fusion device designed to operate extensively with the deuterium–tritium fusion reaction. Deuterium can be extracted from water and is a virtually inexhaustible resource, while tritium is present naturally only in trace amounts and is much more difficult to obtain.

Tritium Origin

Tritium is produced naturally in the upper atmosphere by interactions between cosmic rays and atmospheric gases. Additional tritium remains in the environment from atmospheric nuclear testing, while limited quantities are also produced as a by-product in heavy-water reactors such as CANDU reactors.

Production

Tritium has a half-life of only 12.3 years, so stored inventories continuously decrease through radioactive decay. The limited tritium available from heavy-water reactors is expected to supply ITER’s deuterium–tritium programme, but future fusion power plants would require much larger quantities. A large fusion power plant could consume on the order of 100–200 kg of tritium per year, making tritium self-sufficiency essential.

Transmutation

Tritium can instead be bred from lithium, an element widely available in the Earth’s crust and in seawater. In the D–T fusion reaction, a helium nucleus and a high-energy neutron are produced. Because the neutron has no electric charge, it is not confined by the magnetic field and escapes from the plasma. In a future fusion power plant, these neutrons would be absorbed mainly in a breeding blanket, where their energy would be converted into heat for electricity production.

When a neutron is absorbed by a lithium-6 nucleus, it splits into a helium-4 nucleus and a tritium nucleus, releasing about 4.8 MeV of energy: ⁶Li + n → ⁴He + T + 4.8 MeV. This tritium-breeding reaction can provide an in-situ source of fuel for future fusion power plants while also contributing additional heat to the blanket. Once a closed tritium fuel cycle is established, the main external fuel resources required by a fusion power plant would be deuterium and lithium, both widely available elements. Figuratively speaking, fusion energy could therefore be produced from water and rock.

Tritium Breeding Tests

ITER will test tritium-breeding technologies through four Test Blanket Module concepts installed in two equatorial ports during deuterium–tritium operation. The concepts combine different breeder materials and cooling systems: a European water-cooled lead-lithium system, a Japanese water-cooled ceramic breeder, a Chinese helium-cooled ceramic breeder, and a European-Korean helium-cooled ceramic-pebble system. Together, they will provide data on tritium production and extraction, heat removal, materials and the performance of complete breeding-blanket systems.

Liquid Breeders

In the liquid-breeder concept, a lead–lithium alloy (PbLi) serves as both the tritium-breeding material and a neutron multiplier. Neutron reactions in lead can increase the number of neutrons available for further interactions with lithium, helping to improve the tritium breeding ratio. Tritium can be continuously extracted from the liquid PbLi breeder, although the electrically conducting alloy interacts with the strong magnetic field and creates additional magnetohydrodynamic challenges. These magnetohydrodynamic effects can be reduced by limiting the flow velocity and by electrically insulating the liquid metal from the conducting channel walls where appropriate.

Solid Breeders

Solid-breeder concepts use lithium-containing ceramic materials, often in the form of small pebbles. Unlike liquid metals, they do not produce significant magnetohydrodynamic effects. Tritium released from the heated breeder material is continuously carried away by a purge gas and transferred to a tritium extraction system; the breeder module therefore does not have to be removed each time tritium is collected.

Cooling Options

The heat generated in a breeding blanket must be removed by a coolant. Water-cooled concepts use pressurised water to permit operation above the normal boiling point, while helium-cooled concepts can operate at substantially higher temperatures but require high pressure and significant pumping power because of helium’s low density and volumetric heat capacity. The ITER Test Blanket Modules are compact experimental systems, approximately 1.7 × 0.5 × 0.6 metres, intended to measure tritium breeding and extraction in a real fusion-neutron environment.

A future fusion power plant would surround much of the plasma with a breeding blanket designed to produce enough tritium to sustain its own fuel cycle. Demonstrating that such tritium self-sufficiency can be achieved is one of the important goals of the ITER Test Blanket Module programme. In ITER itself, less than 1 gram of fusion fuel will be present in the vacuum vessel at any one time.

Fuelling

For efficient D–T fusion, deuterium and tritium are supplied in approximately equal proportions and introduced into the plasma. Fuelling the hot plasma core is challenging because neutral gas entering from the edge is rapidly ionised, so ordinary gas injection becomes inefficient at delivering particles deep into a burning plasma.

ITER will use two principal fuelling methods. A gas injection system will provide the initial chamber fill and help control plasma density, with a design throughput of about 200 Pa·m³/s on average and up to 400 Pa·m³/s at peak. Gas fuelling is most effective near the plasma edge and becomes less efficient for core fuelling in high-performance plasmas.

The second method uses frozen deuterium–tritium pellets several millimetres in size, accelerated by a pneumatic injector and guided into the plasma. Fuelling pellets of about 5 mm are designed to travel at speeds of up to roughly 300 m/s, allowing them to penetrate much deeper than injected gas and provide effective core fuelling. Pellet injection can also be used for controlled ELM pacing. Edge Localised Modes are instabilities at the plasma edge that cause short bursts of heat and particles towards plasma-facing components.