The 18 Toroidal Field (TF) magnets produce a magnetic field around the torus, whose primary function is to confine the plasma particles. The ITER TF coils are designed to have a total magnetic energy of 41 gigajoules and a maximum magnetic field of 11.8 tesla. The coils will weigh 6540 tons total; besides the Vacuum Vessel, they are the biggest components of the ITER machine.
The coils will be made of Cable-In-Conduit superconductors, in which a bundle of superconducting strands is cabled together and cooled by flowing Helium, and contained in a structural jacket. The strands necessary for the ITER TF coils have a total length of 80,000 kilometres.
Poloidal Field System
The Poloidal Field coil system consists of six independent coils placed outside the Toroidal Magnet structure.
The Poloidal Field (PF) magnets pinch the plasma away from the walls and contribute in this way to maintaining the plasma's shape and stability. The PF field is induced both by the Magnets and by the current drive in the plasma itself.
The Poloidal Field coil system consists of six horizontal coils placed outside the Toroidal Magnet structure. Due to their size, the actual winding of five of the six PF coils will take place in a dedicated, 250-metre-long coil winding building on the ITER site in Cadarache. The smallest of the PF coils will be manufactured offsite and delivered finished.
The ITER PF coils are also made of Cable-in-Conduit conductors. Two different types of strands are used according to operating requirements, each displaying differences in high-current and high-temperature behaviour.
The Central Solenoid - the backbone of the Magnet system - is essentially a large transformer.
The main plasma current is induced by the changing current in the Central Solenoid which is essentially a large transformer, and the 'backbone' of the Magnet System. It contributes to the inductive flux that drives the plasma, to the shaping of the field lines in the Divertor region, and to vertical stability control. The Central Solenoid is made of six independent coil packs that use a Niobium-Tin (Nb3Sn) Cable-in-Conduit superconducting conductor, held together by a vertical precompression structure. This design enables ITER to access a wide operating window of plasma parameters, enabling the testing of different operating scenarios up to 17 MA and covering inductive and non-inductive operation.
Each coil is based on a stack of multiple pancake winding units that minimizes joints. A glass-polyimide electrical insulation, impregnated with epoxy resin, gives a high voltage operating capability, tested up to 29 kV. The conductor jacket material has to resist the large electromagnetic forces arising during operation and be able to demonstrate good fatigue behaviour. The conductor will be produced in unit lengths up to 910 metres.
The large stainless steel Vacuum Vessel provides an enclosed, vacuum environment for the fusion reaction.
The Vacuum Vessel is a hermetically-sealed steel container inside the Cryostat that houses the fusion reaction and acts as a first safety containment barrier. In its doughnut-shaped chamber, or torus, the plasma particles spiral around continuously without touching the walls.
The size of the Vacuum Vessel dictates the volume of the fusion plasma; the larger the vessel, the greater the amount of power that can be produced. The ITER Vacuum Vessel will be twice as large and sixteen times as heavy as any previous tokamak, with an internal diametre of 6 metres. It will measure a little over 19 metres across by 11 metres high, and weigh in excess of 5,000 tons.
The ITER Vacuum Vessel with its 44 ports. At 8,000 tons, the stainless steel vacuum vessel weighs slightly more than the Eiffel Tower.
The Vacuum Vessel will have double steel walls, with passages for Cooling Water to circulate between them. The inner surfaces of the Vessel will be covered with Blanket Modules that will provide shielding from the high-energy neutrons produced by the fusion reactions. Some of the Blanket Modules will also be used at later stages to test materials for Tritium Breeding concepts.
Forty-four ports will provide access to the Vacuum Vessel for Remote Handling operations, Diagnostic systems, Heating, and Vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports.
Blanket modules provide shielding from the high thermal loads within the Vacuum Vessel and the high-energy neutrons produced by the fusion reactions. In later experiments some modules may be used to test Tritium Breeding concepts.
The Blanket covers the interior surfaces of the Vacuum Vessel, providing shielding to the Vessel and the superconducting Magnets from the heat and neutron fluxes of the fusion reaction. The neutrons are slowed down in the Blanket where their kinetic energy is transformed into heat energy and collected by the coolants. In a fusion power plant, this energy will be used for electrical power production.
The ITER Blanket is one of the most critical and technically challenging components in ITER: together with the Divertor it directly faces the hot plasma. Because of its unique physical properties, Beryllium has been chosen as the element to cover the first wall. The rest of the Blanket shield will be made of high-strength copper and stainless steel.
At a later stage of the ITER project, test breeding modules will be used to test materials for Tritium Breeding concepts. A future fusion power plant producing large amounts of power will be required to breed all of its own Tritium. ITER will test this essential concept of Tritium self-sustainment.
The ITER Divertor
The Divertor is one of the key components of the ITER machine. Situated along the bottom of the Vacuum Vessel, its function is to extract heat and Helium ash — both products of the fusion reaction — and other impurities from the plasma, in effect acting like a giant exhaust system. It will comprise two main parts: a supporting structure made primarily from stainless steel, and the plasma-facing components, weighing about 700 tons. The plasma-facing components will be made of Tungsten, a high-refractory material.
Located at the very bottom of the Vacuum Vessel, the ITER Divertor is made up of 54 remotely-removable cassettes, each holding three plasma-facing components, or targets. These are the inner and the outer vertical targets, and the dome. The targets are situated at the intersection of magnetic field lines where the high-energy plasma particles strike the components. Their kinetic energy is transformed into heat; the heat flux received by these components is extremely intense and requires active water cooling. The choice of the surface material for the Divertor is an important one. Only very few materials are able to withstand temperatures of up to 3,000°C for the projected 20-year lifetime of the ITER machine; these will be tested in ITER.
ITER will begin operations with a Carbon fibre-reinforced Carbon composite (CFC) Divertor target. This material presents the advantage of high thermal conductivity and it enables an easier learning process for the first years of ITER operation. A second Divertor set will be made of Tungsten which has the advantage of a lower rate of erosion and thus a longer lifetime.
About 50 individual measurement systems will help to control, evaluate and optimize plasma performance in ITER and to further understanding of plasma physics.
An extensive diagnostic system will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times.
The system will comprise about 50 individual measuring systems drawn from the full range of modern plasma diagnostic techniques, including lasers, X-rays, neutron cameras, impurity monitors, particle spectrometres, radiation bolometers, pressure and gas analysis, and optical fibres.
Because of the harsh environment inside the Vacuum Vessel, these systems will have to cope with a range of phenomena not previously encountered in diagnostic implementation, while all the while performing with great accuracy and precision. The levels of neutral particle flux, neutron flux and fluence will be respectively about 5, 10 and 10,000 times higher than the harshest experienced in today's machines. The pulse length of the fusion reaction—or the amount of time the reaction is sustained—will be about 100 times longer.
External Heating Systems
The ITER Tokamak will rely on three sources of external heating to bring the plasma to the temperature necessary for fusion.
The temperatures inside the ITER Tokamak must reach 150 million° Celsius—or ten times the temperature at the core of the Sun—in order for the gas in the vacuum chamber to reach the plasma state and for the fusion reaction to occur. The hot plasma must then be sustained at these extreme temperatures in a controlled way in order to extract energy.
The ITER Tokamak will rely on three sources of external heating that work in concert to provide the input heating power of 50 MW required to bring the plasma to the temperature necessary for fusion. These are neutral beam injection and two sources of high-frequency electromagnetic waves.
Ultimately, researchers hope to achieve a "burning plasma"—one in which the energy of the Helium nuclei produced by the fusion reaction is enough to maintain the temperature of the plasma. The external heating can then be strongly reduced or switched off altogether. A burning plasma in which at least 50 percent of the energy needed to drive the fusion reaction is generated internally is an essential step to reaching the goal of fusion power generation.