Tokamak
Magnets
The TF magnet consists of 18 superconducting D-shaped coils containing circular cross-section conductor, composed of Nb3Sn strands, embedded in grooved radial plates. The central solenoid (CS) uses square cross-section Nb3Sn conductor and has six modules which can be powered separately. The six poloidal field (PF) coils are made using NbTi conductor in double pancakes. The lower PF coils are designed with redundant turns and a margin in current to avoid the need to replace the coils in case of local damage in one of the coil pancakes. To accommodate field errors due to manufacturing inaccuracies or to misalignments during assembly of the magnet coils, as well as to control resistive wall mode plasma instabilities, superconducting saddle-shaped correction coils are placed around the machine outside the TF magnets.
When energised, the TF coils press together along their straight sections, forming a vault. The coils are encased to aid their support and to transfer loads across keys between the cases. The poloidal field crossing the TF coils creates overturning moments and circumferential torques on each TF coil. A shell-like structure between the coils, to the extent port penetrations allow, permits these forces to be reacted within the magnet structure, and provides a strong support for the poloidal field coils.
Vessel, Blanket and Divertor
The reaction chamber consists of a vacuum vessel supporting remotely exchangeable modular in-vessel components. The vacuum vessel consists of 9 toroidal sectors, welded in situ. The vessel is a double-walled stainless steel welded ribbed shell, with internal shield plates and ferromagnetic inserts inthe plane of the TF coils on the outboard to reduce toroidal field ripple.
The 440 blanket modules have a single-curvature faceted separate first wall attached to a shielding block which is remotely attached to the vessel through 3 cm diameter access holes in the first wall. To accommodate differential thermal expansion and electromagnetic loads, these attachments are stiff radially, but flexible transversely. The plasma-facing components are beryllium armour attached to a copper substrate, mounted on a water-cooled stainless steel support. The outboard modules may later be replaced with tritium-breeding modules.
The 54-cassette single null divertor has carbon targets and tungsten high heat flux components, again mounted on a copper substrate, and water-cooled stainless steel structure bolted to rails on the vessel floor. The targets can accommodate heat loads of more than 20 MW/m2 for 20 s, but the more normal peak heat load will be 5 - 10 MW/m2.
Ports
Six of the 17 accessible vessel equatorial port plugs are used for heating antennae and neutral beam ducts, three are used for power reactor test blankets, two for plasma limiters, and the remainder for plasma diagnostics. The limiter and two diagnostic ports are also used for remote blanket module replacement. The 9 divertor ports accommodate eight torus cryopumps, diagnostics, glow-discharge cleaning system, pellet and gas injection, and an in-vessel viewing system. Three divertor ports are also used for the remote replacement of the divertor cassettes, which are inserted radially and then slid toroidally and clamped to vessel rails. The 18 upper ports are mainly used for diagnostics. Three contain electron cyclotron antennas to control plasma instabilities (neo-classical tearing modes).
Cryostat
A cryostat surrounds the coils. It is essentially a reinforced single-shell cylinder 24 m high and 28 m diameter. Shielding thicknesses are arranged to permit personnel access at the port terminations or, exceptionally, for repairs in the cryostat-coil interspace, after shutdown. To reduce heat inleak to the coils from radiation from surrounding warm surfaces, thermal shields are used between the vacuum vessel and the toroidal field coils.
Cooling
The tokamak is water-cooled by separate circuits feeding the blanket (3 circuits in parallel), divertor and limiter (1 circuit), and vacuum vessel (2 circuits in parallel). The vessel cooling circuit alone can remove, by natural convection, all decay heat after shutdown in all vessel and in-vessel components. Typical water inlet temperature is 100°C, and pressures are in the range of 3-4.2 MPa. Baking of in-vessel components to remove adsorbed impurities is carried out at 240°C (200°C for the vessel).
Heating
The plasma is heated (and current may be driven) by a combination of electron cyclotron, ion cyclotron, lower hybrid and 1 MeV negative-ion-accelerated neutral beam systems. The initial setup will involve two neutral beams and electron and ion cyclotron systems, but the radio-frequency systems are designed in exchangeable modular units (20 MW/port) to allow various mixes to be tried, and three neutral beams can be accommodated on the machine. A heating power in excess of 110 MW is thus attainable.
Plasma Diagnostics
There are over 40 diagnostic system types with sensors located in ports and in most other available spaces where plasma behaviour can be measured. These systems fall into seven main types: magnetic, neutron, optical/infra-red, bolometric, spectroscopic, microwave and plasma-facing component. The also are split into three categories: required for basic machine protection/control, required for advanced plasma control, and required for evaluation/physics studies.
Assembly
ITER is assembled inside a cylindrical "pit" embedded up to the equatorial port level. After installation of the lower cryostat, PF coils and supports, 40° sectors of the vacuum vessel are combined with two TF coils and appropriate thermal shielding, and welded to adjacent sectors in the pit. The upper coils, ports and services are connected, and the cryostat is closed by a flat lid with heavy segmented shielding.
Remote Handling
Components in the vessel during DT operation will become radioactive. If they need replacement this must be done remotely, to casks which then remove the component to the hot cells. There they can be refurbished or disposed of and meanwhile a replacement component can be remotely installed in the tokamak. A vehicle mounted on a rail deployed in the plasma chamber after operation is used to remove and replace blanket modules. The divertor has a cantilevered transporter which brings the divertor cassette into the chamber radially. It is then shunted toroidally into position and clamped. Port plugs are removed by a cantilevered transporter into the cask.
Plant Systems
Cryoplant
The cryoplant consists of a liquid nitrogen plant, liquid helium plant, distribution system, and storage system. The main cryogenic load is the magnets, which are cooled by liquid helium. The torus cryopumpos also create a significant demand. The storage system must be able to take the liquid helium load during a coil quench, and a gas balloon is provided to accommodate any backlog in liquefaction. Due to the activation that woudl occur in high radiation regions, an 80K gaseous helium circuit is also included, particularly for the thermal shields. The system is modular, allowing expansion during later operating scenarios.
Power Supplies
The total steady power load on the ITER site is up to about 120 MW continuous power during periods of repetitive plasma operation. In addition the ITER plant demands pulsed loads to establish the plasma current and to heat the plasma. These can reach up to an additional 210 MW during the burn pulse.
Heat Rejection System
The main cooling systems of the tokamak, plus all other significant heat sources on site, reject heat to the atmosphere via a heat rejection system, currently conceived as standard cooling towers.
Fuel Cycle
The fuel cycle extracts plasma exhaust consisting of deuterium and tritium, helium from fusion reactions, and impurities, from the vacuum pumping ports, and separates it into its constituents. The deuterium and tritium are reinjected to the plasma, either as frozen pellets from the inboard plasma side, to achieve good penetration into the plasma core, or by gas injection at the top of the plasma region. The impurities are removed as waste, and the helium can be used elsewhere if sufficiently pure. The "tritium plant" part of the fuel cycle also takes care of any tritiated waste streams from contaminated atmospheres coming via the heating and ventilation systems of the plant, or from waste handling, as well as from coolant contaminated with tritium.
Control
The control system consists of a supervisory control system, superior to individual plant system control systems. In parallel there is an independent interlock system, to guard against mistaken access and actions. The supervisory system contains the operational control system, data management system, and supervisory synchronisation system. The operational control system consists of the plasma control system, forming part of a discharge control system.