glossary (a-z) of iter terms
All A B C D E F G H I J K L M N O P Q R S T U V W

Additional Heating

The application of neutral particle beams and/or high-frequency microwave radiation to the plasma from external sources, in order to provide the input heating power necessary to reach the temperatures required for fusion. Additional heating bridges the gap between resistive (or ohmic) heating due to plasma toroidal current (which gets weaker with increased temperature) and alpha-particle heating due to the slowing down of the helium reaction product in the plasma (which gets stronger with higher temperature).

Alcator C-Mod

A tokamak experiment run by the Massachusetts Institute of Technology (MIT) Boston, USA. One ot the three major US tokamaks (along with DIII-D and NSTX). See


Heating antennae providing radiofrequency power at electron cyclotron (EC) and ion cyclotron (IC) frequencies to the plasma.

ASDEX Upgrade

The ASDEX Upgrade divertor tokamak at the Max Planck Institute for Plasma Physics (IPP) in Garching is Germany's largest fusion device. See more here.


The flight tube connecting the neutral beam box to the plasma torus. Sometimes, colloquially, the complete beam system including the beam box.


A thick wall of concrete surrounding the cryostat and designed to absorb the bulk of the remaining neutron radiation from the plasma. The wall shields the region outside the cryostat so that it can be accessed, at maximum a couple of weeks after shutdown, for major hands-on repairs.


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. In ITER, some of the 440 individual blanket modules will be used to test materials for tritium breeding concepts.

Blanket Module

In ITER, the blanket is subdivided in modules to allow it to be relatively easily replaced through equatorial access ports. There are 440 individual segments, each measuring 1x1.5 metres and weighing approximately 4.5 tons. Each segment has a detachable first wall which directly faces the plasma and removes the plasma heat load, and a semi-permanent blanket shield dedicated to neutron shielding.

Blanket Shield

The first component facing the plasma; it removes heat and protects the Vacuum Vessel and Magnets from radiation damage.


The plasma energy breakeven point describes the moment when plasmas in a fusion device release at least as much energy as is required to produce them. Fusion performance is measured by Q. Plasma energy breakeven, or Q=1, has never been achieved: the current record for energy release is held by JET, which succeeded in generating nearly 70% of input power. ITER has been designed to produce more power than it consumes: for 50 MW of input power, 500 MW of output power will be produced (Q=10).

Breeding Blanket

If blanket modules contain lithium, a reaction occurs: the incoming neutron is absorbed by the lithium atom, which recombines into an atom of tritium and an atom of helium. The tritium can be removed from the blanket and recycled into the plasma as fuel. Blankets containing lithium are thus considered to be "breeding blankets" for tritium. Within the fusion reaction, tritium can be 'bred' indefinitely.
This will be important technology for future fusion power reactors.

Breeding Technologies

A number of different combinations of tritium breeding material, neutron multiplier, structural material, and coolant will be tried out on ITER to determine the best combination for tritium and power production. Each one of these solutions is referred to as a "breeding technology". A future fusion 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.

Broader Approach

A "Broader Approach" agreement for complementary research and development was signed in February 2007 between the European Atomic Energy Community (known by its initials EURATOM) and the Japanese government. It established a framework for Japan to conduct research and development in support of ITER and the next-stage device, DEMO, over a period of ten years.


The period of roughly constant and maximum fusion power during the plasma pulse.

Burning Plasma

A plasma 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 methods can then be strongly reduced or switched off altogether. A burning plasma in which at least 50 percent of the energy to drive the fusion reaction is generated internally is an essential step to reaching the goal of fusion power generation.

Central Solenoid

Part of the magnet system in ITER, the central solenoid acts like a large transformer, producing and sustaining the plasma current which heats and shapes the plasma.


The process during which plant or reactor components and systems, after construction, are made operational and verified to be in accordance with design assumptions and performance criteria.


Restriction of a hot plasma to a given volume as long as possible by magnets and pinch effects.

Confinement Time

The time the plasma is maintained at a temperature above the critical ignition temperature. To yield more energy from fusion than has been invested to heat the plasma, the plasma must be held up to this temperature for some minimum length of time, calculated from scaling laws.


Usually refers to the cooling period necessary to remove heat from a large superconducting magnet system to lower the temperature to the operating point.

Cooling Water System

The cooling water system provides for the rejection of heat from a variety of ITER systems and consists of the tokamak cooling water system, the component cooling water system, the chilled water system, and the heat rejection system.

Correction Coils

Coils whose purpose is to compensate small errors in the confining magnetic field arising from fabricating misalignments.


Term applied to substances and materials at very low temperatures (below -150°C).


The plant used to liquify helium and nitrogen to cool magnets, vacuum pumping panels, etc.


A vacuum pump system using panels cooled by liquid helium.


A vacuum vessel built around a superconducting tokamak, capable of being evacuated at room temperature, which provides thermal insulation to maintain the magnets at low temperature.

Current Drive

A means for producing the toroidal plasma current.


The process by which the facility is permanently taken out of operation at the end of the plant lifecycle with adequate regard for the health and safety of workers and the public, and protection of the environment.


Demonstration fusion reactor. The next experimental device to follow ITER, and predecessor to a commercial-sized fusion reactor. DEMO would generate electricity at the level of a few hundred MW and utilize all technologies necessary for a commercial device.


An isotope of hydrogen. Its nucleus contains one neutron and one proton.

Deuterium-Tritium Plasma

A plasma obtained by using deuterium and tritium as the fusion fuels. Also called a DT plasma.

Deuteurium-Tritium Reactions

Reactions between the nuclei of two isotopes of hydrogen, deuterium and tritium, to release energy by nuclear fusion. Helium nuclei and neutrons are produced.


Equipment for determining/monitoring the properties and behaviour of a plasma during an experiment.


The DIII-D tokamak was developed in the 1980s by General Atomics in San Diego, USA, as part of the ongoing effort to achieve magnetically confined fusion. See this link.


The component of the ITER device that removes helium "ash" and plasma heat during operation of the tokamak. 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.

Divertor Channel

The region of the divertor into which field lines in the plasma scrape-off layer are conducted.


The Experimental Advanced Superconducting Tokamak (EAST) is an experimental superconducting tokamak magnetic fusion energy reactor in Hefei, the capital city of Anhui Province, in eastern China. See this link.

Electron Beam

A stream of electrons moving with the same velocity and direction in neighbouring paths and usually emitted from a single source such as a cathode.

Electron Cyclotron Resonance Heating

(ECRH) An external mode of heating the plasma through resonant absorption of energy by introducing electromagnetic waves into the plasma at the cyclotron frequency of electrons.


Edge Localized Mode. Regular, energetic bursts of energy and particles that escape from the magnetic field surrounding the plasma and cause loss of energy. The mitigation of this phenomenon is an important preoccupation of tokamak physics.

Energy Confinement Time

The ratio of instantaneous plasma energy content to the net power flow into the plasma required to maintain that energy content.

Equatorial Port

Mid-level entryways into the ITER vacuum vessel. The 17 equatorial ports will provide access into the vessel for remote handling, diagnostics, heating, and vacuum systems.


EUROfusion, the European Consortium for the Development of Fusion Energy, manages European fusion research activities on behalf of Euratom. See:

First Wall

The interior surface of the tokamak closest to the plasma.


The process by which a neutron strikes a nucleus and splits it into fragments; during the process of nuclear fission, neutrons are released at high speed, and heat and radiation are released.

Fuel Cycle System

The system which extracts deuterium, tritium and impurities from the plasma exhaust stream and prepares deuterium and tritium for re-injection into the plasma.

Fuel Pellets

Small slugs of frozen deuterium and tritium fuel in the 3-6 mm diameter range fired frequently (up to 20 pellets per second) into the plasma to maintain sufficient fuel density in the plasma core. Pellet injection is also efficient in controlling Edge Localized Modes, or ELMs. Special technology is being developed to allow these pellets to fly along curved trajectories, thereby attaining specific zones within the plasmas where ELMs are particularly disruptive.


The merging of two light atomic nuclei into a heavier nucleus, with a resultant loss in the combined mass and a massive release of energy.

Fusion Performance

The level of power amplification, Q, or the energy confinement time during a fusion reaction.

Fusion Triple Product

The "triple product" of density, confinement time and plasma temperature is used by researchers to measure the performance of a fusion plasma. The triple product has seen an increase of a factor of 10,000 in the last thirty years of fusion experimentation; another factor of six is needed to arrive at the level of performance required for a power plant.

Gas Puffing

By releasing puffs of fuel or impurity gas from valves into the plasma chamber it is possible to fuel the outer regions of the plasma. Fuel pellets are used to fuel deeper into the plasma.


A type of radiofrequency power transmission tube (valve) used to produce electromagnetic waves in the GHz range for plasma heating in the range of the plasma electron cyclotron resonance.


The H-mode is the baseline mode of plasma operation on all of today's major tokamaks. As the plasma auxiliary heating exceeds a certain threshold power the energy confinement of the plasma spontaneously doubles. This phenomenon was first discovered on ASDEX in 1982.

Hot Cell

A concrete-shielded chamber with a controlled atmosphere that can be used to work on radioactive materials and components with a view to repairing them and refurbishing them for future re-use, or dismantling them for disposal. The chamber is equipped with remote manipulators or robotic devices for this purpose. No human access is foreseen.


International Atomic Energy Agency, Vienna, Austria.


International Fusion Materials Irradiation Facility, Naka, Japan. Part of the Broader Approach agreement, IFMIF is an international scientific research program designed to test materials for suitability for use in a fusion reactor. Jointly developed by Europe and Japan, IFMIF will use a particle accelerator-based neutron source to produce a large neutron flux, in a suitable quantity and time period to test the long-term behavior of materials under conditions similar to those expected at the inner wall of a fusion reactor. Engineering validation and engineering design activities (EVEDA) are currently underway. See more at IFMIF/EVEDA.


The point at which a fusion reaction becomes self-sustaining. At ignition, fusion self-heating is sufficient to compensate for all energy losses, external sources of heating power are no longer necessary to sustain the reaction.


Atoms of unwanted elements in the plasma usually originating from the surrounding walls.

Induction Coil

A transformer for producing high voltage pulses in the secondary winding, obtained from interrupted direct current in the primary, as for a gas engine.

In-vessel components

The in-vessel components comprise the blanket, the divertor, the fuelling and internal pumping systems, the port plugs, the in-vessel coils and diagnostic sensors mounted directly on the vessel.



An atom which has become charged as a result of gaining or losing one or more orbiting electrons. A completely ionized atom is one that is stripped of all its electrons.

Ion Cyclotron Resonance Heating

(ICRH) An external mode of heating the plasma through resonant absorption of energy by introducing electromagnetic waves into the plasma at the cyclotron frequency of ions.


The removing or adding of an electron to a neutral atom, thereby creating an ion.


One of several versions of the same element, possessing different numbers of neutrons but the same number of protons in their nuclei.

ITER (the name)

The acronym ITER (pronounced "eater") is the Latin word for "the way." In choosing this name, the participants in the early Conceptual Design Activities for ITER (1988-1992) thus expressed their common hopes that the project would lead to international cooperation on the development of a new form of energy. (The acronym also originally stood for International Thermonuclear Experimental Reactor—a name that is no longer used.)


International Tokamak Physics Activity. ITPA aims at cooperation in development of the physics basis for burning tokamak plasma physics, covering designs and issues broader than those represented by ITER. See the ITPA page hosted on the ITER website.


The cost estimates for the construction and operation phases of the ITER Project have been quantified using an internal currency called the "ITER Unit of Account" or IUA, established in 1989. The basis of conversion from IUA to Euro has been agreed between the Members and is updated each year.

Because seven ITER Members are collaborating to build ITER, each with responsibility for the procurement of in-kind hardware in its own territory with its own currency, the IUA was devised to measure the value of in-kind contributions consistently over time, and to neutralize market fluctuations. 


The European Joint European Torus tokamak experiment hosted by the Culham Centre for Fusion Energy (CCFE, UK). See


The upgrade of the Japanese JT-60 tokamak. See theJT-60SA website.


The KSTAR, or Korea Superconducting Tokamak Advanced Research, is a magnetic fusion device at the National Fusion Research Institute in Daejon, South Korea. See


Present in minerals and salt in the Earth's crust, lithium is the lightest metal.

Lower Ports

Low-level entryways into the ITER vacuum vessel. The 9 lower ports will provide access into the vessel for remote handling, diagnostics, heating, and vacuum systems.

Magnetic Confinement

The containment of a plasma during fusion experiments by applying a specific pattern of magnetic fields. Also referred to as a magnetic bottle.

Magnetic Fusion

The use of magnetic fields to confine a plasma that is undergoing fusion.

Major Plasma Radius

The centre of the last closed flux surface at the height of the maximum width of the plasma. The major radius of the torus.


The Mega Amp Spherical Tokamak located at the Culham Centre for Fusion Energy (CCFE, UK).

NbTi Conductor

Flexible superconductor made of niobium titanium compound suitable for use up to 10T with helium coolant at 4.5 K.

Net Energy

ITER will be the first fusion device to produce net energy. This means that the total energy created during a fusion plasma pulse will surpass the amount of energy required to power the machine's systems (heating). See Breakeven.

Neutral Beam

High-energy beams of neutral atoms, typically a hydrogen isotope such as deuterium, that are injected into the core of the plasma through neutral beam injection. These energetic atoms transfer their energy to the plasma, raising the overall temperature.

Neutron Wall Loading

Energy flux carried by fusion neutrons into the first physical boundary that surrounds the plasma.


The National Spherical Torus Experiment is a spherical tokamak that has been operated since 1999 by the Princeton Plasma Physics Laboratory (PPPL), USA. See

Ohmic Heating

The heating effect resulting from the resistance a medium offers to the flow of electric current. In a plasma subjected to ohmic heating, ions are heated almost entirely by the transfer of energy from the hotter electrons. Also known as resistive heating.

Pellet Injector

A device that shoots small frozen quantities of hydrogen isotopes at high speed into the inner regions of a hot plasma. This method has some penetration advantages over conventional gas injection.


The fourth state of matter. At extreme temperatures, electrons are separated from nuclei and a gas becomes a plasma - a hot, electrically charged gas. In a star as in a fusion device, plasmas provide the environment in which light elements can fuse and yield energy. Some 99% of the known universe is in the plasma state. Examples of plasmas include the sun, fluorescent light bulbs, and other gas-discharge tubes.

Plasma Current

The electrical current going the long way around the torus.

Plasma Disruption

A rapid deposition of plasma energy resulting from the loss of plasma confinement to part of the plasma-facing structure as a result of instabilities.

Plasma Shutdown

The orderly process of extinguishing the plasma at the end of a plasma burn pulse, involving reduction of the plasma thermal energy, and reduction of plasma current to zero.

Plasma Temperature

Temperature expressed in Kelvin (thermodynamic temperature) or electron volts (kinetic temperature). A measure of the random kinetic energy (energy of motion) of the ions or electrons present.

Plasma-Facing Components

Tokamak components which directly interact with the plasma, and are subject to high heat fluxes. In ITER, this includes the first wall and the divertor.

Poloidal Direction

Movement in the vertical plane intersecting the plasma torus along projections in that plane of any of the tokamak's nested toroidal flux surfaces.

Poloidal Field

The magnetic field generated by an electric current flowing in a ring. In toroidal devices, the magnetic field that encircles the plasma axis. (i.e. loops around the torus the short way.)

Poloidal Field Coils

Components of a tokamak that assist in stabilizing the plasma. In ITER, the poloidal field coil system consists of six horizontal coils placed outside the toroidal magnet structure.

Procurement Arrangement

Procurement Arrangements are a unique ITER invention. Each one of these documents governs the procurement of plant systems, components, or site construction and details all the necessary technical specifications and management requirements. The value of each Procurement Arrangement is expressed in ITER Units of Account (IUAs). About 140 individual Procurement Arrangements are currently planned to implement the work packages for building ITER.


Plasma power amplification; the ratio of fusion power input to the plasma divided by external power supplied to the plasma. In ITER, the programmatic goal - Q≥10 - signifies delivering ten times more power than that which is consumed by operation.


A quench is an abnormal termination of magnet operation that occurs when part of the superconducting coil loses its superconductive state, and reenters the normal, resistive state. Resistance results in ohmic heating in a specific area; this heat then rapidly causes other areas of the magnet to quench. ITER will be equipped with quench detection systems, and rapid discharge units to dissipate the excess magnet energy during a quench.

Ramp-Up Time

The time necessary to initiate the plasma and heat it up to burn temperature, involving a phase of increase (ramp-up) of plasma current, followed by an increase (ramp-up) of plasma temperature.

Remote Handling

Handling of tools or components by machines with the controls at a remote location.

Remote Maintenance

Maintenance and modification of the radioactive elements and components of the tokamak using machines and tools controlled remotely to avoid human exposure to radioactivity.

Resistive Heating

See Ohmic Heating.

Shut Down Time

The time span between the end of burn and the end of the plasma state; part of the operating time.


Symbol Sv. An international unit for biological radiation dosage measurement. A chest X-ray, for example, delivers a dose of 150µSv.

Steady State Operation

In ITER, the operation of the plasma in a way in which termination of the pulse is not determined by plasma behaviour, but is rather a choice of the operator. Operation that, in principle, can continue indefinitely.

Steady State Tokamak

A tokamak in which conditions such as temperature, reaction rate, and neutron flux do not change appreciably with time.


A device invented by Lyman Spitzer (USA) for the containment of a plasma inside a racetrack-shaped tube. The toroidal device produces a poloidal field in a plasma with the use of external magnetic field coils.

Superconducting Coil

Magnetic coils which use superconductors that have zero resistivity when cooled below the critical temperature.

Superconducting Magnets

Superconducting magnet coils which circumscribe the torus to confine the plasma within and away from its inner surface.


The flow of electric current without resistance in certain metals and alloys at temperatures near absolute zero.


A type of electrical conductor that permits a current to flow with zero resistance.

Supercritical Helium

Helium will remain liquid in a bath at 1 atmosphere pressure provided the temperature does not rise above 4.2K. If the ITER coils are placed in such a coolant bath and a high pulse of heat ensues in their operation, most of the helium must be vented to avoid large overpressures. To avoid this, the coils of ITER operate with pumped supercritical helium, just above the critical temperature, which retains a large measure of the heat transfer properties of liquid helium without the risk of overpressure.


Tokamak-15, an experiment at the Kurchatov Institute, Moscow. It was the first tokamak to use superconducting magnets to control the plasma.


The Tokamak Fusion Test Reactor (TFTR) was an experimental tokamak built at Princeton Plasma Physics Laboratory (New Jersey, USA), and operated from 1982-1997. See more information here.

Thermal Shield

The tokamak component that absorbs neutrons and transfers heat, protecting the vacuum vessel and the magnets.


A fusion device for containing a plasma inside a torus chamber through the use of two magnetic fields--one created by electric coils around the torus, the other created by intense electric current in the plasma itself. The tokamak was invented in the 1950s by Soviet physicists Igor Yevgenyevich Tamm and Andrei Sakharov. The term tokamak is a transliteration of a Russian expression (toroidalnaya kamera + magnitnaya katushka) meaning toroidal chamber with magnetic coils.

Tore Supra

A superconducting fusion experiment at the Institute for Magnetic Fusion Research, IRFM (CEA Cadarache research centre) in France, which aims particularly at demonstrating long-pulse tokamak operation. Currently, Tore Supra is being upgraded with an actively cooled tungsten divertor (the WEST project) to serve as a test bed for ITER.

Toroidal Current

The basic means of driving toroidal current in the tokamak plasma uses the fact that most field lines created by the central solenoid pass down its bore and do not return on themselves until they pass outboard of (i.e., radially beyond) the plasma. This "inductive linkage" between the solenoid and plasma allows a change in current in the solenoid to drive current in the plasma (Maxwell's Laws).

Toroidal Direction

In a doughnut-shaped torus, the direction parallel to the large circumference.

Toroidal Field

The magnetic field generated by an electrical currrent flowing around a torus.

Toroidal Field Coils

Components of a tokamak that assist in stabilizing the plasma, by creating a "magnetic bottle" for confinement. In ITER, the toroidal field coil system consists of 18 D-shaped vertical coils placed around the vacuum vessel.


A surface of revolution generated by revolving a circle in three-dimensional space about an axis coplanar with and not touching the circle. Examples of tori include the surfaces of doughnuts and inner tubes. The solid contained by the surface is known as a toroid.


The third isotope of hydrogen, containing one proton and two neutrons in the nucleus.

Tritium Handling

The processes of tritium removal from gas streams, including plasma exhaust and the plant atmosphere, and returning it to use as fuel in the plasma or to storage, in a safe manner.

Upper Ports

Upper-level entryways into the ITER vacuum vessel. The 18 upper ports will provide access into the vessel for remote handling, diagnostics, heating, and vacuum systems.

Vacuum Pumps

Pumps which remove gas, usually air, from a chamber, leaving it under vacuum. A variety of types are used in ITER, depending on the quality and degree of the vacuum needed.

Vacuum Vessel

The Vacuum Vessel establishes the confinement barrier in a tokamak and limits the heat flux to the toroidal field coils. It provides low-impurity volume for the reacting plasma.


The upgrade of the Tore Supra tokamak in France to a test bed for ITER, with supplementary magnetic coils and a new ITER-like tungsten divertor. See the West website here.