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FR

Fusion glossary

A

ADITYA (synonym of Sun in Hindi) is the first indigenously designed and fabricated tokamak in India. Located at the Institute for Plasma Research in Gujarat and operated since 1989, this medium-size tokamak conducts experiments with high plasma current at high temperature. It was upgraded in 2016 to ADITYA-U to realize shaped-plasma operations in an open diverter configurations. See this website.

A tokamak experiment run by the Massachusetts Institute of Technology (MIT) in Boston, USA up to the end of funding in 2016. One of the three major US tokamaks (along with DIII-D and NSTX). See MIT's Plasma Science and Fusion Center.

Oscillations caused by interacting magnetic fields and electric currents in plasma.
The fusion between the nuclei of the hydrogen isotopes deuterium (D) and tritium (T) produces one helium nucleus, also called an "alpha particle," and one neutron. The helium nucleus, which carries 20% of the energy produced by the fusion reaction, is electrically charged and remains confined by the magnetic fields of the tokamak. The heating provided by these alpha particles contributes to maintaining the temperature of the plasma. When heating by the helium nuclei is dominant ("alpha heating") the plasma is said to be a "burning plasma."

In fusion, created by fusing deuterium and tritium nuclei. The particle is the nucleus of a helium atom, made of two protons and two neutrons bound together.

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

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.  

In fusion, this is the ratio of the major to minor radius of the toroidal plasma.

The number of protons in an atom’s nucleus. Uniquely characterizes every element.

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. Auxiliary 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).

B

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

Light metal with atomic number 4. ITER originally selected beryllium as the plasma-facing material in the first wall of the blanket; however, the decision was made in 2023 to replace beryllium with tungsten—a material that is more relevant for future “DEMO” machines and commercial fusion devices.

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, composed of the shield block module and the first wall, 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.

An intrinsic toroidal current created in the presence of a pressure gradient. Arises from collisions between trapped and passing particles and was experimentally verified at JET and TFTR. Valuable for the design of advanced tokamak experiments.

Application of a thin layer of boron over the plasma-facing components to reduce impurity radiation. The layer will capture–or “getter”–oxygen that could otherwise increase radiative losses of the plasma.

The point where the amount of energy produced by fusion equals the energy required to fuel the reaction (Q = 1).
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 an important technology for future fusion power reactors.
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.

An agreement for complementary research and development between the European Atomic Energy Community (Euratom) and the Japanese government. Signed in 2007 and renewed in 2020, the Broader Approach establishes a framework for advanced research and development in support of ITER and the next-stage device, DEMO.

The period of roughly constant and maximum fusion power during the plasma pulse.
A plasma in which the energy of the helium nuclei (alpha particles) 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. At Q = 10 (ITER), approximately 66% of the plasma heating is contributed by the alpha particles.

C

The French Alternative Energies and Atomic Energy Commission, a multidisciplinary research organization. CEA Cadarache, next to ITER, is home to the Institute for Magnetic Fusion Research IRFM and the WEST tokamak.

Part of the magnet system in ITER, the central solenoid acts like a large transformer, allowing a powerful current to be induced in the ITER plasma and maintained during long plasma pulses.

Control system that provides the interface “language” for over 30 ITER plant systems.

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 for as long as possible through magnets and pinch effects.
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.
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.
Coils whose purpose is to compensate small errors in the confining magnetic field arising from fabrication misalignments.
Term applied to substances and materials at very low temperatures (below -150 °C).
The ITER plant used to liquify helium and nitrogen to cool magnets, thermal shields, vacuum pumping panels, etc.
A vacuum pump system using panels cooled by liquid helium to trap condensable gases from the plasma.
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.
A means for producing the toroidal plasma current.

D

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.  

DEMO (DEMOnstration fusion reactor) is a generic term referring to the next class of experimental device to follow ITER, predecessor to a demonstration power plant. DEMO would generate electricity at the level of a few hundred MW and utilize all technologies necessary for a commercial device. See the After ITER page of the website for more information about the projects underway in the ITER Members.

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

A plasma obtained by using deuterium and tritium as the fusion fuels. Also called a D-T plasma.
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. It is currently operated by General Atomics for the US Department of Energy. See this link.

A disruption is an instability that may develop within the tokamak plasma. Disruptions lead to the degradation or loss of the magnetic confinement of the plasma and, because of the high amount of energy contained within the plasma, the loss of confinement during a disruption can cause a significant thermal loading of in-vessel components together with high mechanical strains on the in-vessel components, the vacuum vessel and the magnet coils. Disruptions have been observed, avoided and mitigated in most operating tokamaks. 
A system to protect machine components against excessive heat loads and electromagnetic forces resulting from plasma disruptions, in which there is a sudden loss of stored thermal and magnetic energy. See shattered pellet injection.
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.
The region of the divertor into which field lines in the plasma scrape-off layer are conducted.
Agencies created in each of the seven ITER Members to fulfil in-kind procurement responsibilities to ITER. In-kind procurement means that instead of supplying only cash to the ITER Organization for the construction phase of the project, Members are supplying 90% of their participation "in kind," in the form of systems, components and, in the case of Europe, buildings.
The building block of ITER's toroidal field and poloidal field superconducting magnets. Double pancakes are double layers of spiralled superconductor wound into the precise shape of the coil (D-shape for the toroidal field magnets, ring-shape for the poloidal field magnets). Six to nine double pancakes are stacked to form the final magnet assemblies.

E

The Experimental Advanced Superconducting Tokamak (EAST) is a superconducting tokamak in operation at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), in Hefei since 2006. See this link.

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.
(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 physicists.
The condition of Q = 1, meaning that the total fusion power produced during a plasma pulse is equal to the power injected into the systems that heat the plasma.
The ratio of instantaneous plasma energy content to the net power flow into the plasma required to maintain that energy content.

The Swiss Federal Institute of Technology in Lausanne, home to the Swiss Plasma Center and the TCV tokamak.

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 this site. 

F

The interior surface of the tokamak, closest to the plasma. Detachable, front-facing elements of the ITER blanket (called first wall panels) will be installed to withstand the heat flux from the plasma.
The system which extracts deuterium, tritium and impurities from the plasma exhaust stream and prepares deuterium and tritium for re-injection into the plasma.
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.
The level of power amplification, Q, or the energy confinement time during a fusion reaction.

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 fifty years of fusion experimentation; less than an additional factor of ten is needed to arrive at the level of performance required for a fusion power plant. See also: Lawson's Criteria

G

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.

H

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.

Ash is the name given to the helium nuclei produced by fusion reactions in a deuterium-tritium plasma, even though helium bears no physical resemblance to ash from a fire. The helium nuclei are termed "ash" because they have no further use once they have shared their energy with the rest of the plasma; their removal and replacement by deuterium-tritium fuel is required to prevent dilution of the plasma.

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.

The lightest element with atomic number 1. Has three isotopes: protium, deuterium, and tritium. Protium is the most common with a >99% abundance; ITER will use deuterium and tritium as fuel for its fusion reactions.

I

International Atomic Energy Agency, Vienna, Austria. See this website.

The International Fusion Materials Irradiation Facility/DEMO Oriented NEutron Source (IFMIF-DONES) will be a research infrastructure for the testing, validation and qualification of the materials to be used in future fusion power plants like DEMO (a demonstration fusion reactor prototype). It is under construction now in Granada, Spain, with Spain, Croatia, and Japan as principal project leads. See this website.

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 completely 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.
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.  

Use of lasers to compress a fuel pellet, raising it to temperatures sufficient for nuclear fusion.

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.
(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.

Max Planck Institute for Plasma Physics (IPP) in Germany, home to the ASDEX Upgrade tokamak (Garching) and the Wendelstein 7-X stellarator (Greifswald). See this site.

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

The International Tokamak Physics and Engineering Activity (ITPEA) provides a framework for internationally coordinated fusion research activities. The ITPEA continues the tokamak physics and engineering R&D activities that have been conducted on an international level for many years resulting in the achievement of a broad engineering and physics basis essential for the ITER design and useful for all fusion programs and for progress toward fusion energy generally. (The name was changed from the ITPA in late 2025.)

The ITPEA operates under the auspices of ITER. See the ITPEA 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.

J

The Joint European Torus, JET, operated between 1983 and 2023 as a joint European project at the Culham Centre for Fusion Energy, UK. It completed its 40-year operational lifetime in autumn 2023 and is now slated for repurposing and decommissioning (2024-2040). Read more about the JET device and its many milestones here. Read more about decommissioning and repurposing here.

The upgrade of the Japanese JT-60 tokamak. On achieving first plasma in October 2023, JT-60SA became the world's largest functioning tokamak. See the JT-60SA website.

K

KSTAR (Korea Superconducting Tokamak Advanced Research) is a magnetic fusion device at the Korea Institute of Fusion Energy (KFE) in Daejeon, South Korea. KSTAR achieved its first plasma in 2008. See this page.

L

In 1955, British physicist John Lawson (1923-2008) demonstrated that the conditions for fusion rely on three vital parameters: temperature (T), density (n) and confinement time (τ). (Lawson's Criteria) 

Present in minerals and salt in the Earth's crust, lithium is the lightest metal, with atomic number 3. ITER will test specific lithium-containing wall modules to "breed" tritium (see Tritium breeding). In effect, tritium can be produced within the tokamak when neutrons escaping the plasma interact with lithium contained in the blanket.

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.

M

The containment of a plasma during fusion experiments by applying a specific pattern of magnetic fields. Also referred to as a magnetic bottle.
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 (UK). See this webpage.

N

At the Neutral Beam Test Facility at Consorzio RFX, in Padua, Italy, ITER neutral beam injection will be tested in advance of operation on two test beds: SPIDER (an ITER-scale negative ion source) and MITICA (a full-size ITER neutral beam injector). See this ITER webpage.

Flexible superconductor made of niobium titanium compound suitable for use up to 10T with helium coolant at 4.5 K.
When the total power produced during a fusion plasma pulse surpasses the thermal power injected to heat the plasma. See Energy Breakeven.
ITER's heating neutral beam injectors shoot uncharged high-energy particles into the plasma where, by way of chaotic motion and collision, they will transfer their energy to the charged plasma particles and raise plasma temperature.    

NSTX-U is an upgrade of the National Spherical Torus Experiment (NSTX, 1999-2010) located at Princeton Plasma Physics Laboratory (USA). Currently completing repairs, NSTX-U is expected to resume operations in 2025. See this page.

O

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.

P

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.    
The electrical current going the long way around the torus.
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.
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.
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.  
Movement in the vertical plane intersecting the plasma torus along projections in that plane of any of the tokamak's nested toroidal flux surfaces.
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.)  
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 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.

Q

Also called "Fusion Gain." The ratio between the power produced by the fusion reactions and the external power required to sustain them via plasma heating. In ITER, the programmatic goal, Q≥10, signifies delivering ten times more power than that which is consumed by the heating systems. Breakeven corresponds to Q=1; ignition corresponds to Q=infinity. A burning plasma has a Q value of >1.
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.  

R

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.
Handling of tools or components by machines with the controls at a remote location.
Maintenance and modification of the radioactive elements and components of the tokamak using machines and tools controlled remotely to avoid human exposure to radioactivity.
See Ohmic Heating.
Fast, concentrated beams of electrons (runaway electrons) may be produced during disruptions in ITER, and could cause damage to plasma-facing components.

S

The scrape-off layer (or SOL) is the plasma periphery—the critical buffer region between the hot core and the solid wall elements.
Defines the plasma edge and acts as a boundary between open and closed magnetic field lines.
The technique selected for disruption mitigation on ITER. A shattered pellet injector pre-empts plasma disruptions by releasing a spray of frozen deuterium-neon pellets into the plasma. The frozen pellet fragments, injected at speeds up to 250 metres per second, rapidly decrease the plasma temperature, thereby dissipating energy and minimizing potential damage to plasma-facing surfaces during a disruption.
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 1 x 1.5 metres and weighing approximately 4.5 tonnes. 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. (see also "First Wall").
The time span between the end of burn and the end of the plasma state; part of the operating time.

The Indian Steady State Superconducting Tokamak (SST-1) was fully commissioned in 2013 at the Institute for Plasma Research in Gujarat, and upgraded in 2019. SST-1 is a medium-sized tokamak producing repeatable plasma discharges up to ~ 500 ms with plasma currents in excess of 75000 A at a central field of 1.5 T. See this website.

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.
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.
Magnetic coils which use superconductors that have zero resistivity when cooled below the critical temperature.
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.
Helium will remain liquid in a bath at 1 atmosphere pressure provided the temperature does not rise above 4.2 K. 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.

T

Tokamak in operation at the Kurchatov Institute, Moscow. T-15MD is an upgrade from the historic T-15 machine—the first tokamak to use superconducting magnets to control the plasma. 

A tokamak operated by the Swiss Federal Institute of Technology in Lausanne (EPFL) as part of the Swiss Plasma Center. See this webpage.

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.

The tokamak component that prevents heat transfer to the ultra-cold superconducting 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.

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. Tore Supra has been upgraded with an actively cooled tungsten divertor (the WEST project) to serve as a test bed for ITER.

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).
In a doughnut-shaped torus, the direction parallel to the large circumference.
The magnetic field generated by an electrical currrent flowing around a torus.
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. Is unstable and decays through beta radiation with a half-life of 12.3 years. Its low natural abundance is why future fusion power plants would need to “breed” their own tritium (see Tritium breeding).

Production of tritium by way of a reaction between a high-speed neutron produced in a fusion reaction and the light metal lithium. ITER will be the first fusion device to test specific lithium-containing wall modules to "breed" tritium.

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.

Metal with atomic number 74 and a high melting point of 3687 K. ITER will use tungsten as the material for both of its plasma-facing components—the blanket and divertor.

U

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.

V

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.
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.

W

The process of cleaning the first wall of a vacuum vessel to remove contaminants.
Stresses on the plasma-facing surfaces of the tokamak from plasma particles, neutrons and electromagnetic radiation, which should be minimized to prevent damage.

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.