This article is about the fusion reaction device. For other uses, see Tokamak (disambiguation).

A tokamak (Russian: токама́к) is a device that uses a powerful magnetic field to confine plasma in the shape of a torus. Achieving a stable plasma equilibrium requires magnetic field lines that move around the torus in a helical shape. Such a helical field can be generated by adding a toroidal field (traveling around the torus in circles) and a poloidal field (traveling in circles orthogonal to the toroidal field). In a tokamak, the toroidal field is produced by electromagnets that surround the torus, and the poloidal field is the result of a toroidal electric current that flows inside the plasma. This current is induced inside the plasma with a second set of electromagnets.

The tokamak is one of several types of magnetic confinement devices, and is one of the most-researched candidates for producing controlled thermonuclear fusion power. Magnetic fields are used for confinement since no solid material could withstand the extremely high temperature of the plasma. An alternative to the tokamak is the stellarator. The world's largest tokamak project is the ITER (International Thermonuclear Experimental Reactor) being constructed in Saint-Paul-lès-Durance, south of France. Scheduled to begin operation in 2020, it is expected to produce an output power of 500 megawatts.

Tokamaks were invented in the 1950s by Soviet physicists Igor Tamm and Andrei Sakharov, inspired by an original idea of Oleg Lavrentiev.[1]


The word tokamak is a transliteration of the Russian word токамак, an acronym of either:



A USSR stamp, 1987. Tokamak thermonuclear system.

Although nuclear fusion research began soon after World War II, the programs in various countries were each initially classified as secret. It was not until after the 1955 United Nations International Conference on the Peaceful Uses of Atomic Energy in Geneva that programs were declassified and international scientific collaboration could take place.

Experimental research of tokamak systems started in 1956 in Kurchatov Institute, Moscow by a group of Soviet scientists led by Lev Artsimovich. The group constructed the first tokamaks, the most successful being T-3 and its larger version T-4. T-4 was tested in 1968 in Novosibirsk, conducting the first ever quasistationary thermonuclear fusion reaction.[3]

In 1968, at the third IAEA International Conference on Plasma Physics and Controlled Nuclear Fusion Research at Novosibirsk, Soviet scientists announced that they had achieved electron temperatures of over 1000 eV in a tokamak device. British and American scientists met this news with skepticism since they were far from reaching that benchmark; they remained suspicious until laser scattering tests confirmed the findings the next year.[4]

In 1973 design work on JET, the Joint European Torus, began.

In 1978, Bob Guccione, publisher of Penthouse Magazine met Robert Bussard and became the world's biggest and most committed private investor in fusion technology, ultimately putting $20 Million ($60 Million in 2016 dollars) of his own money into Bussard's Compact Tokamak.[5]

Toroidal design

Tokamak magnetic field and current. Shown is the toroidal field and the coils (blue) that produce it, the plasma current (red) and the poloidal field produced by it, and the resulting twisted field when these are overlaid.

Positively and negatively charged ions and negatively charged electrons in a fusion plasma are at very high temperatures, and have correspondingly large velocities. In order to maintain the fusion process, particles from the hot plasma must be confined in the central region, or the plasma will rapidly cool. Magnetic confinement fusion devices exploit the fact that charged particles in a magnetic field experience a Lorentz force and follow helical paths along the field lines.

Early fusion research devices were variants on the Z-pinch and used electric current to generate a poloidal magnetic field to contain the plasma along a linear axis between two points. Researchers discovered that a simple toroidal field, in which the magnetic field lines run in circles around an axis of symmetry, confines a plasma hardly better than no field at all. This can be understood by looking at the orbits of individual particles. The particles not only spiral around the field lines, they also drift across the field. Since a toroidal field is curved and decreases in strength moving away from the axis of rotation, the ions and the electrons move parallel to the axis, but in opposite directions. The charge separation leads to an electric field and an additional drift, in this case outward (away from the axis of rotation) for both ions and electrons. Alternatively, the plasma can be viewed as a torus of fluid with a magnetic field frozen in. The plasma pressure results in a force that tends to expand the torus. The magnetic field outside the plasma cannot prevent this expansion. The plasma simply slips between the field lines.

For a toroidal plasma to be effectively confined by a magnetic field, there must be a twist to the field lines. There are then no longer flux tubes that simply encircle the axis, but, if there is sufficient symmetry in the twist, flux surfaces. Some of the plasma in a flux surface will be on the outside (larger major radius, or "low-field side") of the torus and will drift to other flux surfaces farther from the circular axis of the torus. Other portions of the plasma in the flux surface will be on the inside (smaller major radius, or "high-field side"). Since some of the outward drift is compensated by an inward drift on the same flux surface, there is a macroscopic equilibrium with much improved confinement. Another way to look at the effect of twisting the field lines is that the electric field between the top and the bottom of the torus, which tends to cause the outward drift, is shorted out because there are now field lines connecting the top to the bottom.

When the problem is considered even more closely, the need for a vertical (parallel to the axis of rotation) component of the magnetic field arises. The Lorentz force of the toroidal plasma current in the vertical field provides the inward force that holds the plasma torus in equilibrium.

Advanced tokamaks

Since about 1990 tokamaks are designed to operate in high-confinement mode to reduce plasma and energy losses.

Advanced or 2nd generation tokamaks generally use a 'C' or 'D' shaped plasma cross-section.

Plasma disruptions

At the necessarily large toroidal currents (15 megaamperes in ITER) the tokamak concept suffers from a fundamental problem of stability. The nonlinear evolution of magnetohydrodynamical instabilities leads to a dramatic quench of the plasma current within milliseconds. Very energetic electrons are created (runaway electrons) and finally a global loss of confinement happens. At that point very intense radiation is inflicted on small areas. This phenomenon is called a major disruption.[6] The occurrence of major disruptions in running tokamaks has always been rather high, of the order of a few percent of the total numbers of the shots. In currently operated tokamaks, the damage is often large but rarely dramatic. In the ITER tokamak, it is expected that the occurrence of a limited number of major disruptions will definitively damage the chamber with no possibility to restore the device.[7][8][9]

A large amplitude of the central current density can also result in internal disruptions, or sawteeth, which do not generally result in termination of the discharge.[10]

Plasma heating

In an operating fusion reactor, part of the energy generated will serve to maintain the plasma temperature as fresh deuterium and tritium are introduced. However, in the startup of a reactor, either initially or after a temporary shutdown, the plasma will have to be heated to its operating temperature of greater than 10 keV (over 100 million degrees Celsius). In current tokamak (and other) magnetic fusion experiments, insufficient fusion energy is produced to maintain the plasma temperature.

Ohmic heating ~ inductive mode

Since the plasma is an electrical conductor, it is possible to heat the plasma by inducing a current through it; in fact, the induced current that heats the plasma usually provides most of the poloidal field. The current is induced by slowly increasing the current through an electromagnetic winding linked with the plasma torus: the plasma can be viewed as the secondary winding of a transformer. This is inherently a pulsed process because there is a limit to the current through the primary (there are also other limitations on long pulses). Tokamaks must therefore either operate for short periods or rely on other means of heating and current drive. The heating caused by the induced current is called ohmic (or resistive) heating; it is the same kind of heating that occurs in an electric light bulb or in an electric heater. The heat generated depends on the resistance of the plasma and the amount of electric current running through it. But as the temperature of heated plasma rises, the resistance decreases and ohmic heating becomes less effective. It appears that the maximum plasma temperature attainable by ohmic heating in a tokamak is 20-30 million degrees Celsius. To obtain still higher temperatures, additional heating methods must be used.

Neutral-beam injection

Neutral-beam injection involves the introduction of high energy (rapidly moving) atoms (molecules) into an ohmically heated, magnetically confined plasma within the tokamak. The high energy atoms (molecules) originate as ions in an arc chamber before being extracted through a high voltage grid set. The term "ion source" is used to generally mean the assembly consisting of a set of electron emitting filaments, an arc chamber volume, and a set of extraction grids. The extracted ions travel through a neutralizer section of the beamline where they gain enough electrons to become neutral atoms (molecules) but retain the high velocity imparted to them from the ion source. Once the neutral beam enters the tokamak, interactions with the main plasma ions occur which significantly heat the bulk plasma and bring it closer to fusion-relevant temperatures. Ion source extraction voltages are typically of the order 50-100 kV, and high voltage, negative ion sources (-1 MV) are being developed for ITER. The ITER Neutral Beam Test Facility in Padova will be the first ITER facility to start operation.[11] While neutral beam injection is used primarily for plasma heating, it can also be used as a diagnostic tool and in feedback control by making a pulsed beam consisting of a string of brief 2-10 ms beam blips. Deuterium is a primary fuel for neutral beam heating systems and hydrogen and helium are sometimes used for selected experiments.

Magnetic compression

A gas can be heated by sudden compression. In the same way, the temperature of a plasma is increased if it is compressed rapidly by increasing the confining magnetic field. In a tokamak system this compression is achieved simply by moving the plasma into a region of higher magnetic field (i.e., radially inward). Since plasma compression brings the ions closer together.

Set of hyperfrequency tubes (84 GHz and 118 GHz) for plasma heating by electron cyclotron waves on the Tokamak à Configuration Variable (TCV). Courtesy of CRPP-EPFL, Association Suisse-Euratom.

Radio-frequency heating

High-frequency electromagnetic waves are generated by oscillators (often by gyrotrons or klystrons) outside the torus. If the waves have the correct frequency (or wavelength) and polarization, their energy can be transferred to the charged particles in the plasma, which in turn collide with other plasma particles, thus increasing the temperature of the bulk plasma. Various techniques exist including electron cyclotron resonance heating (ECRH) and ion cyclotron resonance heating. This energy is usually transferred by microwaves.

Tokamak particle inventory

Plasma discharges within the tokamak's vacuum chamber consist of energized ions and atoms and the energy from these particles eventually reaches the inner wall of the chamber through radiation, collisions, or lack of confinement. The inner wall of the chamber is water-cooled and the heat from the particles is removed via conduction through the wall to the water and convection of the heated water to an external cooling system. Turbomolecular or diffusion pumps allow for particles to be evacuated from the bulk volume and cryogenic pumps, consisting of a liquid helium-cooled surface, serve to effectively control the density throughout the discharge by providing an energy sink for condensation to occur. When done correctly, the fusion reactions produce large amounts of high energy neutrons. Being electrically neutral and relatively tiny, the neutrons are not affected by the magnetic fields nor are they stopped much by the surrounding vacuum chamber. The neutron flux is reduced significantly at a purpose-built neutron shield boundary that surrounds the tokamak in all directions. Shield materials vary, but are generally materials made of atoms which are close to the size of neutrons because these work best to absorb the neutron and its energy. Good candidate materials include those with much hydrogen, such as water and plastics. Boron atoms are also good absorbers of neutrons. Thus, concrete and polyethylene doped with boron make inexpensive neutron shielding materials. Once freed, the neutron has a relatively short half-life of about 10 minutes before it decays into a proton and electron with the emission of energy. When the time comes to actually try to make electricity from a tokamak-based reactor, some of the neutrons produced in the fusion process would be absorbed by a liquid metal blanket and their kinetic energy would be used in heat-transfer processes to ultimately turn a generator.

Experimental tokamaks

Currently in operation

(in chronological order of start of operations)

Alcator C-Mod

Previously operated

The control room of the Alcator C tokamak at the MIT Plasma Science and Fusion Center, in about 1982–1983.


See also


  1. Bondarenko B D "Role played by O. A. Lavrent'ev in the formulation of the problem and the initiation of research into controlled nuclear fusion in the USSR" Phys. Usp. 44 844 (2001) available online
  2. "Tokamak - Definition of tokamak by Merriam-Webster".
  3. Great Soviet Encyclopedia, 3rd edition, entry on "Токамак", available online here
  4. Peacock, N. J.; Robinson, D. C.; Forrest, M. J.; Wilcock, P. D.; Sannikov, V. V. (1969). "Measurement of the Electron Temperature by Thomson Scattering in Tokamak T3". Nature. 224 (5218): 488–490. doi:10.1038/224488a0.
  5. Penthouse founder had invested his fortune in fusion, ITER News Oct 25, 2010
  6. Kruger, S. E.; Schnack, D. D.; Sovinec, C. R., (2005). "Dynamics of the Major Disruption of a DIII-D Plasma". Phys. Plasmas 12, 056113. doi:10.1063/1.1873872.
  7. Wurden, G., (2011) International Workshop "MFE Roadmapping in the ITER Era", Princeton
  8. Baylor, L. R.; Combs, S. K.; Foust, C. R.; Jernigan, T.C.; Meitner, S. J.; Parks, P. B.; Caughman, J. B.; Fehling, D. T.; Maruyama, S.; Qualls, A. L.; Rasmussen, D. A.; Thomas, C. E., (2009). "Pellet Fuelling, ELM Pacing and Disruption Mitigation Technology Development for ITER". Nucl. Fusion 49 085013. doi:10.1088/0029-5515/49/8/085013. >
  9. Thornton, A. J.; Gibsonb, K. J.; Harrisona, J. R.; Kirka, A.; Lisgoc, S. W.; Lehnend, M.; Martina, R.;, Naylora, G.; Scannella, R.; Cullena, A. and MAST Team Thornton, A., (2011). "Disruption mitigation studies on the Mega Amp Spherical Tokamak (MAST)". Journal Nucl. Mat. 415, 1, Supplement, 1, S836-S840. doi:10.1016/j.jnucmat.2010.10.029.
  10. Goeler, V. et al. (1974). Studies of internal disruptions and m= 1 oscillations in tokamak discharges with soft—x-ray techniques, Physical Review Letters, vol. 20, p. 1201.
  12. Vojtěch Kusý. "GOLEM @ FJFI.CVUT".
  13. 1 2 "Tokamak Department, Institute of Plasma Physics".
  14. History of Golem
  15. Tore Supra Archived November 15, 2012, at the Wayback Machine.
  16. DIII-D (video)
  17. "Centro de Fusão Nuclear".
  18. Fusion Research: Australian Connections, Past and Future B. D. Blackwell, (1) M.J. Hole, J. Howard and J. O'Connor
  19. "MIT Plasma Science & Fusion Center: research>alcator>".
  20. "Pegasus Toroidal Experiment".
  21. The SST-1 Tokamak Page Archived June 20, 2014, at the Wayback Machine.
  22. "Tokamak". Retrieved 2012-06-28.
  23. 1 2 Tokamak. "Tokamak Energy – About Us".
  24. Ramos, J.; Meléndez, L.; et al. (1983). "Diseño del Tokamak Novillo" (PDF). Rev. Mex. Fís. 29 (4): 551–592.
  25. "ITER & Beyond. The Phases of ITER.". Retrieved 12 September 2012.
  27. "Concept design of CFETR superconducting magnet system based on different maintenance ports". Fusion Engineering and Design. 88: 2960–2966. doi:10.1016/j.fusengdes.2013.06.008.
  28. Song, Yun Tao; et al. (2014). "Concept Design of CFETR Tokamak Machine". IEEE Transactions on Plasma Science. 42 (3): 503–509. doi:10.1109/TPS.2014.2299277.


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