A tokamak (Russian: токамак) 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.
- 1 Etymology
- 2 History
- 3 Toroidal design
- 4 Plasma heating
- 5 Tokamak particle inventory
- 6 Experimental tokamaks
- 7 See also
- 8 Notes
- 9 References
- 10 External links
- "тороидальная камера с магнитными катушками" (toroidal'naya kamera s magnitnymi katushkami) — toroidal chamber with magnetic coils;
- "тороидальная камера с аксиальным магнитным полем" (toroidal'naya kamera s aksial'nym magnitnym polem) — toroidal chamber with axial magnetic field.
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.
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.
In 1973 design work on JET, the Joint European Torus, began.
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.
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.
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. 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.[dubious ][page needed]
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 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 (-500 kV) are proposed for ITER. 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.
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, the process has the additional benefit of facilitating attainment of the required density for a fusion reactor.
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.
Currently in operation
(in chronological order of start of operations)
- 1960s: TM1-MH (since 1977 Castor; since 2007 Golem) in Prague, Czech Republic. In operation in Kurchatov Institute since early 1960s but renamed to Castor in 1977 and moved to IPP CAS, Prague; in 2007 moved to FNSPE, Czech Technical University in Prague and renamed to Golem.
- 1975: T-10, in Kurchatov Institute, Moscow, Russia (formerly Soviet Union); 2 MW
- 1983: Joint European Torus (JET), in Culham, United Kingdom
- 1985: JT-60, in Naka, Ibaraki Prefecture, Japan; (Currently undergoing upgrade to Super, Advanced model)
- 1987: STOR-M, University of Saskatchewan; Canada; first demonstration of alternating current in a tokamak.
- 1988: Tore Supra, at the CEA, Cadarache, France
- 1989: Aditya, at Institute for Plasma Research (IPR) in Gujarat, India
- 1980s: DIII-D, in San Diego, USA; operated by General Atomics since the late 1980s
- 1989: COMPASS, in Prague, Czech Republic; in operation since 2008, previously operated from 1989 to 1999 in Culham, United Kingdom
- 1990: FTU, in Frascati, Italy
- 1991: Tokamak ISTTOK, at the Instituto de Plasmas e Fusão Nuclear, Lisbon, Portugal;
- 1991: ASDEX Upgrade, in Garching, Germany
- 1992: H-1NF (H-1 National Plasma Fusion Research Facility) based on the H-1 Heliac device built by Australia National University's plasma physics group and in operation since 1992
- 1992: Alcator C-Mod, MIT, Cambridge, USA
- 1992: Tokamak à configuration variable (TCV), at the EPFL, Switzerland
- 1994: TCABR, at the University of São Paulo, São Paulo, Brazil; this tokamak was transferred from Centre des Recherches en Physique des Plasmas in Switzerland
- 1995: HT-7, in Hefei, China
- 1999: MAST, in Culham, United Kingdom
- 1999: NSTX in Princeton, New Jersey
- 1999: Globus-M in Ioffe Institute, Saint Petersburg, Russia
- 1990s: Pegasus Toroidal Experiment at the University of Wisconsin-Madison; in operation since the late 1990s
- 2002: HL-2A, in Chengdu, China
- 2006: EAST (HT-7U), in Hefei, China (ITER member)
- 2008: KSTAR, in Daejon, South Korea (ITER member)
- 2010: JT-60SA, in Naka, Japan (ITER member); upgraded from the JT-60.
- 2012: SST-1, in Gandhinagar, India (ITER member); the Institute for Plasma Research reports 1000 seconds operation.
- 2012: IR-T1, Islamic Azad University, Science and Research Branch, Tehran, Iran
- 2012: ST25 at Tokamak Energy at Culham, Oxfordshire, UK (now at Milton Park)
- 2014: ST25 (HTS) the first tokamak to have all magnetic fields formed from high temperature superconducting magnets, at Tokamak Energy based in Oxfordshire, UK
- 1960s: T-3 and T-4, in Kurchatov Institute, Moscow, Russia (formerly Soviet Union); T-4 in operation in 1968.
- 1963: LT-1, Australia National University's plasma physics group built the first tokamak outside of Soviet Union c. 1963
- 1971–1980: Texas Turbulent Tokamak, University of Texas at Austin, USA
- 1973–1976: Tokamak de Fontenay aux Roses (TFR), near Paris, France
- 1973–1979: Alcator A, MIT, USA
- 1978–1987: Alcator C, MIT, USA
- 1978–2013: TEXTOR, in Jülich, Germany
- 1979–1998: MT-1 Tokamak, Budapest, Hungary (Built at the Kurchatov Institute, Russia, transported to Hungary in 1979, rebuilt as MT-1M in 1991)
- 1980–2004: TEXT/TEXT-U, University of Texas at Austin, USA
- 1982–1997: TFTR, Princeton University, USA
- 1983–2000: Novillo Tokamak, at the Instituto Nacional de Investigaciones Nucleares,in Mexico City, Mexico
- 1984–1992: HL-1 Tokamak, in Chengdu, China
- 1987–1999: Tokamak de Varennes; Varennes, Canada; operated by Hydro-Québec and used by researchers from Institut de recherche en électricité du Québec (IREQ) and the Institut national de la recherche scientifique (INRS)
- 1988–2005: T-15, in Kurchatov Institute, Moscow, Russia (formerly Soviet Union); 10 MW
- 1991–1998: START in Culham, United Kingdom
- 1990s–2001: COMPASS, in Culham, United Kingdom
- 1994–2001: HL-1M Tokamak, in Chengdu, China
- 1999–2005: UCLA Electric Tokamak, in Los Angeles, USA
- ITER, international project in Cadarache, France; 500 MW; construction began in 2010, first plasma expected in 2020.
- DEMO; 2000 MW, continuous operation, connected to power grid. Planned successor to ITER; construction to begin in 2024 according to preliminary timetable.
- CFETR, also known as "China Fusion Engineering Test Reactor"; 200 MW; Next generation Chinese fusion reactor, is a new tokamak device.
- Magnetic mirrors
- Edge-Localized Mode
- Reversed-field pinch
- List of plasma (physics) articles
- Ball-pen probe
- The section on Dimensionless parameters in tokamaks in the article on Plasma scaling
- ARC fusion reactor
- 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
- "Tokamak - Definition of tokamak by Merriam-Webster". merriam-webster.com.
- Great Soviet Encyclopedia, 3rd edition, entry on "Токамак", available online here 
- 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.
- 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.
- Wurden, G., (2011) International Workshop "MFE Roadmapping in the ITER Era", Princeton
- 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. >
- 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.
- Vojtěch Kusý. "GOLEM @ FJFI.CVUT". cvut.cz.
- "Tokamak Department, Institute of Plasma Physics". cas.cz.
- History of Golem[dead link]
- Tore Supra Archived November 15, 2012, at the Wayback Machine.
- DIII-D (video)
- "Centro de Fusão Nuclear". utl.pt.
- Fusion Research: Australian Connections, Past and Future B. D. Blackwell, (1) M.J. Hole, J. Howard and J. O'Connor
- "MIT Plasma Science & Fusion Center: research>alcator>". mit.edu.
- "Pegasus Toroidal Experiment". wisc.edu.
- The SST-1 Tokamak Page Archived June 20, 2014, at the Wayback Machine.
- "Tokamak". Pprc.srbiau.ac.ir. Retrieved 2012-06-28.
- Tokamak. "Tokamak Energy – About Us". tokamakenergy.co.uk.
- Ramos, J.; Meléndez, L.; et al. (1983). "Diseño del Tokamak Novillo" (PDF). Rev. Mex. Fís. 29 (4): 551–592.
- "ITER & Beyond. The Phases of ITER.". Retrieved 12 September 2012.
- "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.
- 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.
- Braams, C.M. & Stott, P.E. (2002). Nuclear Fusion: Half a Century of Magnetic Confinement Research. Institute of Physics Publishing. ISBN 0-7503-0705-6.
- Dolan, Thomas J. (1982). Fusion Research, Volume 1 - Principles. Pergamon Press. LCC QC791.D64.
- Nishikawa, K. & Wakatani, M. (2000). Plasma Physics. Springer-Verlag. ISBN 3-540-65285-X.
- Raeder, J.; et al. (1986). Controlled Nuclear Fusion. John Wiley & Sons. ISBN 0-471-10312-8.
- Wesson, John (2000). The Science of JET (PDF).
- Wesson, John; et al. (2004). Tokamaks. Oxford University Press. ISBN 0-19-850922-7.
|Wikimedia Commons has media related to Tokamaks.|
- CCFE - site from the UK fusion research centre CCFE.
- Int'l Tokamak research - various that relate to ITER
- Plasma Science - site on tokamaks from the French CEA.
- Fusion Programs at General Atomics, including the DIII-D National Fusion Facility, an experimental tokamak.
- General Atomics DIII-D Program
- Fusion and Plasma Physics Seminar at MIT OCW
- Unofficial ITER fan club, Club for fans of the biggest tokamak planned to be built in near future.
- www.tokamak.info Extensive list of current and historic tokamaks from around the world.
- SSTC-1 Overview video of a small scale tokamak concept.
- on YouTube Section View Video of a small scale tokamak concept.
- on YouTube Fly Through Video of a small scale tokamak concept.
- LAP_Tokamak_Development Information on conditions necessary for nuclear reaction in a tokamak reactor
- A. P. Frass (1973). "Engineering Problems In The Design Of Controlled Thermonuclear Reactors" (PDF). Oak Ridge National Laboratory. Retrieved September 2013.
- Observer Newspaper Article on Tokomak Nuclear fusion and the promise of a brighter tomorrow