A tokamak (Russian: Токамáк) is a device that uses a powerful magnetic field to confine a hot plasma in the shape of a torus. The tokamak is one of the several types of magnetic confinement devices being developed to produce controlled thermonuclear fusion power. As of 2018, it is the leading candidate for a practical fusion reactor.
Tokamaks were invented in the 1950s by Soviet physicists Igor Tamm and Andrei Sakharov, inspired by an original idea of Oleg Lavrentiev. It had earlier been demonstrated that a stable plasma equilibrium requires magnetic field lines that wind around the torus in a helix. Devices like the z-pinch and stellarator had attempted this, but demonstrated serious instabilities anyway. It was the development of the concept now known as the safety factorthat guided tokamak development; by arranging the reactor so this critical factor q was always greater than 1, the tokamaks strongly suppressed the instabilities that plagued earlier designs.
The first tokamak, T-1, began operation in 1958. By the mid-1960s the tokamak designs began to show greatly improved performance. Initial results were released in 1965, but were ignored; Lyman Spitzer dismissed them out of hand. A second set of results was published in 1968, this time claiming performance far in advance of any other machine. Recognizing these claims might be dismissed as well, the Soviet delegation invited a team from the United Kingdom to verify the result. The team’s 1969 publication confirmed the improvements, resulting in a stampede of tokamak construction. By the mid-1970s, dozens of tokamaks were in use around the world.
By the late 1970s, newer machines had reached all of the conditions needed for practical fusion, although not at the same time nor in a single reactor. With the goal of breakeven now in sight, in the late 1970s a new series of machines were designed that would run on a fuel of deuterium and tritium, as opposed to test gasses. These machines, notably the Joint European Torus (JET) and Tokamak Fusion Test Reactor (TFTR), had the explicit goal of reaching breakeven. Instead, they demonstrated new problems that limited their performance and solving these would require a much larger and more expensive machine. After an initial agreement between Ronald Reagan and Mikhail Gorbachev in November 1985, the ITER reactor effort emerged, and remains the primary international effort to develop practical fusion power. Many smaller designs, and offshoots like the spherical tokamak, continue to be used to investigate performance parameters and other issues.
- 3Tokamak design
- 4Plasma heating
- 5Tokamak particle inventory
- 6Experimental tokamaks
- 7See also
- 10External 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.
The term was created by Igor Golovin, the vice-director of the Laboratory of Measuring Apparatus of Academy of Science, today’s Kurchatov Institute. A similar term, “tokamag”, was also proposed for a time.
The history of magnetic confinement fusion can be broken into three periods. The early era started in the 1940s but really began in earnest in 1951, a period in which it was widely believed that practical fusion devices would be relatively simple and that the only issue was to find a working design among the many proposals. The second period, “the doldrums”, started in 1958 when a public release of information from the US, UK and USSR demonstrated that all three efforts were beset by serious problems and there was no clear path forward. The third era begins with the release of information on the tokamak in 1968 and continues to this day.
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 of which were the 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.
Unlike previous toroidal devices that used an electrical current in the plasma as a primary confinement method, the tokamak owed its stability to deliberately reducing this current in relation to the external magnetic fields. In earlier machines, the current was also used to heat the plasma, but this could no longer be used because the current had to be limited to values far below the levels needed for practical heating. With the stability problem solved, heating now emerged as the major area of study.
This led to a series of new designs that explored a variety of ways to heat the plasma. Among these were the Adiabatic Toroidal Compressor that explored the creation of a warm plasma and then heating it through rapid compression. Radio frequency heating, essentially a plasma microwave oven, was another common concept, but limited in how rapidly it could deliver its energy. The most successful method during these experiments was neutral beam injection, in which fuel atoms were shot into the plasma using a particle accelerator, where collisions with other atoms would ionize and trap them. This had the added advantage of delivering new fuel to the plasma.
By the mid-1970s these second-generation machines had reached all of the conditions needed to produce fusion, although not in a single machine, and using only “test gasses”, typically hydrogen, instead of the real fusion fuel consisting of deuterium and tritium. At this point, the first serious attempts to produce a breakeven tokamak design emerged, notably the Joint European Torus in the UK, the Tokamak Fusion Test Reactor in the US, and the T-10 in the USSR.
The excitement was so widespread that several commercial ventures to produce commercial tokamaks began around this time. Best known among these, 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. Funding by the Riggs Bank led to this effort being known as the Riggatron.
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.
The simplest magnetic confinement system is a solenoid. A plasma in a solenoid will spiral about the lines of force running down its center, preventing motion towards the sides. However, this does not prevent motion towards the ends. The obvious solution is to bend the solenoid around into a circle, forming a torus. However, it was demonstrated that such an arrangement is not uniform; for purely geometric reasons, the field on the outside edge of the torus is lower than on the inside edge. This asymmetry causes the electrons and ions to drift across the field, and eventually hit the walls of the torus.
The solution is to shape the lines so they do not simply run around the torus, but are also twist like the stripes on a barber pole or candycane. In such a field any single particle will find itself at the outside edge where it will drift one way, say up, and then as it follows its magnetic line around the torus it will find itself on the inside edge, where it will drift the other way. This cancellation is not perfect, but calculations showed it was enough to allow the fuel to remain in the reactor for a useful time.
The two first solutions to making a design with the required twist were the stellarator which did so through a mechanical arrangement, twisting the entire torus, and the z-pinch design which ran an electrical current through the plasma to create a second magnetic field to the same end. Both demonstrated improved confinement times compared to a simple torus, but both also demonstrated a variety of effects that caused the plasma to be lost from the reactors at rates that were not sustainable.
The tokamak is essentially identical to the z-pinch concept in its physical layout. Its key innovation was the realization that the instabilities that were causing the pinch to lose its plasma could be controlled. The issue was how “twisty” the fields were; fields that caused the particles to transit inside and out more than once per orbit around the long axis torus were much more stable than devices that had less twist. This ratio of twists to orbits became known as the safety factor, denoted q. Previous devices operated at q about ⅓, while the tokamak operates at q >> 1. This increases stability by orders of magnitude.
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.
While the tokamak addresses the issue of plasma stability in a gross sense, plasmas are also subject to a number of dynamic instabilities. One of these, the kink instability, is strongly suppressed by the tokamak layout, a side-effect of the high safety factors of tokamaks. The lack of kinks allowed the tokamak to operate at much higher temperatures than previous machines, and this allowed a host of new phenomenon to appear.
One of these, the banana orbits, is caused by the wide range of particle energies in a tokamak – much of the fuel is hot but a certain percentage is much cooler. Due to the high twist of the fields in the tokamak, particles following their lines of force rapidly move towards the inner edge and then outer. As they move inward they are subject to increasing magnetic fields due to the smaller radius concentrating the field. The low-energy particles in the fuel will reflect off this increasing field and begin to travel backwards through the fuel, colliding with the higher energy nuclei and scattering them out of the plasma. This process causes fuel to be lost from the reactor, although this process is slow enough that a practical reactor is still well within reach.
One of the first goals for any controlled fusion devices is to reach breakeven, the point where the energy being released by the fusion reactions is equal to the amount of energy being used to maintain the reaction. The ratio of input to output energy is denoted Q, and breakeven corresponds to a Q of 1. A Q of at least one is needed for the reactor to generate net energy, but for practical reasons, it is desirable for it to be much higher.
Once breakeven is reached, further improvements in confinement generally lead to a rapidly increasing Q. That is because some of the energy being given off by the fusion reactions of the most common fusion fuel, a 50-50 mix of deuterium and tritium, is in the form of alpha particles. These can collide with the fuel nuclei in the plasma and heat it, reducing the amount of external heat needed. At some point, known as ignition, this internal self-heating is enough to keep the reaction going without any external heating, corresponding to an infinite Q.
In the case of the tokamak, this self-heating process is maximized if the alpha particles remain in the fuel long enough to guarantee they will collide with the fuel. As the alphas are electrically charged, they are subject to the same fields that are confining the fuel plasma. The amount of time they spend in the fuel can be maximized by ensuring their orbit in the field remains within the plasma. It can be demonstrated that this occurs when the electrical current in the plasma is about 3 MA.
The safety factor varies across the axis of the machine; for purely geometrical reasons, it is always smaller at the inside edge of the plasma closest to the machine’s center because the long axis is shorter there. That means that a machine with an average q = 2 might still be less than 1 in certain areas. In the 1970s, it was suggested that one way to counteract this and produce a design with a higher average q would be to shape the magnetic fields so that the plasma only filled the outer half of the torus, shaped like a D or C when viewed end-on, instead of the normal circular cross section.
One of the first machines to incorporate a D-shaped plasma was the JET, which began its design work in 1973. This decision was made both for theoretical reasons as well as practical; because the force is larger on the inside edge of the torus, there is a large net force pressing inward on the entire reactor. The D-shape also had the advantage of reducing the net force, as well as making the supported inside edge flatter so it was easier to support. Code exploring the general layout noticed that a non-circular shape would slowly drift vertically, which led to the addition of an active feedback system to hold it in the center. Once JET had selected this layout, the General Atomics Doublet III team redesigned that machine into the D-IIID with a D-shaped cross-section. This layout has been largely universal since then.
One problem seen in all fusion reactors is that the presence of heavier elements causes energy to be lost at an increased rate, cooling the plasma. During the very earliest development of fusion power, a solution to this problem was found, the divertor, essentially a large mass spectrometer that would cause the heavier elements to be flung out of the reactor. However, designing a divertor for a tokamak proved to be a very difficult design problem.
Another problem seen in all fusion designs is the heat load that the plasma places on the wall of the confinement vessel. There are materials that can handle this load, but they are generally undesirable and expensive heavy metals. When such materials are sputtered in collisions with hot ions, their atoms mix with the fuel and rapidly cool it. A solution found on most tokamak designs is the limiter, a small ring of light metal that projected into the chamber so that the plasma would hit it before hitting the walls. This eroded the limiter and caused its atoms to mix with the fuel, but these cause less disruption than the wall materials.
When reactors moved to the D-shaped plasmas it was quickly noted that the escaping particle flux of the plasma could be shaped as well. Over time, this led to the idea of using the fields to create an internal divertor that flings the heavier elements out of fuel, typically towards the bottom of the reactor. There, a pool of liquid lithiummetal is used as a sort of limiter; the particles hit it and are rapidly cooled, remaining in the lithium. This internal pool is much easier to cool, due to its location, and although some lithium atoms are released into the plasma, its light weight makes it a much smaller problem than even the lightest metals used previously.
As machines began to explore this newly shaped plasma, they noticed that certain arrangements of the fields and plasma parameters would sometimes enter what is now known as the high-confinement mode, or H-mode, which operated stably at higher temperatures and pressures. Operating in the H-mode, which can also be seen in stellarators, is now a major design goal of the tokamak design.
Finally, it was noted that when the plasma had a non-uniform density would give rise to internal electrical currents. This is known as the bootstrap current. This allows a properly designed reactor to generate some of the internal current needed to twist the magnetic field lines without having to supply it from an external source. This has a number of advantages, and modern designs all attempt to generate as much of their total current through the bootstrap process as possible.
By the early 1990s, the combination of these features and others collectively gave rise to the “advanced tokamak” concept. This forms the basis of modern research, including ITER.
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 (-1 MV) are being developed for ITER. The ITER Neutral Beam Test Facility in Padova will be the first ITER facility to start operation. 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 as Castor; since 2007 as Golem) in Prague, Czech Republic. In operation in Kurchatov Institutesince the 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; (Being upgraded 2015-2018 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, United States; 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: 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: 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 Swiss Plasma Centerin Switzerland
- 1995: HT-7, in Hefei, at the Institute of Plasma Physics, 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, at The Hefei Institutes of Physical Science, 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: Medusa CR, in Cartago, at The Instituto Tecnológico de Costa Rica, Costa Rica
- 2012: SST-1, in Gandhinagar, at the Institute for Plasma Research, India (ITER member)
- 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 a device to explore toroidal configurations, independently discovering the tokamak layout
- 1970: Stellarator C reopens as the Symmetric Tokamak in May at PPPL
- 1971–1980: Texas Turbulent Tokamak, University of Texas at Austin, US
- 1972: The Adiabatic Toroidal Compressor begins operation at PPPL
- 1973–1976: Tokamak de Fontenay aux Roses (TFR), near Paris, France
- 1973–1979: Alcator A, MIT, US
- 1975: Princeton Large Torus begins operation at PPPL
- 1978–1987: Alcator C, MIT, US
- 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)
- 1980-1990: Tokoloshe Tokamak, Atomic Energy Board, South Africa
- 1980–2004: TEXT/TEXT-U, University of Texas at Austin, US
- 1982–1997: TFTR, Princeton University, US
- 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, US
- 1992–2016: Alcator C-Mod, MIT, Cambridge, US
- 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
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- Vojtěch Kusý. “GOLEM @ FJFI.CVUT”. cvut.cz.
- “Tokamak Department, Institute of Plasma Physics”. cas.cz. Archived from the original on 2015-09-01.
- History of Golem
- Tore Supra Archived November 15, 2012, at the Wayback Machine.
- DIII-D (video)
- EMazzitelli, Giuseppe. “ENEA-Fusion: FTU”. http://www.fusione.enea.it.
- “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
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- “MIT Plasma Science & Fusion Center: research>alcator>”. mit.edu. Archived from the original on 2015-07-09.
- “ITER & Beyond. The Phases of ITER”. Archived from the original on 22 September 2012. 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. Bibcode:2014ITPS…42..503S. 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.
- Wesson, John (November 1999). The Science of JET (PDF). JET Joint Undertaking. p. 20.
|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. Check date values in:
- Observer Newspaper Article on Tokomak Nuclear fusion and the promise of a brighter tomorrow