List of fusion experiments

From Wikipedia, the free encyclopedia
Target chamber of the Shiva laser, used for inertial confinement fusion experiments from 1978 until decommissioned in 1981
Plasma chamber of TFTR, used for magnetic confinement fusion experiments, which produced 11 MW of fusion power in 1994

Experiments directed toward developing fusion power are invariably done with dedicated machines which can be classified according to the principles they use to confine the plasma fuel and keep it hot.

The major division is between magnetic confinement and inertial confinement. In magnetic confinement, the tendency of the hot plasma to expand is counteracted by the Lorentz force between currents in the plasma and magnetic fields produced by external coils. The particle densities tend to be in the range of 1018 to 1022 m−3 and the linear dimensions in the range of 0.1 to 10 m. The particle and energy confinement times may range from under a millisecond to over a second, but the configuration itself is often maintained through input of particles, energy, and current for times that are hundreds or thousands of times longer. Some concepts are capable of maintaining a plasma indefinitely.

In contrast, with inertial confinement, there is nothing to counteract the expansion of the plasma. The confinement time is simply the time it takes the plasma pressure to overcome the inertia of the particles, hence the name. The densities tend to be in the range of 1031 to 1033 m−3 and the plasma radius in the range of 1 to 100 micrometers. These conditions are obtained by irradiating a millimeter-sized solid pellet with a nanosecond laser or ion pulse. The outer layer of the pellet is ablated, providing a reaction force that compresses the central 10% of the fuel by a factor of 10 or 20 to 103 or 104 times solid density. These microplasmas disperse in a time measured in nanoseconds. For a fusion power reactor, a repetition rate of several per second will be needed.

Magnetic confinement[]

Within the field of magnetic confinement experiments, there is a basic division between toroidal and open magnetic field topologies. Generally speaking, it is easier to contain a plasma in the direction perpendicular to the field than parallel to it. Parallel confinement can be solved either by bending the field lines back on themselves into circles or, more commonly, toroidal surfaces, or by constricting the bundle of field lines at both ends, which causes some of the particles to be reflected by the mirror effect. The toroidal geometries can be further subdivided according to whether the machine itself has a toroidal geometry, i.e., a solid core through the center of the plasma. The alternative is to dispense with a solid core and rely on currents in the plasma to produce the toroidal field.

Mirror machines have advantages in a simpler geometry and a better potential for direct conversion of particle energy to electricity. They generally require higher magnetic fields than toroidal machines, but the biggest problem has turned out to be confinement. For good confinement there must be more particles moving perpendicular to the field than there are moving parallel to the field. Such a non-Maxwellian velocity distribution is, however, very difficult to maintain and energetically costly.

The mirrors' advantage of simple machine geometry is maintained in machines which produce compact toroids, but there are potential disadvantages for stability in not having a central conductor and there is generally less possibility to control (and thereby optimize) the magnetic geometry. Compact toroid concepts are generally less well developed than those of toroidal machines. While this does not necessarily mean that they cannot work better than mainstream concepts, the uncertainty involved is much greater.

Somewhat in a class by itself is the Z-pinch, which has circular field lines. This was one of the first concepts tried, but it did not prove very successful. Furthermore, there was never a convincing concept for turning the pulsed machine requiring electrodes into a practical reactor.

The dense plasma focus is a controversial and "non-mainstream" device that relies on currents in the plasma to produce a toroid. It is a pulsed device that depends on a plasma that is not in equilibrium and has the potential for direct conversion of particle energy to electricity. Experiments are ongoing to test relatively new theories to determine if the device has a future.

Toroidal machine[]

Toroidal machines can be axially symmetric, like the tokamak and the reversed field pinch (RFP), or asymmetric, like the stellarator. The additional degree of freedom gained by giving up toroidal symmetry might ultimately be usable to produce better confinement, but the cost is complexity in the engineering, the theory, and the experimental diagnostics. Stellarators typically have a periodicity, e.g. a fivefold rotational symmetry. The RFP, despite some theoretical advantages such as a low magnetic field at the coils, has not proven very successful.

Tokamak[]

[1]

Device name Status Construction Operation Location Organisation Major/minor radius B-field Plasma current Purpose Image
T-1 (Tokamak-1) Shut down ? 1957–1959 Soviet Union Moscow Kurchatov Institute 0.625 m/0.13 m 1 T 0.04 MA First tokamak T-1
T-3 (Tokamak-3) Shut down ? 1962–? Soviet Union Moscow Kurchatov Institute 1 m/0.12 m 2.5 T 0.06 MA
ST (Symmetric Tokamak) Shut down Model C 1970–1974 United States Princeton Princeton Plasma Physics Laboratory 1.09 m/0.13 m 5.0 T 0.13 MA First American tokamak, converted from Model C stellarator
ORMAK (Oak Ridge tokaMAK) Shut down 1971–1976 United States Oak Ridge Oak Ridge National Laboratory 0.8 m/0.23 m 2.5 T 0.34 MA First to achieve 20 MK plasma temperature ORMAK plasma vessel
ATC (Adiabatic Toroidal Compressor) Shut down 1971–1972 1972–1976 United States Princeton Princeton Plasma Physics Laboratory 0.88 m/0.11 m 2 T 0.05 MA Demonstrate compressional plasma heating Schematic of ATC
Pulsator[2] Shut down 1970–1973 1973–1979 Germany Garching Max Planck Institute for Plasma Physics 0.7 m/0.12 m 2.7 T 0.125 MA Discovery of high-density operation with tokamaks
TFR (Tokamak de Fontenay-aux-Roses) Shut down 1973–1984 France Fontenay-aux-Roses CEA 1 m/0.2 m 6 T 0.49 MA
T-10 (Tokamak-10) Operational 1975- Soviet Union Moscow Kurchatov Institute 1.50 m/0.37 m 4 T 0.8 MA Largest tokamak of its time Model of the T-10
PLT (Princeton Large Torus) Shut down 1975–1986 United States Princeton Princeton Plasma Physics Laboratory 1.32 m/0.4 m 4 T 0.7 MA First to achieve 1 MA plasma current Construction of the Princeton Large Torus
Microtor[3] Shut down ? 1976–1983? United States Los Angeles UCLA 0.3 m/0.1 m 2.5 T 0.12 MA Plasma impurity control and diagnostic development
Macrotor[3] Shut down ? 1970s–80s United States Los Angeles UCLA 0.9 m/0.4 m 0.4 T 0.1 MA Understanding plasma rotation driven by radial current
ISX-B Shut down ? 1978–? United States Oak Ridge Oak Ridge National Laboratory 0.93 m/0.27 m 1.8 T 0.2 MA Superconducting coils, attempt high-beta operation
T-7 (Tokamak-7) Recycled →HT-7[4] ? 1979–1985 Soviet Union Moscow Kurchatov Institute 1.2 m/0.31 m 3 T 0.3 MA First tokamak with superconducting toroidal field coils
ASDEX (Axially Symmetric Divertor Experiment)[5] Recycled →HL-2A 1973–1980 1980–1990 Germany Garching Max-Planck-Institut für Plasmaphysik 1.65 m/0.4 m 2.8 T 0.5 MA Discovery of the H-mode in 1982
TEXTOR (Tokamak Experiment for Technology Oriented Research)[6][7] Shut down 1976–1980 1981–2013 Germany Jülich Forschungszentrum Jülich 1.75 m/0.47 m 2.8 T 0.8 MA Study plasma-wall interactions
TFTR (Tokamak Fusion Test Reactor)[8] Shut down 1980–1982 1982–1997 United States Princeton Princeton Plasma Physics Laboratory 2.4 m/0.8 m 6 T 3 MA Attempted scientific break-even, reached record fusion power of 10.7 MW and temperature of 510 MK TFTR plasma vessel
JET (Joint European Torus)[9] Operational 1978–1983 1983- United Kingdom Culham Culham Centre for Fusion Energy 2.96 m/0.96 m 4 T 7 MA Record for fusion output power 16.1 MW JET in 1991
Novillo[10][11] Shut down NOVA-II 1983–2004 Mexico Mexico City Instituto Nacional de Investigaciones Nucleares 0.23 m/0.06 m 1 T 0.01 MA Study plasma-wall interactions
JT-60 (Japan Torus-60)[12] Recycled →JT-60SA 1985–2010 Japan Naka Japan Atomic Energy Research Institute 3.4 m/1.0 m 4 T 3 MA High-beta steady-state operation, highest fusion triple product
CCT (Continuous Current Tokamak) Shut down ? 1986–199? United States Los Angeles UCLA 1.5 m/0.4 m 0.2 T 0.05 MA H-mode studies
DIII-D[13] Operational 1986[14] 1986- United States San Diego General Atomics 1.67 m/0.67 m 2.2 T 3 MA Tokamak Optimization DIII-D vacuum vessel
STOR-M (Saskatchewan Torus-Modified)[15] Operational 1987- Canada Saskatoon Plasma Physics Laboratory (Saskatchewan) 0.46 m/0.125 m 1 T 0.06 MA Study plasma heating and anomalous transport
T-15 Recycled →T-15MD 1983–1988 1988–1995 Soviet Union Moscow Kurchatov Institute 2.43 m/0.7 m 3.6 T 1 MA First superconducting tokamak T-15 on a stamp
Tore Supra[16] Recycled →WEST 1988–2011 France Cadarache Département de Recherches sur la Fusion Contrôlée 2.25 m/0.7 m 4.5 T 2 MA Large superconducting tokamak with active cooling
ADITYA (tokamak) Operational 1989- India Gandhinagar Institute for Plasma Research 0.75 m/0.25 m 1.2 T 0.25 MA
COMPASS (COMPact ASSembly)[17][18] Operational 1980- 1989- Czech Republic Prague 0.56 m/0.23 m 2.1 T 0.32 MA COMPASS plasma chamber
FTU (Frascati Tokamak Upgrade) Operational 1990- Italy Frascati ENEA 0.935 m/0.35 m 8 T 1.6 MA
START (Small Tight Aspect Ratio Tokamak)[19] Recycled →Proto-Sphera 1990–1998 United Kingdom Culham Culham Centre for Fusion Energy 0.3 m/? 0.5 T 0.31 MA First full-sized Spherical Tokamak
ASDEX Upgrade (Axially Symmetric Divertor Experiment) Operational 1991- Germany Garching Max-Planck-Institut für Plasmaphysik 1.65 m/0.5 m 2.6 T 1.4 MA ASDEX Upgrade plasma vessel segment
Alcator C-Mod (Alto Campo Toro)[20] Operational (funded by Fusion Startups) 1986- 1991–2016 United States Cambridge Massachusetts Institute of Technology 0.68 m/0.22 m 8 T 2 MA Record plasma pressure 2.05 bar Alcator C-Mod plasma vessel
ISTTOK (Instituto Superior Técnico TOKamak)[21] Operational 1992- Portugal Lisbon Instituto de Plasmas e Fusão Nuclear 0.46 m/0.085 m 2.8 T 0.01 MA
TCV (Tokamak à Configuration Variable)[22] Operational 1992- Switzerland Lausanne École Polytechnique Fédérale de Lausanne 0.88 m/0.25 m 1.43 T 1.2 MA Confinement studies TCV plasma vessel
HBT-EP (High Beta Tokamak-Extended Pulse) Operational 1993- United States New York City Columbia University Plasma Physics Laboratory 0.92 m/0.15 m 0.35 T 0.03 MA High-Beta tokamak HBT-EP sketch
HT-7 (Hefei Tokamak-7) Shut down 1991–1994 1995–2013 China Hefei Hefei Institutes of Physical Science 1.22 m/0.27 m 2 T 0.2 MA China's first superconducting tokamak
Pegasus Toroidal Experiment[23] Operational ? 1996- United States Madison University of Wisconsin–Madison 0.45 m/0.4 m 0.18 T 0.3 MA Extremely low aspect ratio Pegasus Toroidal Experiment
NSTX (National Spherical Torus Experiment)[24] Operational 1999- United States Plainsboro Township Princeton Plasma Physics Laboratory 0.85 m/0.68 m 0.3 T 2 MA Study the spherical tokamak concept National Spherical Torus Experiment
Globus-M (UNU Globus-M)[25] Operational 1999- Russia Saint Petersburg Ioffe Institute 0.36 m/0.24 m 0.4 T 0.3 MA Study the spherical tokamak concept
ET (Electric Tokamak) Recycled →ETPD 1998 1999–2006 United States Los Angeles UCLA 5 m/1 m 0.25 T 0.045 MA Largest tokamak of its time The Electric Tokamak.jpg
CDX-U (Current Drive Experiment-Upgrade) Recycled →LTX 2000–2005 United States Princeton Princeton Plasma Physics Laboratory 0.3 m/? 0.23 T 0.03 MA Study Lithium in plasma walls CDX-U setup
MAST (Mega-Ampere Spherical Tokamak)[26] Recycled →MAST-Upgrade 1997–1999 2000–2013 United Kingdom Culham Culham Centre for Fusion Energy 0.85 m/0.65 m 0.55 T 1.35 MA Investigate spherical tokamak for fusion Plasma in MAST
HL-2A (Huan-Liuqi-2A) Operational 2000–2002 2002–2018 China Chengdu 1.65 m/0.4 m 2.7 T 0.43 MA H-mode physics, ELM mitigation [1]
SST-1 (Steady State Superconducting Tokamak)[27] Operational 2001- 2005- India Gandhinagar Institute for Plasma Research 1.1 m/0.2 m 3 T 0.22 MA Produce a 1000 s elongated double null divertor plasma
EAST (Experimental Advanced Superconducting Tokamak)[28] Operational 2000–2005 2006- China Hefei Hefei Institutes of Physical Science 1.85 m/0.43 m 3.5 T 0.5 MA Superheated plasma for over 101 s at 120 M°C and 20 s at 160 M°C[29] Drawing of EAST
J-TEXT (Joint TEXT) Operational TEXT (Texas EXperimental Tokamak) 2007- China Wuhan Huazhong University of Science and Technology 1.05 m/0.26 m 2.0 T 0.2 MA Develop plasma control [2]
KSTAR (Korea Superconducting Tokamak Advanced Research)[30] Operational 1998–2007 2008- South Korea Daejeon 1.8 m/0.5 m 3.5 T 2 MA Tokamak with fully superconducting magnets, 20 s-long operation at 100 MK[31] KSTAR
LTX (Lithium Tokamak Experiment) Operational 2005–2008 2008- United States Princeton Princeton Plasma Physics Laboratory 0.4 m/? 0.4 T 0.4 MA Study Lithium in plasma walls Lithium Tokamak Experiment plasma vessel
QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak)[32] Operational 2008- Japan Kasuga Kyushu University 0.68 m/0.4 m 0.25 T 0.02 MA Study steady state operation of a Spherical Tokamak QUEST
Kazakhstan Tokamak for Material testing (KTM) Operational 2000–2010 2010- Kazakhstan Kurchatov National Nuclear Center of the Republic of Kazakhstan 0.86 m/0.43 m 1 T 0.75 MA Testing of wall and divertor
ST25-HTS[33] Operational 2012–2015 2015- United Kingdom Culham Tokamak Energy Ltd 0.25 m/0.125 m 0.1 T 0.02 MA Steady state plasma ST25-HTS with plasma
WEST (Tungsten Environment in Steady-state Tokamak) Operational 2013–2016 2016- France Cadarache Département de Recherches sur la Fusion Contrôlée 2.5 m/0.5 m 3.7 T 1 MA Superconducting tokamak with active cooling WEST chamber
ST40[34] Operational 2017–2018 2018- United Kingdom Didcot Tokamak Energy Ltd 0.4 m/0.3 m 3 T 2 MA First high field spherical tokamak ST40 engineering drawing
MAST-U (Mega-Ampere Spherical Tokamak Upgrade)[35] Operational 2013–2019 2020- United Kingdom Culham Culham Centre for Fusion Energy 0.85 m/0.65 m 0.92 T 2 MA Test new exhaust concepts for a spherical tokamak
HL-2M (Huan-Liuqi-2M)[36] Operational 2018–2019 2020- China Leshan 1.78 m/0.65 m 2.2 T 1.2 MA Elongated plasma with 200 MK HL-2M
JT-60SA (Japan Torus-60 super, advanced)[37] Operational 2013–2020 2021– Japan Naka Japan Atomic Energy Research Institute 2.96 m/1.18 m 2.25 T 5.5 MA Optimise plasma configurations for ITER and DEMO with full non-inductive steady-state operation panorama of JT-60SA
T-15MD Operational 2010–2020 2021- Russia Moscow Kurchatov Institute 1.48 m/0.67 m 2 T 2 MA Hybrid fusion/fission reactor T-15MD coil system
ITER[38] Under construction 2013–2025? 2025? France Cadarache ITER Council 6.2 m/2.0 m 5.3 T 15 MA ? Demonstrate feasibility of fusion on a power-plant scale with 500 MW fusion power Small-scale model of ITER
DTT (Divertor Tokamak Test facility)[39][40] Planned 2022–2025? 2025? Italy Frascati ENEA 2.14 m/0.70 m 6 T ? 5.5 MA ? Superconducting tokamak to study power exhaust [3]
SPARC[41][42] Planned 2021–? 2025? United States Devens Commonwealth Fusion Systems and MIT Plasma Science and Fusion Center 1.85 m/0.57 m 12.2 T 8.7 MA Compact, high-field tokamak with ReBCO coils and 100 MW planned fusion power
IGNITOR[43] Planned[44] ? >2024 Russia Troitzk ENEA 1.32 m/0.47 m 13 T 11 MA ? Compact fusion reactor with self-sustained plasma and 100 MW of planned fusion power
SST-2 (Steady State Tokamak-2)[45] Planned 2027? India Gujarat Institute for Plasma Research 4.42 m/1.47 m 5.42 T 11.2 MA Full-fledged fusion reactor with tritium breeding and up to 500 MW output
CFETR (China Fusion Engineering Test Reactor)[46] Planned 2020? 2030? China Institute of Plasma Physics, Chinese Academy of Sciences 5.7 m/1.6 m ? 5 T ? 10 MA ? Bridge gaps between ITER and DEMO, planned fusion power 1000 MW [4]
ST-F1[47] Planned 2027? United Kingdom Didcot Tokamak Energy Ltd ? 4 T 5 MA Spherical tokamak with Q=3 and hundreds of MW planned electrical output
STEP (Spherical Tokamak for Energy Production) Planned 2032? 2040? United Kingdom Culham Culham Centre for Fusion Energy 3 m/2 m ? ? ? Spherical tokamak with hundreds of MW planned electrical output
(Korean fusion demonstration tokamak reactor)[48] Planned 2037? South Korea 6.8 m/2.1 m 7 T 12 MA ? Prototype for the development of commercial fusion reactors with around 2200 MW of fusion power Engineering drawing of planned KDEMO
DEMO (DEMOnstration Power Station) Planned 2031? 2044? ? 9 m/3 m ? 6 T ? 20 MA ? Prototype for a commercial fusion reactor Artist's conception of DEMO

Stellarator[]

Device name Status Construction Operation Type Location Organisation Major/minor radius B-field Purpose Image
Model A Shut down 1952–1953 1953–? Figure-8 United States Princeton Princeton Plasma Physics Laboratory 0.3 m/0.02 m 0.1 T First stellarator [5]
Model B Shut down 1953–1954 1954–1959 Figure-8 United States Princeton Princeton Plasma Physics Laboratory 0.3 m/0.02 m 5 T Development of plasma diagnostics
Model B-1 Shut down ?-1959 Figure-8 United States Princeton Princeton Plasma Physics Laboratory 0.25 m/0.02 m 5 T Yielded 1 MK plasma temperatures
Model B-2 Shut down 1957 Figure-8 United States Princeton Princeton Plasma Physics Laboratory 0.3 m/0.02 m 5 T Electron temperatures up to 10 MK [6]
Model B-3 Shut down 1957 1958- Figure-8 United States Princeton Princeton Plasma Physics Laboratory 0.4 m/0.02 m 4 T Last figure-8 device, confinement studies of ohmically heated plasma
Model B-64 Shut down 1955 1955 Square United States Princeton Princeton Plasma Physics Laboratory ? m/0.05 m 1.8 T
Model B-65 Shut down 1957 1957 Racetrack United States Princeton Princeton Plasma Physics Laboratory [7]
Model B-66 Shut down 1958 1958–? Racetrack United States Princeton Princeton Plasma Physics Laboratory
Wendelstein 1-A Shut down 1960 Racetrack Germany Garching Max-Planck-Institut für Plasmaphysik 0.35 m/0.02 m 2 T ℓ=3
Wendelstein 1-B Shut down 1960 Racetrack Germany Garching Max-Planck-Institut für Plasmaphysik 0.35 m/0.02 m 2 T ℓ=2
Model C Recycled →ST 1957–1962 1962–1969 Racetrack United States Princeton Princeton Plasma Physics Laboratory 1.9 m/0.07 m 3.5 T Found large plasma losses by Bohm diffusion
L-1 Shut down 1963 1963–1971 Soviet Union Lebedev Lebedev Physical Institute 0.6 m/0.05 m 1 T
SIRIUS Shut down 1964–? Racetrack Soviet Union Kharkiv
TOR-1 Shut down 1967 1967–1973 Soviet Union Lebedev Lebedev Physical Institute 0.6 m/0.05 m 1 T
TOR-2 Shut down ? 1967–1973 Soviet Union Lebedev Lebedev Physical Institute 0.63 m/0.036 m 2.5 T
Uragan-1 Shut down ? 1967–? Racetrack Soviet Union Kharkiv National Science Center, Kharkiv Institute of Physics and Technology (NSC KIPT) 1.1 m/0.1 m 1 T ?
Wendelstein 2-A Shut down 1965–1968 1968–1974 Heliotron Germany Garching Max-Planck-Institut für Plasmaphysik 0.5 m/0.05 m 0.6 T Good plasma confinement “Munich mystery” Wendelstein 2-A
Wendelstein 2-B Shut down ?-1970 1971–? Heliotron Germany Garching Max-Planck-Institut für Plasmaphysik 0.5 m/0.055 m 1.25 T Demonstrated similar performance as tokamaks Wendelstein 2-B
L-2 Shut down ? 1975–? Soviet Union Lebedev Lebedev Physical Institute 1 m/0.11 m 2.0 T
WEGA (Wendelstein Experiment in Greifswald für die Ausbildung) Recycled →HIDRA 1972–1975 1975–2013 Classical stellarator Germany Greifswald Max-Planck-Institut für Plasmaphysik 0.72 m/0.15 m 1.4 T Test lower hybrid heating WEGA
Wendelstein 7-A Shut down ? 1975–1985 Classical stellarator Germany Garching Max-Planck-Institut für Plasmaphysik 2 m/0.1 m 3.5 T First "pure" stellarator without plasma current
Heliotron-E Shut down ? 1980–? Heliotron Japan 2.2 m/0.2 m 1.9 T
Heliotron-DR Shut down ? 1981–? Heliotron Japan 0.9 m/0.07 m 0.6 T
( [uk])[49] Operational ? 1982–?[50] Torsatron Ukraine Kharkiv National Science Center, Kharkiv Institute of Physics and Technology (NSC KIPT) 1.0 m/0.12 m 1.3 T ?
Auburn Torsatron (AT) Shut down ? 1984–1990 Torsatron United States Auburn Auburn University 0.58 m/0.14 m 0.2 T Auburn Torsatron
Wendelstein 7-AS Shut down 1982–1988 1988–2002 Modular, advanced stellarator Germany Garching Max-Planck-Institut für Plasmaphysik 2 m/0.13 m 2.6 T First H-mode in a stellarator in 1992 Wendelstein 7-AS
Advanced Toroidal Facility (ATF) Shut down 1984–1988[51] 1988–? Torsatron United States Oak Ridge Oak Ridge National Laboratory 2.1 m/0.27 m 2.0 T High-beta operation
Compact Helical System (CHS) Shut down ? 1989–? Heliotron Japan Toki 1 m/0.2 m 1.5 T
Compact Auburn Torsatron (CAT) Shut down ?-1990 1990–2000 Torsatron United States Auburn Auburn University 0.53 m/0.11 m 0.1 T Study magnetic flux surfaces Compact Auburn Torsatron
H-1 (Heliac-1)[52] Operational 1992- Heliac Australia Canberra Research School of Physical Sciences and Engineering, Australian National University 1.0 m/0.19 m 0.5 T H-1NF plasma vessel
(Tokamak de la Junta Kiel)[53] Operational TJ-IU 1994- Torsatron Germany Kiel, Stuttgart University of Stuttgart 0.60 m/0.10 m 0.5 T Teaching
TJ-II (Tokamak de la Junta II)[54] Operational 1991-1996 1997- flexible Heliac Spain Madrid National Fusion Laboratory, Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas 1.5 m/0.28 m 1.2 T Study plasma in flexible configuration CAD drawing of TJ-II
LHD (Large Helical Device)[55] Operational 1990–1998 1998- Heliotron Japan Toki 3.5 m/0.6 m 3 T Determine feasibility of a stellarator fusion reactor LHD cross section
HSX (Helically Symmetric Experiment) Operational 1999- Modular, quasi-helically symmetric United States Madison University of Wisconsin–Madison 1.2 m/0.15 m 1 T Investigate plasma transport HSX with clearly visible non-planar coils
Heliotron J (Heliotron J)[56] Operational 2000- Heliotron Japan Kyoto 1.2 m/0.1 m 1.5 T Study helical-axis heliotron configuration
Columbia Non-neutral Torus (CNT) Operational ? 2004- Circular interlocked coils United States New York City Columbia University 0.3 m/0.1 m 0.2 T Study of non-neutral plasmas
(M)[49] Operational 1988–2006 2006-[57] Heliotron, Torsatron Ukraine Kharkiv National Science Center, Kharkiv Institute of Physics and Technology (NSC KIPT) 1.7 m/0.24 m 2.4 T ?
Quasi-poloidal stellarator (QPS)[58][59] Cancelled 2001–2007 - Modular United States Oak Ridge Oak Ridge National Laboratory 0.9 m/0.33 m 1.0 T Stellarator research Engineering drawing of the QPS
NCSX (National Compact Stellarator Experiment) Cancelled 2004–2008 - Helias United States Princeton Princeton Plasma Physics Laboratory 1.4 m/0.32 m 1.7 T High-β stability CAD drawing of NCSX
Compact Toroidal Hybrid (CTH) Operational ? 2007?- Torsatron United States Auburn Auburn University 0.75 m/0.2 m 0.7 T Hybrid stellarator/tokamak CTH
HIDRA (Hybrid Illinois Device for Research and Applications)[60] Operational 2013–2014 (WEGA) 2014- ? United States Urbana, IL 0.72 m/0.19 m 0.5 T Stellarator and tokamak in one device HIDRA after its reassembly in Illinois
UST_2[61] Operational 2013 2014- modular three period quasi-isodynamic Spain Madrid Charles III University of Madrid 0.29 m/0.04 m 0.089 T 3D-printed stellarator UST_2 design concept
Wendelstein 7-X[62] Operational 1996–2015 2015- Helias Germany Greifswald Max-Planck-Institut für Plasmaphysik 5.5 m/0.53 m 3 T Steady-state plasma in fully optimized stellarator Schematic diagram of Wendelstein 7-X
SCR-1 (Stellarator of Costa Rica) Operational 2011–2015 2016- Modular Costa Rica Cartago Costa Rica Institute of Technology 0.14 m/0.042 m 0.044 T SCR-1 vacuum vessel drawing
CFQS (Chinese First Quasi-Axisymmetric Stellarator)[63] Under construction 2017 – Helias China Chengdu Southwest Jiaotong University, National Institute for Fusion Science in Japan 1 m/0.25 m 1 T m=2 quasi-axisymmetric stellarator, modular CFQS coils and field

Magnetic mirror[]

Toroidal Z-pinch[]

  • Perhapsatron (1953, USA)
  • ZETA (Zero Energy Thermonuclear Assembly) (1957, United Kingdom)

Reversed field pinch (RFP)[]

  • ETA-BETA II in Padua, Italy (1979–1989)
  • RFX (Reversed-Field eXperiment), Consorzio RFX, Padova, Italy[64]
  • MST (Madison Symmetric Torus), University of Wisconsin–Madison, United States[65]
  • T2R, Royal Institute of Technology, Stockholm, Sweden
  • TPE-RX, AIST, Tsukuba, Japan
  • KTX (Keda Torus eXperiment) in China (since 2015)[66]

Spheromak[]

Field-reversed configuration (FRC)[]

Open field lines[]

Plasma pinch[]

  • Trisops – 2 facing theta-pinch guns
  • FF-2B, Lawrenceville Plasma Physics, United States[67]

Levitated dipole[]

Inertial confinement[]

Main article: Inertial confinement fusion

Laser-driven[]

Current or under construction experimental facilities[]

Solid state lasers[]
Gas lasers[]
  • NIKE laser at the Naval Research Laboratories, Krypton Fluoride gas laser
  • PALS, formerly the "Asterix IV", at the Academy of Sciences of the Czech Republic,[72] 1 kJ max. output iodine laser at 1.315 micrometre fundamental wavelength

Dismantled experimental facilities[]

Solid-state lasers[]
Gas lasers[]
  • "Single Beam System" or simply "67" after the building number it was housed in, a 1 kJ carbon dioxide laser at Los Alamos National Laboratory
  • , 2 beams, 2.5 kJ carbon dioxide laser at LANL
  • , 8 beam, ~10 kJ carbon dioxide laser at LANL — Media at Wikimedia Commons
  • at LANL. (40 kJ CO2 laser, largest ever built, production of hot electrons in target plasma due to long wavelength of laser resulted in poor laser/plasma energy coupling)
  • 96 beam 1.3 kJ total krypton fluoride (KrF) laser at LANL
  • few joules/pulse laser at the Central Laser Facility, Rutherford Appleton Laboratory

Z-pinch[]

Main article: Z-pinch

Inertial electrostatic confinement[]

Main article: Inertial electrostatic confinement

Magnetized target fusion[]

Main article: Magnetized target fusion

References[]

  1. ^ "International tokamak research". ITER.
  2. ^ "Pulsator".
  3. ^ a b Taylor, R. J., Lee, P., & Luhmann Jr, N. C. (1981). ICRF heating, particle transport and fluctuations in tokamaks (PDF) (Report).{{cite report}}: CS1 maint: uses authors parameter (link)
  4. ^ Robert Arnoux (2009-05-18). "From Russia with love".
  5. ^ "ASDEX". www.ipp.mpg.de.
  6. ^ "Forschungszentrum Jülich – Plasmaphysik (IEK-4)". fz-juelich.de (in German).
  7. ^ Progress in Fusion Research – 30 Years of TEXTOR
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