Tokamak: Difference between revisions
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A tokamak is a magnetic confinement device in which the poloidal component of the magnetic field is generated mainly by currents flowing in the plasma. | A tokamak is a [[Magnetic confinement|magnetic confinement]] device in which the poloidal component of the magnetic field is generated mainly by currents flowing in the plasma. | ||
The relative simplicity of the tokamak design has led to an initial headway of this design with respect to other prospective designs for a [[Nuclear fusion|fusion]] reactor, and the top performance among current fusion experiments has been achieved in tokamaks. As a consequence, next-step devices are based on this design. | The relative simplicity of the tokamak design has led to an initial headway of this design with respect to other prospective designs for a [[Nuclear fusion|fusion]] reactor, and the top performance among current fusion experiments has been achieved in tokamaks. As a consequence, next-step devices are based on this design. | ||
However, the intrinsic limitations of tokamaks when operated at high values of the operational parameters may lead to an eventual preference for the [[Stellarator reactor|stellarator]] design, in spite of its increased complexity. | However, the intrinsic limitations of tokamaks when operated at high values of the operational parameters may lead to an eventual preference for the [[Stellarator reactor|stellarator]] design, in spite of its increased complexity. | ||
Revision as of 10:45, 26 October 2009
A tokamak is a magnetic confinement device in which the poloidal component of the magnetic field is generated mainly by currents flowing in the plasma. The relative simplicity of the tokamak design has led to an initial headway of this design with respect to other prospective designs for a fusion reactor, and the top performance among current fusion experiments has been achieved in tokamaks. As a consequence, next-step devices are based on this design. However, the intrinsic limitations of tokamaks when operated at high values of the operational parameters may lead to an eventual preference for the stellarator design, in spite of its increased complexity.
Defunct tokamaks
- Alcator A (USA)
- Alcator C (USA)
- CASTOR (Prague, Czech Republic)
- Electric Tokamak (USA)
- LT-1 (Australia)
- PBX-M (Princeton, NJ, USA)
- RTP (Rijnhuizen, The Netherlands)
- START (UK)
- T-3 (Russia)
- T-4 (Russia)
- T-15 (Russia)
- TEXT (USA)
- TFTR (USA)
- TJ-I (Spain)
- Tokamak de Varennes (Canada)
Operational tokamaks
- Aditya (Gujarat, India)
- Alcator C-Mod (Cambridge, USA)
- ASDEX Upgrade (Garching, Germany)
- COMPASS (Prague, Czech Republic - previously in Culham, UK)
- DIII-D (San Diego, USA)
- EAST (HT-7U) (Hefei, China)
- FTU (Frascati, Italy)
- HT-7 (Hefei, China)
- ISTTOK (Lisbon, Portugal)
- JET (UK - European)
- JT-60 (Naka, Japan)
- KSTAR (Daejon, South Korea)
- MAST (Culham, UK)
- NSTX (Princeton, NJ, USA)
- STOR-M (Canada)
- T-10 (Russia)
- TCABR (Sao Paulo, Brazil - previously in Switzerland)
- TCV (Switzerland)
- TEXTOR (Jülich, Germany)
- Tore Supra (Cadarache, France)
Future tokamaks
See also
- Wikipedia:Tokamak
- www.tokamak.info - comprehensive list of tokamaks