Fusion research facilities across Europe
The task of defining ITER operational concepts and pushing forward the research for DEMO and future fusion power plant takes place in the laboratories of the European Fusion Associations. These laboratories boast a range of experimental fusion devices both in terms of size and configuration, as well as specific facilities for technology development.
In addition to the JET tokamak there are many other experimental facilities already contributing to European fusion research.
EU Fusion Physics Devices
This suite of fusion experimental devices includes conventional tokamaks, spherical tokamaks, stellarators and reverse pinch devices (see explanatory page below).
The small- and medium-sized tokamaks within the European fusion programme include a wide range of configurations, sizes, shapes and heating systems, exploring a variety of operational scenarios. This work is building a greatly increased database for decisions on future designs of fusion devices.
The ASDEX Upgrade device is run by the German IPP Association. Based at Garching in Germany, it investigates ITER-relevant divertors, plasma-wall interactions, operation with an inner wall made entirely of tungsten, and advanced operational scenarios.
The Czech IPP.CR Association based in Prague hosts the COMPASS-D device, a small tokamak with a D-shaped plasma cross-section (like ITER). It was originally constructed and operated at the Culham site. After being mothballed there for a few years it was transferred to Prague where it has been upgraded for a variety of ITER relevant experiments and to act as a training facility.
The spherical tokamak MAST at Culham works on the operational benefits that this type of tokamak may offer to commercial plants.
FTU is a high-magnetic field, high-plasma density, high-current tokamak operated by ENEA at Frascati, Italy. The device investigates new radio-frequency plasma-heating techniques using electron cyclotron technology and techniques to reduce plasma contamination.
The Portuguese Association IST operates ISTTOK, a small tokamak in Lisbon. The device is involved in fundamental physics studies, in particular the development of theoretical descriptions of plasma and novel diagnostics.
TCV is a variable configuration tokamak for the study of specially shaped cross-sections of the plasma run by the Swiss Association in Lausanne.
TEXTOR, a tokamak run by three Associations (FZJ from Germany, FOM from the Netherlands and the Belgian State Association), is located in Julich, Germany. Experiments on the device include wall interactions, plasma confinement under additional heating, and testing of new diverter technology.
The French Association CEA runs the large tokamak TORE SUPRA at Cadarache. This device was the first large tokamak to use a series of superconducting coils to generate a permanent magnetic field. It has a circular plasma cross-section and has the capability of running long pulse plasmas on a regular basis. This allows TORE SUPRA to explore new scientific questions in ITER-relevant conditions such as erosion and hydrogen wall trapping, real-time discharge control and performance optimisation.
TJ-II is a highly flexible stellarator device with a helical magnetic axis constructed at Madrid, Spain by the CIEMAT Association. It works on novel confinement and high-efficiency operations.
A large, advanced stellarator device, the Wendelstein 7-X is being constructed at Greifswald in Germany. The device is based on data obtained with the Wendelstein 7-AS device at Garching, also operated by IPP. It studied plasma behaviour in a modular Stellarator design and tested an island diverter concept.
Reversed Field Pinch Devices (RFP)
RFX, run by the Italian ENEA Association at Padua in Italy, is investigating toroidal confinement and transport whilst evaluating future prospects for RFP technology.
EXTRAP-T2R is situated in Stockholm, Sweden, and run by the NFR Association. It supports RFX by looking, in particular, at wall stabilisation of the plasma and advanced control techniques.
There are also a large number of devices devoted to technological aspects of the fusion programme. Many of these have been, or are being, developed for specific tasks in the testing and characterisation of components for ITER. Some of the key facilities are:
Divertor Test Platform (DTP2) - a full-scale model of the ITER divertor to develop and test remote handling concepts, in particular the techniques for replacement of the large and radioactive divertor components.
NBTF - a planned Neutral Beam Test Facility in Italy to support the development, testing and remote handling of the neutral beam plasma heating system for ITER.
TLK - the Tritium Laboratory at Forschungszentrum Karlsruhe in Germany, a facility to develop the technologies for processing the fusion fuel tritium.
TOSKA - a facility for testing large superconducting coils operated by the Forschungszentrum Karlsruhe in Germany.
MAGNUM-PSI - a facility at the FOM Institute in the Netherlands for simulating the plasma conditions in the ITER divertor. It will be used to study plasma surface interaction and the behaviour of candidate plasma facing materials.
FE 200 - a thermal fatigue test facility using a 200kW electron gun, run by CEA in France and Areva.
SULTAN - a test facility for superconductor and joint samples, run by CRPP in Switzerland.
Diverter test and refurbishment platforms run by ENEA in Italy.
Tokamak, Stellarator or RPF?
Magnetic confinement devices for fusion experiments come in a variety of designs. Here is a quick guide.
In a tokamak, plasma is confined in a toroidal vessel by a magnetic field with two main components. The first (toroidal) field is produced by a set of coils equally spaced around the doughnut-shaped reactor vessel, and keeps the plasma away from the walls. However, this field is not enough to confine the plasma by itself, requiring a second (poloidal) field to counteract the natural pressure inside the plasma which tries to make it expand. This poloidal field is generated by the toroidal current flowing in the plasma itself.
In a spherical tokamak (ST) the plasma is confined in basically the same way as the conventional tokamak. The main difference, as the name suggests, is in the magnetic geometry. In an ST device the aspect ratio (the ratio of the plasma diameter to the overall device size) is low. An ST looks like a sphere with a hole through the middle rather than the doughnut shape of a conventional tokamak. One possible advantage is that an ST may be more efficient than a conventional tokamak in terms of plasma performance for a given engineering cost.
A stellarator relies entirely on magnetic fields produced by external coils to produce the magnetic confinement, eliminating the need for a toroidal plasma current but requiring a more complex shape for the coils with tokamaks. However, a stellarator is intrinsically able to maintain the confinement configuration without the use of systems to drive the plasma current, which are needed in tokamaks. Disruptions, instabilities and other plasma events associated with the free energy of a large (several million amp) toroidal current either do not occur or are strongly reduced. This means that stellarators offer an intrinsic potential for steady-state, continuous operation.
Reversed Field Pinches (RFP) devices are toroidal plasma confinement devices that are very much like tokamaks but use the electromagnetic properties of the plasma itself to generate the confining magnetic fields. RFP systems are being studied for possible low magnetic field, for high plasma density confinement designs, and to enhance understanding of the physics of toroidal confinement in operating regions outside the range of standard tokamak devices.