Experimental facilities

The many fusion research institutions across Europe together operate more than twenty specialised fusion experiments, of which the largest is the Joint European Torus (JET) in the UK. In addition, engineering R & D is carried out on a number of technological facilities. These various devices all contribute to the development of the scientific understanding and technology that is needed for future fusion power plants.

The Joint European Torus (JET)

Figure 1:A look inside the plasma vessel
of the Joint European Torus (JET).
JET is located in Culham, GB.

JET, based in Culham, Great Britain, is the central research facility of the European Fusion Programme, and the largest and most successful fusion experiment in the world. JET was approved in 1974, began operations in 1983, and met its planned operational goals on schedule in 1990. Since the beginning of 2000, the JET experimental programme has been managed under EFDA. The UK Atomic Energy Authority (UKAEA) has been contracted to maintain and operate the facility, with experimental work being carried out by visiting teams of scientists from all the Associations working on the fusion programme.

The JET machine is a large tokamak device, approximately 15 metres in diameter and 12 metres high. At the heart of the machine there is a toroidal vacuum vessel of almost 6 metres diameter with a "D"-shaped cross-section which measures 2.5 by 4.2 metres.

Figure 2:JET tokamak © Image: JET

The main toroidal magnetic field is provided by 32 "D"-shaped coils surrounding the vacuum vessel. This field combines with the field produced by the current flowing through the plasma to form the basic magnetic confinement of the plasma in the vessel. Additional coils positioned around the tight-fitting mechanical shell are used to shape and position the plasma.

The tokamak is surrounded by a very extensive range of diagnostic systems that are able to measure a wide variety of plasma properties. These include magnetic based measurements of the plasma position, shape and current; measurement of plasma density and temperature; a variety of spectroscopic measurements (from visible to X-ray and neutron spectrometry); video imaging of the plasma and many other techniques.

Figure 3:JET Diagnostics © Image: JET

JET has produced significant fusion power in deuterium/tritium plasmas - up to 16 MW - in the short pulses characteristic of existing experimental devices. "Break-even" conditions, where the fusion output power equals the external input power required to heat the plasma, were almost reached. Moreover, JET has demonstrated that fusion devices can be operated safely with tritium fuel and that radioactive structures can be maintenanced and modified using remote handling techniques.

JET, which is a unique tool in the preparative work for ITER, is the closest to ITER in size, shape and plasma parameters. It is currently the only device in the world capable of operating with the fuel mixture that will be used in a commercial fusion power station, and is able to make technological tests on advanced systems - including heating and control systems, plasma facing components, and remote handling - in an "ITER relevant" environment.

Physics devices

Figure 4:Plasma inside the ASDEX
Upgrade tokamak
© Image: IPP

The small and medium-sized tokamaks in the European fusion programme encompass a wide range of configurations, size, shape and heating systems and are able to explore a variety of operational scenarios. This provides for a greatly increased database on which to base decisions on future designs of fusion devices.

The ASDEX Upgrade device is run by the German IPP Association. Based at Garching in Germany, it is investigating ITER-relevant divertors, plasma-wall interactions, and advanced operational scenarios.

The IPP.CR Association in Prague, Czech Republic, hosts the CASTOR device which is studying how to control fluctuations in the confined plasma, new methods for increasing plasma current, and diagnostic developments.

COMPASS-D is a compact spherical tokamak, constructed at Culham in the UK, with, originally, a circular cross-section, but now with a 'D'-shaped vessel. The device was operated by UKAEA for plasma stability studies.

The spherical tokamak MAST is also situated at Culham in the UK and run by UKAEA. This device is continuing the work on spherical tokamaks initiated by the START device at Culham, and which may be able to offer operational benefits to commercial plants.

FTU is a high magnetic field, high plasma density, high current tokamak operated by ENEA at Frascati in Italy. The device also investigates new radio- frequency plasma heating techniques using electron cyclotron technology.

The Portuguese Association IST operates ISTTOK a small tokamak in Lisbon. The device is engaged in fundamental physics studies, in particular developing 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 CRPP Association in Lausanne, Switzerland.

TEXTOR-94, a tokamak run by three Associations (FZJ from Germany, FOM from The Netherlands and The Belgian State Association), is sited in Jülich in Germany. Experiments on the device include wall interactions, plasma confinement under additional heating, and testing of new divertor technology.

Figure 5:Tore Supra is designed to work with
long duration plasma burn © Image: CEA

The French Association CEA runs one of the largest tokamaks operating today: the TORE SUPRA at Cadarache in France. This device is the first tokamak to use a series of superconducting coils to generate a permenant magnetic field. It has a circular plasma cross-section and has the capability to run 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.

Stellarators offer an intrinsic potential for steady-state, continuous operation. A Stellarator project is currently in operation (TJ-II in Madrid, Spain) with construction of a further device (Wendelstein 7-X in Greifswald, Germany) well under way.

Figure 6:The TJ-II experiment
© Image: CIEMAT

TJ-II is a highly flexible device with helical magnetic axis constructed at Madrid, Spain by the CIEMAT Association. It works on novel confinement and high efficiency operations.

The recently closed Wendelstein 7-AS device at Garching in Germany was operated by IPP. It studied plasma behaviour in a modular Stellarator design and tested an island divertor concept. It was also used as the basis for the larger advanced Stellarator Wendelstein 7-X designed at Garching and being constructed at Greifswald in Germany.

Two Reversed Field Pinch devices (RFP) contribute to studies on achieving high-performance operation and controlling plasma modes.

RFX, run by the Italian ENEA Association at Padova in Italy, is investigating toroidal confinement and transport whilst evaluating future prospects for RFP technology.

EXTRAP-T2, sited at Stockholm in Sweden and run by the NFR Association, supports RFX by looking, in particular, at wall stabilisation: a situation where changes to some electrical properties of the chamber wall can affect the stabilisation of the plasma.

Technology devices

Figure 7:Gas handling system
at JET © Image: JET

There are also a large number of other devices devoted to technological aspects of the fusion programme. These include:

MANTIS, which is a neutral beam test bed and FE 200, a thermal fatigue test facility using a 200kW electron gun, run by CEA in France and Framatome respectively.
SULTAN - a test facility for superconductor and joint samples, run by CRPP in Switzerland.
The divertor test and refurbishment platforms are run by ENEA in Italy.
The TOSKA facility for testing large superconducting coils and the Tritium laboratory is being operated by FZK in Germany.
The Active Gas Handling System is based at JET in the UK.