Objective
The decision of the Council of Ministers states that the JET Joint Undertaking's mandate is to:
"...construct operate and exploit as part of the Euratom fusion programme and for the benefit of its participants in this programme, a large torus facility of tokamak-type and its auxiliary facilities in order to extend the parameter range applicable to controlled thermonuclear fusion experiments up to conditions close to those needed in a thermonuclear reactor."
The principal objective of JET is to enable the essential requirements of a tokamak reactor to be defined. To do this, a plasma approaching reactor condition must be created and studied.
There are four main areas of work:
- the dusty of scaling of plasma behaviour as parameters approach the reactor range;
- the study of plasma-wall interaction in these conditions;
- the study of plasma heating;
- the study of alpha-particle production, confinement and consequent plasma heating.
In addition, JET is pioneering two of the key technologies that will be required in subsequent fusion reactors. These are the use of tritium and the application of remote maintenance and repair techniques.
Joint European torus (JET) plasmas display most of the characteristics to be expected from a reactor plasma. Densities up to nbar Rq(cyl)/B(t) = 33 m{-2}T{-1} with nbar = 1.4E20 m{-3} are obtained and appear to be limited by edge radiation loss processes. Beta values up to 5.5% equal to the Troyon limit value at B(t) = 0.9 T are obtained. Strongly peaked ion temperatures above 28 keV with T(i)(0)/Tbar(i) of about 6 are obtained in low density H-mode discharges. T(i)(0) and T(e)(0) are simultaneously in excess of 10 keV in higher density discharges.
Energetic (MeV) particles both with isotropic and with transverse velocities are contained for a least 2 s while they slow down by apparently classical processes. This is adequately long enough for energy transfer in a reacting plasma, where even better confinement should be expected given the increased size and plasma current.
The duration of fusion relevant conditions in JET H-mode discharges is at present limited to less than 1 s by the influx of carbon impurities form localized plasma target plate interaction. 100 kW of fusion power in charged particles has been produced in deuterium (helium-3) plasmas. n(D)(0)T(i)(0)tau(E) values of 7E20 to 9E20 m{-3)keVs with Q(DD) greater than 2E-3 have been produced in deuterium-deuterium plasmas. For the same plasma parameters in a deuterium-tritium plasma this is equivalent to Q(DT) of about 0.7 P(alpha) of about 2.4 MW, P(fusion) of about 12 MW with P(alpha)/P(Loss) above 20%.
Recent experiments in the joint European torus (JET) using beryllium, both as a limiter material and for gettering, have resulted in a significant extension of JET operating regimes. The power range of edge localized mode (elm) free H-mode experiments has been extended, by almost a factor of 2, and the previously observed confinement behaviour confirmed. In addition, an ion cyclotron resonance heating (ICRH) generated H-mode has been obtained. The confinement properties of both ICRH and combined heating H-modes are now equivalent to neutral beam injection (NBI) H-modes. Improved control of particle recycling due to the use of beryllium has also improved the performance of limiter discharges. Sawtooth free L-mode discharges with Te(0) of about 12 keV have been produced at 5MA and, at 3 MA, a range of discharges with peaked density and temperature profiles have been obtained both on the inner wall and on the outside belt limiters.
Analysis of the local heat flux has confirmed that in many JET regimes ion transport dominates electron transport in the outer half of the plasma. In regimes with peaked density and temperature profiles, however, ion heat diffusivity falls to near neoclassical values in the plasma centre. Quantitative comparison of transport derived from theories of ion temperature gradient driven turbulence with that observed experimentally has found that the transport predicted was too strong. A study of ion heat and momentum diffusivities over many discharges from several JET regimes concluded that the values are strongly correlated. Investigation of electron transport by perturbative means has confirmed previous results and, in addition, obtained evidence of direct coupling between heat and particle transport. Discharges with beta(phi) values close to the Troyon limit exhibit substantial magnetohydrodynamic activity which can modulate or reduce beta(phi). Studies of the sawtooth instability have confirmed the internal self consistency of several of th e critical diagnostic measurements. A quantitative assessment of sawtooth stabilization experiments supports the idea that energetic ions are central to the stabilization mechanism.
Beryllium was examined as material for walls and limiters. Studies were carried out with initially a thin layer of beryllium evaporated onto the walls of the machine. Later, in addition the graphite material of the belt limiter was exchanged for beryllium. The use of this material was generally beneficial for the plasma behaviour. Combined with a reduction in the oxygen content, strong pumping of hydrogen isotopes was found which allowed the joint European torus project (JET) to widen considerably the operational space with respect to ion temperatures, densities and plasma purity. Higher plasma purity levels and higher density operation were achieved with beryllium limiters in comparison with graphite ones. These improvements resulted mainly from the elimination of oxygen and the strong wall pumping capability of beryllium.
Despite very accurate alignment of the belt limiter, hot spots and local melting of the limiter surface was found. Consequently, plasma facing components must be designed in such a way that alignment is not critical and that the heat load is distributed evenly.
The H-mode performance was considerably improved compared to earlier experiments, but this mode of operation was prevented from reaching its full potential by a strong influx of carbon 1 s after the commencement of the high power heating. Calculations show that an equivalent Q(DT) value above unity would have been achieved by delaying the carbon bloom by another second.
Improvements to the performance of joint European torus (JET) limiter plamas have been made by increasing the plasma current, by suppressing sawteeth, by peaking the density profile and by using beryllium evaporation and employing beryllium limiters. Central electron temperatures of about 12 keV, ion temperatures of about 18 keV, central electron densities of about 2.3E20 m{-3} have been demonstrated in limiter plasmas with currents 3 to 5 MA. A 7 MA discharge with q(psi) of about 3.2 has been demonstrated.
The experiments have shown that the edge localized mode (ELM) free H-mode is a transient plasma regime. The analysis of the results of impurities injection experiments and of the spectroscopical data suggest that if one could control the plasma density and drastically reduce the influx of impurities, the H-mode could reach a steady state transport equilibrium.
The H-mode is achieved in a configuration with a magnetic separatrix, which can be marginally limiting on the x-point dump plates or on the inner wall protection plates. A parameter related to the shear at the plasma edge, for this marginally limiting configuration, could be the ratio between q(psi) and q(cyl). H-mode has been achieved with q(psi) larger or equal to 7 to 8 while q(cyl) was 2 to 3.
In the H-mode the global energy confinement time shows an improvement of approximately a factor of 2 over the limiter L-mode independently of the heating method and in the range of additional power up to 25 MW.
The radiation collapse, which terminates the H-mode, determines the values of the density limit, which scales with square root of the input power. An improvement in plasma purity and reduction of fuel dilution has been achieved with beryllium gettering of graphite tiles. In low density ohmic target, and largely with ion cyclotron resonance frequency (ICRF) heating it as been possible to achieve transient stabilization of sawteeth during the H-mode. With neutral beams (NB) injection in low density deuterium plasmas a hot ion plasma has been created during the H-mode. In these discharges the highest plasma thermonuclear reactivities have been obtained.
The study of energy and particle confinement in joint European torus (JET) plasmas has been performed by means of various and interpretive and predictive techniques, concentrating on local rather than on global analysis. Electron and ion energy transport and particle transport are studied and their relationship is examined. Comparisons with the predictions of theoretical models are made with special attention to transport coefficients derived from the theory of ion temperature gradient driven turbulence and to the critical electron temperature transport model. The analysis showed that no pure theoretical mode was sufficiently correct or complete enough to explain all features exhibited by the wide variety of plasma regimes found in JET. Time dependent simulations of a variety of phenomena evolving on different time scales indicated a correlation between energy and particle transport. In particular, D chi(e) as normally defined in interpretive codes was larger than D{P}/chi(e) {P} from pulse propagation analysis. Neoclassical theory may be relevant for impurity transport under various circumstances. These results are generally consistent with the picture of the energy and particle transport assumed by the critical electron temperature model. This model, which has been successfully tested in a variety of JET plasma conditions, is a good candidate for predicting tokamak plasma performance.
The density limits in JET discharges with graphite (carbon), beryllium gettered graphite and beryllium limiters were examined. There is a clear improvement in the case of beryllium limiters. The beryllium gettered phase showed no increase in the gas fuelled density limit, except with ion cyclotron resonance heating (ICRH), but the limit changed character. During asymmetric edge radiation (MARFE) formation, any further increase in density was prevented, leading to a soft density limit. The soft density limit was a function of input power and impurity content with a weak dependence on q Helium and pellet fuelled discharges exceeded the gas fuelled global density limits, but essentially had the same edge limit. High beta operation in low-B double null (DH) X-point configurations with beryllium gettered carbon target plates was studied. The Troyon limit was reached during H-mode discharges and toroidal beta values of 5.5% were obtained. At high beta, the sawteeth were modified and characterized by very rapid heat waves and fishbone like pre and postcursors with strongly ballooning character.
The entrenchment of impurities in a forced flow of plasma towards the target plates is a candidate concept for impurity control, and is the focus of the pumped divertor planned for the joint European torus (JET). Analysis shows that the effectiveness of the friction force in overcoming the ion thermal force is strongly reduced at high ion temperature in the scrape off layer (SOL), which occurs at high ion power flows (Pi). The hydrogenic flow required for effective impurity retention increases approximately as Pi{5/7}. At high SOL densities, recycling at the target plates is very high, and this natural recycling flux is sufficient to retain the impurities effectively. At moderate and low SOL densities, however, the natural recycling flux must be augmented by external recirculation to retain impurities. The fraction of the target flux needed for impurity retention increases as density decreases, but decreases with decreasing ion power flow. Pumping the required external recirculation flows is likely to be possible only at rather high edge densities, approaching about 1E20 m{-3}. The situation improves at lower Pi. Operation with clean plasmas is likely, therefore, only at high densities. This may have severe consequences, eg for current drive. First steps towards validating these complex SOL model have been taken by establishing consistency with observed edge data in JET x-point. In particular, the predicted conditions Ti above Te in the SOL, and divertor temperatures Ted in the range 10 to 20 eV at high edge density have been observed.
The joint European torus (JET) has been successfully operated with beryllium evaporation onto graphite surfaces and with beryllium limiters, ion cyclotron resonance heating (ICRH) Faraday shields and X-point target tiles. Although hot spots and surface melting have been observed on the beryllium components, the plasmas can be optimized so that improved performance over the graphite phase is obtained. Beryllium has shown itself to be a viable replacement for graphite as a plasma facing material. The use of beryllium has increased the density limit, significantly reduced deconditioning following disruptions, allowed heavy gas fuelling for impurity control, reduced the impurity influx from the ICRH antennae so that ICRH only H modes were possible for the first time and permitted hot ion plasmas on the outer limiters. However, neither material can satisfactorily match the requirements for divertor target tiles, because of the problem of blooms, unless most of the power is radiated from the target plasma.
Impurity influxes in joint European torus (JET) discharges due to ion cyclotron resonance heating (ICRH) have been reduced to insignificant levels. This has allowed high quality H-modes to be produced with ICRH alone and has enhanced the density limit which is now the same as the neutral beam injection (NBI) limit. Improvements in the deuterium fuel fraction has led to the generation of 100 kW of nonthermal helium-3 and deuterium fusion power. Alpha particle simulations using MeV ions created by ICRH show classical energy loss and suggest that alpha heating in a reactor will be highly efficient. A clear demonstration of transit time magnetic pumping (TTMP) damping of the fast wave in high beta plasmas has been achieved. A broadband ICRH system is proposed which allows fast wave current drive and central ion heating for burn control and ignition.
Sawtooth relaxations control the central plasma parameters in most joint European torus (JET) discharges and, as a result, have a significant influence on fusion performance. Several techniques have been developed for the stabilization of sawteeth in various JET operating regimes. Under some circumstances, sawtooth suppression appears to be associated with the central safety factor rising above unity. However, in other cases it remains significantly below unity and it is thought that stabilization is due to the influence of a population of energetic particles. Detailed measurements of the evolution of the current profile in JET during sawteeth have confirmed that the central safety factor remains below unity throughout the sawtooth cycle and thus the trigger for the sawtooth collapse remains problematic. Further diagnostic investigations of the sawtooth collapse have shown that the collapse takes the form of a convective plasma flow, but the cause of the resultant energy transport is not understood.
2 mechanisms for the self sustainment of magnetic islands have been studied in cylindrical geometry. The first one is based on the different behaviour of electrons and ions in the presence of the islands as a results of their different Larmor radii. This mechanism could maintain magnetic turbulence resulting from a mixture between islands of the size of the ion Larmor radius and chaotic regions. The second mechanism is a pseudogravity, used as a simple analogue for pressure gradient and field curvature modes. It could sustain islands much larger than the ion Larmor radius.
It is found that resistive magnetohydrodynamics is inappropriate to describe tokamak sawtooth reconnection. It is found that under usual conditions the electric field resulting from flux reconnection is very large. The resulting acceleration of the electrons is such that reconnection is determined by electron inertia rather than electron collisions.
Radio frequency techniques in the ion cyclotron (ICR), lower hybrid (LH) and electron cyclotron (ECR) ranges of frequency have proved very effective. In the joint European torus (JET), 18 MW of ICR power was achieved for a few seconds and 6 MW for 20 s. The largest fusion yield was 100 kW from helium-3 and deuterium reaction with energetic ions created from ion cyclotron resonance heating (ICRH). The greatest electron temperature was 12 keV.
In JT-60, 9 MW of LH power was achieved for a few second, and the largest noninductive current drive plasma current, of 2 MA. In T-10, 2.2 MW of ECR power wasachieved for 0.2 s and in DIII, 1 MW for about 1 s. The longest moninductive current drive plasma current lasted for more than 1 hour, at 35 kA in Triam-1M.
The use of beryllium as a first wall material has substantially improved the plasma purity and as a consequence a record fusion product 9E20 m{-3}keVs has been achieved. Impurity influxes prevent the achievement of higher plasma parameters and steady state conditions. A new divertor configuration has been proposed to study impurity control, fuelling and exhaust, in conditions relevant for the next generation of machines. The latest design features a multicoil configuration which gives substantive operational flexibility.
The performance of a plasma divertor utilizing toroidal X-point target tiles was investigated. The choice of beryllium tiles for the lower target and carbon fibre composite tiles for the upper target permitted a careful study of the relative merits of the 2 materials. Two designs of carbon fibre composite tiles were used. High performance, but ion H-mode discharges with both sets of tiles were terminated by carbon blooms, but there were significant differences between the threshold energies and the nature of the blooms. In both cases, the neutron production rate remained high for a few tenths of a second after the influence of carbon started to increase at the plasma edge.
Beryllium tiles with sloping edges were used at the lower divertor. At modest heating powers, there was little difference in the confinement properties of medium density H-modes compared to discharges on the carbon tiles, but the ion temperature was higher with beryllium due to reduced radiation cooling. At higher heating powers (above 10 MW) and lower densities, the beryllium tiles were vunerable to melting, which resulted in the rapid distribution of the discharge due to a beryllium was about a half of that obtained with carbon. Beryllium tiles could withstand higher heating powers when combined with strong gas puffing to increase the radiated power and cool the plasma edge.
Test impurities introduced into the plasma showed better confinement during H-modes than during L-modes. This suggested that the H-mode might be a disadvantage in a reactors, where impurities and helium ash could be retained, leading to increased fuel dilution in the plasma centre.
Experiments have improved understanding of the conditions under which disruptions occur and of the evolution of the plasma towards a major disruption. Early detection of a helical nonrotating instability has enabled measures to be taken to minimize the impact of the resultant disruption on the torus. Detection triggers a rapid ramp down in driving current to zero, postphoning the loss of vertical stability until much later in the current decay, with consequent reduction in forces by an order of magnitude.
Solid pellets of hydrogen or deuterium have been injected into the plasma centre to optimize the central density, while minimizing the edge density. A central density of 4E20 m{-3} was achieved by strongly peaking the density profile using a sequence of 4 mm solid deuterium pellets injected at intervals throughout the current rise phase of an X-point discharge.
Techniques have been developed for the suppression of sawtooth instabilities, by the injection of lower hybrid (LH) radio frequency (RF) waves or neutral beams into the plasma. Phasing the ion cyclotron resonance frequency (ICRF) power with the resonance on the high field side of the magnetic axis controlled the sawtooth period and the rate of ejection of fast particles from the plasma centre, suggesting a method for burn control in a reactor.
ICRF heating has been used in the pellet enhanced phase of operation to reheat a dense core with a power deposition profile peaked on axis. Peak ion temperatures ranging from 16 to 18 keV and peak electron temperature between 12 an 14 keV have been achieved.
The lower hybrid current drive system (LHCD) has been used to sustain plasma current, in combination with ICRF power. The current drive efficiency reached nRI/P = 0.4E20 m{-2} A/W.
Neutral beam injection was used to heat the plasma using 14 beams of deuterium, and to inject tritium, ensuring that the tritium reached the centre of the discharge.
Energy confinement times of up to 1.8 s have been attached in joint European torus (JET) ohmically heated discharges, but with temperatures too low for fusion. Experiments were conducted to investigate the correlation of energy confinement time with values of plasma parameters. Confinement was found to depend on total plasma current, its distribution, and internal plasma inductance.
Experiments have confirmed that energy confinement time is not dependent on the square root of atomic mass, which scaling laws predict.
Neutral beam counter injection experiments established that the power threshold for the H-mode was from 10 to 20% lower than for coinjection. With counter injection, frequent, small amplitude sawteeth were observed. During some H-modes, edge localized mode of instability appeared at the plasma edge, leading to some loss of confinement. Under certain conditions, a steady state H-mode was achieved in which particle and impurity influxes were controlled by regular edge localized modes, produced by gas puffing from the midplane. A maximum duration of 18 s has been achieved at 2 MA and 2.3T.
The relative plasma performance and plasma purity of successive cycles of alternating current (AC) operation of the joint European tours (JET) were examined. No significant difference between the first and second phase of as AC discharge was discovered.
Attempts to demonstrate long pulse operation included development of discharges up to 1 minute in duration at 2 MA and 1.9T. However, the particle behaviour did not approach the desired steady state.
An investigation of fusion performance was carried out on a 3 to 4 MA double null X-point configuration, with input power up to 30 MW. Ion cyclotron resonance frequency, neutral beam injection and combined heating were used with bothdeuterium and helium-3 beams. Stored plasma energy of 12.7 MT was obtained with a trial input power of 20 MW.
A preliminary deuterium and tritium experiment was carried out using this configuration. In 2 fuel power pulses, tritium was introduced from 2 neutral beam sources. Each pulse produced fusion power in excess of 1.5 MW.
With beryllium tiles for the lower target and carbon fibre composite tiles for the upper target, comparison of the materials and of tile design was possible. Carbon fibre composite tiles gave better fusion yield in low density hot ion H-modes.
Steady state H-modes, with plasma parameters held constant by edge localized modes, were produced in a 2 MA plasma for a maximum duration of 18 s. Sawteeth were stabilized for periods up to 2 s by varying the phase of the ion cyclotron resonance frequency (ICRF) system antennae, resulting in a significantly peaked temperature profile.
Using ICRF power and lower hybrid current drive to enhance current drive, a discharge at 2 MA plasma current was maintained and heated for a duration of 1 minute.
Reliable operation and additional heating of a 2-cycle plasma was demonstrated at 2 MA and plasma parameters were found to be very similar in the 2 cycles.
The hot ion mode was further enhanced to produce a neutron yield in a deuterium plasma of 4.3E16 s{-1}. The fusion product was 9.2E20 m{-3} keVs, which corresponded to an equivalent fusion amplification factor of 1.14. This exceeded break even conditions and was within a factor of 5 to 8 of that required in a reactor.
Tritium was introduced into hot ion H-mode discharges, resulting in total fusion releases of 1.7 MW peak power and 2 MJ energy.
History
Joint European Torus (JET), Joint Undertaking was set up under Council Decision (78/471/Euratom) on 30 May 1978. The JET Project has been prolonged until the end of 1992, 1994 and then to the end of 1996 (Council Decision 91/677/Euratom). A further extension to the end of 1999 has been proposed.
Present scientific and technical programme
The Mandate of the JET Joint Undertaking is to : "construct, operate and exploit as part of the Euratom fusion programme and for the benefit of its participants, a large torus facility of tokamak-type and its auxiliary facilities in order to extend the parameter range applicable to controlled thermonuclear fusion experiments up to conditions close to those needed in a thermonuclear reactor". The original principal objective of JET was to enable the essential requirement of a tokamak reactor to be defined. To do this, a plasma approaching reactor conditions must be created and studied. Subsequently the extension of the Project to 1996 brought into the programme an objective to establish the effective control of impurities in conditions close to the Next-Step.
There are four areas of work : the study of scaling of plasma behaviour as parameters approach the reactor range; the study of plasma-wall interaction in these conditions; the study of plasma heating; the study of alpha-particle production, confinement and consequent plasma heating. The major focus of this work is now the study of divertor physics and the technology of divertor operations directed to the provision of key information for design of a Next Step fusion device and ultimate commercial systems. JET is also developing two of the new technologies that will be required in subsequent fusion power stations. These are the use of tritium and the application of remote maintenance and repair techniques.
In November 1991, for the first time ever, controlled fusion power in the megawatt range was produced for 2 seconds. In 1992/93 there was a major shutdown for the installation of four divertor coils and the Mark I divertor structure, as well as the upgrading of the current drive and other systems. After the initial phase aimed at bringing the machine back into operation, most of the 1994/95 experimental campaign is to be executed by three Task Forces for divertor assessment, high performance, and tokamak concept improvements. The programme plan to 1999 calls for successive modification and operation of the divertor into forms and operating regimes gradually approaching those thought necessary for a next step device.
Staff
Professionals : 203
Technical support staff : 90
Administrative support staff : 60
Contract staff : varies in range 400-500
Yearly budget : pounds sterling 80 MioECU
Management structure
Head of Research Unit : M. KEILHACKER, Director of JET
Joint Undertaking : Director : M. Keilhacker
Deputy Director : A. Gibson
Associate Directors : J. Jacquinot, G.W. O'Hara
Collaboration with other institutions
Collaborates with all Euratom Associations and other fusion laboratories worldwide (incl. IEA Tripartite Agreement on Large Tokamaks). JET participates extesively in ITER Committees.
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Coordinator
OX14 3EA Abingdon
United Kingdom
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