Joint European Torus

The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)

On the basis of an analysis of the ITER L-mode energy confinement database, two new scaling expressions for tokamak L-mode energy confinement are proposed, namely a power law scaling and an offset-linear scaling. The analysis indicates that the present multiplicity of scaling expressions for the energy confinement time τE in tokamaks (Goldston, Kaye, Odajima-Shimomura, Rebut-Lallia, etc.) is due both to the lack of variation of a key parameter combination in the database, fs = 0.32 R a−0.75 k0.5 ∼ A a0.25k0.5, and to variations in the dependence of τE on the physical parameters among the different tokamaks in the database. By combining multiples of fs and another factor, fq = 1.56 a2 kB/RIp = qeng/3.2, which partially reflects the tokamak to tokamak variation of the dependence of τE on q and therefore implicitly the dependence of τE on Ip and ne, the two proposed confinement scaling expressions can be transformed to forms very close to most of the common scaling expressions. To reduce the multiplicity of the scalings for energy confinement, the database must be improved by adding new data with significant variations in fs, and the physical reasons for the tokamak to tokamak variation of some of the dependences of the energy confinement time on tokamak parameters must be clarified.

Effects of the plasma boundary can have a substantial influence on the behaviour of the entire plasma in tokamaks. Progress in the field, particularly that over the last decade, is reviewed, with emphasis on experimental observation. Simple modelling for interpretation is also included.

In JET, both high density and low-q operation are limited by disruptions. The density limit disruptions are caused initially by impurity radiation. This causes a contraction of the plasma temperature profile and leads to an MHD unstable configuration. There is evidence of magnetic island formation resulting in minor disruptions. After several minor disruptions, a major disruption with a rapid energy quench occurs. This event takes place in two stages. In the first stage there is a loss of energy from the central region. In the second stage there is a more rapid drop to a very low temperature, apparently due to a dramatic increase in impurity radiation. The final current decay takes place in the resulting cold plasma. During the growth of the MHD instability the initially rotating mode is brought to rest. This mode locking is believed to be due to an electromagnetic interaction with the vacuum vessel and external magnetic field asymmetries. The low-q disruptions are remarkable for the precision with which they occur at qψ = 2. These disruptions do not have extended precursors or minor disruptions. The instability grows and locks rapidly. The energy quench and current decay are generally similar to those of the density limit.

<span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">Describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixture. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to produce more than one megawatt of fusion power in a controlled way; (ii) to validate transport codes and provide a basis for accurately predicting the performance of deuterium-tritium plasmas from measurements made in deuterium plasmas; (iii) to determine tritium retention in the torus systems and to establish the effectiveness of discharge cleaning techniques for tritium removal; (iv) to demonstrate the technology related to tritium usage; and (v) to establish safe procedures for handling tritium in compliance with the regulatory requirements. A single-null X-point magnetic configuration, diverted onto the upper carbon target, with reversed toroidal magnetic field was chosen. Deuterium plasmas were heated by high power, long duration deuterium neutral beams from fourteen sources and fuelled also by up to two neutral beam sources injecting tritium. The results from three of these high performance hot ion H-mode discharges are described: a high performance pure deuterium discharge; a deuterium-tritium discharge with a 1% mixture of tritium fed to one neutral beam source; and a deuterium-tritium discharge with 100% tritium fed to two neutral beam sources. The TRANSP code was used to check the internal consistency of the measured data and to determine the origin of the measured neutron fluxes. In the best deuterium-tritium discharge, the tritium concentration was about 11% at the time of peak performance, when the total neutron emission rate was 6.0 </span><strong style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">×</strong><span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px"> 10</span><sup style="vertical-align: baseline; position: relative; bottom: 0.5em; font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">17</sup><span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px"> neutrons/s. The integrated total neutron yield over the high power phase, which lasted about 2 s, was 7.2 </span><strong style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">×</strong><span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px"> 10</span><sup style="vertical-align: baseline; position: relative; bottom: 0.5em; font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">17</sup><span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px"> neutrons, with an accuracy of ±7%. The actual fusion amplification factor, Q</span><sub style="vertical-align: baseline; position: relative; top: 0.25em; font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px">DT</sub><span style="font-family: Arial, Helvetica, Verdana, sans-serif; line-height: 16.200000762939453px"> was about 0.15</span>

The stability of the ion temperature gradient modes is investigated using the kinetic ion response without expansions in ωD/ω. A systematic parameter study is carried out using a low-beta circular flux surface equilibrium in order to determine the stability boundaries in ηi vs εn space (ηi=d ln Ti/ d ln n, εn=Ln/R). Particular attention is devoted to the consequences of the presence of these modes for anomalous ion transport.

The scaling of the energy confinement in H mode plasmas with different hydrogenic isotopes (hydrogen, deuterium, DT and tritium) is investigated in JET. For ELM-free H modes the thermal energy confinement time τth is found to decrease weakly with the isotope mass (τth ∼M-0.25±0.22), whilst in ELMy H modes the energy confinement time shows practically no mass dependence (τth ∼M0.03±0.1). Detailed local transport analysis of the ELMy H mode plasmas reveals that the confinement in the edge region increases strongly with the isotope mass, whereas the confinement in the core region decreases with mass (τthcore ∝ M-0.16), in approximate agreement with theoretical models of the gyro-Bohm type (τgB ∼M-0.2).

The experimental characteristics of divertor detachment in the JET tokamak with the Mark I pumped divertor are presented for ohmic, L mode and ELMy H mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and Dα emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

High fusion power experiments using DT mixtures in ELM-free H mode and\noptimized shear regimes in JET are reported. A fusion power of 16.1 MW\nhas been produced in an ELM-free H mode at 4.2 MA/3.6 T. The transient\nvalue of the fusion amplification factor was 0.95+/-0.17, consistent\nwith the high value of\nnDT(0)τEdiaTi(0) = 8.7\n× 1020+/-20% m-3 s keV, and was maintained\nfor about half an energy confinement time until excessive edge pressure\ngradients resulted in discharge termination by MHD instabilities. The\nratio of DD to DT fusion powers (from separate but otherwise similar\ndischarges) showed the expected factor of 210, validating DD projections\nof DT performance for similar pressure profiles and good plasma mixture\ncontrol, which was achieved by loading the vessel walls with the\nappropriate DT mix. Magnetic fluctuation spectra showed no evidence of\nAlfvénic instabilities driven by alpha particles, in agreement\nwith theoretical model calculations. Alpha particle heating has been\nunambiguously observed, its effect being separated successfully from\npossible isotope effects on energy confinement by varying the tritium\nconcentration in otherwise similar discharges. The scan showed that\nthere was no, or at most a very weak, isotope effect on the energy\nconfinement time. The highest electron temperature was clearly\ncorrelated with the maximum alpha particle heating power and the optimum\nDT mixture; the maximum increase was 1.3+/-0.23 keV with 1.3 MW of alpha\nparticle heating power, consistent with classical expectations for alpha\nparticle confinement and heating. In the optimized shear regime, clear\ninternal transport barriers were established for the first time in DT,\nwith a power similar to that required in DD. The ion thermal\nconductivity in the plasma core approached neoclassical levels. Real\ntime power control maintained the plasma core close to limits set by\npressure gradient driven MHD instabilities, allowing 8.2 MW of DT fusion\npower with\nnDT(0)τEdiaTi(0) approx\n1021 m-3 s keV, even though full optimization was\nnot possible within the imposed neutron budget. In addition,\nquasi-steady-state discharges with simultaneous internal and edge\ntransport barriers have been produced with high confinement and a fusion\npower of up to 7 MW these double barrier discharges show a great\npotential for steady state operation. © 1999, Euratom

The present model of sawtooth oscillations does not appear to be consistent with experimental observations on JET. An alternative theory is proposed, offering possible explanations for the basic aspects of the observed behaviour.

Abstract The JET edge plasma code EDGE2D/U is designed to simulate divertor configurations. We describe here the physical model and the numerical techniques used.

Comparisons between time dependent simulations and experiments in ion cyclotron heated plasmas in the JET tokamak have been made. A time dependent code, PION-T, has been used to simulate the heating. The scenario that has been analysed is minority heating of hydrogen in a deuterium plasma. Two measured quantities have been compared with the calculations, namely the anisotropic plasma energy content and the 2.4 MeV neutron flux from (D,D) reactions. Two versions of the PION-T code have been used, one with zero banana width and one in which a simplified model for taking the finite width of drift orbits into account is employed. The zero banana width calculations yield good agreement between measurements and calculations for relatively low power levels only. However, with the second version of the code, taking the finite width of the drift orbits into account, good agreement can also be obtained for higher power levels. The non-thermal neutron yield caused by second harmonic absorption by the deuterium is simulated and agreement is found provided the hydrogen concentration, for which no reliable measurements are available, is suitably chosen. Finally, by studying the rapid changes in fast ion energy content in connection with sawtooth instabilities, it is found that about 40% of the fast ions are expelled outside the q=1 surface and that prompt losses are negligible

Experiments in the JET tokamak with additional heating power (ion cyclotron resonance heating and/or neutral beam injection) above 5 MW show that the plasma can undergo a transition to a new regime. In this regime, the sawtooth instability is suppressed for periods up to 1.6 s and the level of long-wavelength, coherent MHD activity is very low. An improvement in the global energy confinement time of up to 20% is observed. Possible mechanisms for the stabilization of the $m=1$ instability and the implications for the near-ignition regime are discussed.

Collisionless magnetic reconnection in regimes where the mode structure is characterized by global convection cells is found to exhibit a quasiexplosive time behavior in the early nonlinear stage where the fluid displacement is smaller than the equilibrium scale length. This process is accompanied by the formation of a current density sublayer narrower than the skin depth. This sublayer keeps shrinking with time.

The theory of non-linear tearing modes has been extended to include the effects of rotation and interaction with a resistive wall. The time dependent solution of the resulting equations shows a spatial locking of the mode similar to that observed in tokamak experiments. Time dependent numerical calculations show that the model can explain in a qualitative way the overall frequency evolution of mode locking and the radial magnetic field behaviour observed on JET.

The rapid collapse of a sawtooth oscillation in the JET tokamak has been observed in detail on a fast time scale. Tomographic reconstruction of data from two x-ray cameras and electron-cyclotron-emission temperature profiles show that during the sawtooth collapse the central hot region is rapidly (in approximately 100 \ensuremath{\mu}s) displaced off axis with an $m=1$ component and is then redistributed around a surface of constant minor radius. The theoretical implications of the measurements are discussed.

Preprint of an invited review paper presented to the 18th EPS Conf. on Controlled Fusion and Plasma Physics, Berlin, 3-7 Jun 1991

Analysis of MHD activity in Pellet Enhanced Performance (PEP) pulses is used to determine the position of rational surfaces associated with the safety factor q. This gives evidence for negative shear in the central region of the plasma. The plasma equilibrium calculated from the measured q values yields a Shafranov shift in reasonable agreement with the experimental value of about 0.2 m. The corresponding current profile has two large off-axis maxima in agreement with the bootstrap current calculated from the electron temperature and density measurements. A transport simulation shows that the bootstrap current is driven by the steep density gradient, which results from improved confinement in the plasma core where the shear is negative. During the PEP phase, (m, n) = (1, 1) fast MHD events are correlated with collapses in the neutron rate. The dominant mode preceding these events usually is n = 3, whereas the mode following them is dominantly n = 2. Toroidal linear MHD stability calculations assuming a non-monotonic q-profile with an off-axis minimum decreasing from above 1 to below 1 describe this sequence of modes (n = 3, 1, 2), but always give a larger growth rate for the n = 1 mode than for the n = 2 mode. This large growth rate is due to the high central poloidal beta of 1.5 observed in the PEP pulses. Finally, a rotating (m, n) = (1, 1) mode is observed as a hot spot with a ballooning character on the low field side. The hot spot has some of the properties of a 'hot' island consistent with the presence of a region of negative shear.

Measurements of soft X-ray emission from the JET plasma have been analysed with tomographic reconstruction methods. Because there are two detector arrays, two-dimensional images of X-ray emissivity are obtained without having to resort to rotation models. Several algorithms are employed in order to get as much detail as possible in the images while keeping any guiding assumptions to a minimum. The data analysed so far have been used principally to study MHD instabilities, and illustrative examples of the sawtooth crash and of disruptions are described.

New experiments on JET, COMPASS-D and DIII-D have identified the critical scalings of error field sensitivity and harmonic content effects, enabling predictions to be made of the requirements for larger devices such as ITER. Thresholds are lowest at low density, a regime proposed for H mode access on ITER. Results suggest a moderate error field sensitivity (δB/B ≈ 10-4) for ITER, comparable with the size of its intrinsic error, although there are uncertainties in scaling behaviour. Other studies on COMPASS-D and DIII-D show that sideband harmonics to the (2, 1) component play an important role. Thus, a correction system for ITER will be important, with flexibility to correct sidebands desirable, possibly assisted by beam rotation. Such a system has been designed and is capable of reducing multiple harmonic error levels to ∼ 2 × 10-5.