[en] The study of momentum confinement is a valuable means of studying the mechanisms governing confinement in tokamak plasmas. A dedicated rotation experiment was conducted in TFTR, in September 1988, using the recently installed CHERS diagnostic. Shots at different values of plasma current, magnetic field, injected beam power and injected direction were made, to study the parametric dependence of local fluxes of momentum and energy. A study of momentum confinement is presented. Its purposes are to analyze the data of the TFTR rotation experiment, to document the results and to compare the experimental results with the predictions of neoclassical and anomalous momentum transport theories. Particular attention is devoted to the evaluation of the magnitude of the poloidal variation of densities and rotation frequencies that determine the magnitude of the gyroviscous momentum flux. We find that: (1) Ware's cold ions theory underpredicts the observed viscosity by a few orders of magnitude; (2) each of the anomalous theories considered predicted torque flows which show magnitudes, radial profiles and parametric dependencies on plasma parameters different from those of the experimental torque flow; (3) up-down poloidal asymmetries of density and rotation frequency, evaluated with a model which neglects heat flux and includes the effect of anomalous particle fluxes, are found to be much smaller than epsilon; and (4) the gyroviscous torque is at least one order of magnitude smaller than the experimental torque flow

[en] Unbalanced neutral beam injection, used to heat tokamak plasmas, causes the plasma to rotate. It is well documented that the observed momentum confinement time cannot be explained by the existing neoclassical theory of perpendicular viscous momentum transfer. Analysis of Impurity Study Experiment-B, Poloidal Divertor Experiment, and Princeton Large Torus PDX, and data based on the more recent gyroviscous theory of momentum confinement, however, showed good agreement between theory and experiment. In this paper, results from recent Joint European Torus (JET) and Tokamak Fusion Test Reactor (TFTR) experiments in which the various plasma parameters were measured with a good degree of accuracy are presented and shown to be in agreement with the prediction of the gyroviscous theory. 6 refs., 2 figs

[en] The current driven by neutral beam injection is calculated for three different reactor designs: TIBER II, ITER-US and INTOR. The sensitivity of the current drive efficiency to plasma and beam parameters is studied, and general conclusions are drawn. 12 refs., 8 figs., 1 tab

[en] Full text: Experiments on disruption mitigation are carried out nowadays on all large tokamaks to investigate the influence of the injected impurities on the development of the disruption. The optimal disruption mitigation requires that all three deleterious effects of disruptions, namely the localized thermal load, the energy carried by runaway electrons and the mechanical forces are minimized. The kind of gas, the rate of injection and the injected quantities differ considerably according to the type of gas injection system available, the volume of the vessel and the specific problems affecting the machine after disruption. In this framework, ASDEX Upgrade has been conducting experiments for years and is routinely employing the fast injection of neon for the plasma shut-down of disrupting plasmas and machine protection. The injection is triggered by the locked mode (LM) signal and leads to the onset of a mitigated disruption within 5 ms. The impurity gas is injected into the plasma with two electromagnetic valves. The valves have an opening and closing time of 2 ms and remain open for 4-5 ms; they have been mostly operating with neon gas at a reservoir pressure of 5 bar and have been typically injecting 180 mbar (4.5x10²¹ atoms) of gas. Dedicated experiments have been carried out with 10 and 15 bar reservoir pressure. This paper will (1) describe the phenomenology of mitigation, (2) the experimental condition in ASDEX Upgrade, (3) the plasma response to the injection of impurities, (4) discuss the understanding of the observed phenomena, and (5) outline the future plans for the development of the mitigation system on ASDEX Upgrade. (author)

[en] Disruption generate large thermal and mechanical stresses on the tokamak components. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/ predictive methods must be developed further. The study of disruptions on ASDEX Upgrade is focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon and learning to avoid it or to predict its occurrence (with neural networks, for example) and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows to dimension or reinforce the machine components to withstand these loads and to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions. (author)

[en] Plasma disruptions, i. e. the sudden loss of magnetic confinement, are unavoidable, at least occasionally, in present day and future tokamaks. The expected energy fluxes to the plasma facing components (PFCs) during disruptions in ITER lie in the range of tens of GW/m''2 for timescales of about a millisecond. Since high energy fluxes can cause severe damage to the PFCs, their design heavily depends on the spatial and temporal distribution of the energy fluxes during disruptions. We investigate the nature of power fluxes during the thermal quench phase of disruptions by means of numerical simulations with the B2 SOLPS fluid code. Based on an ASDEX Upgrade shot, steady-state pre-disruption equilibria are generated which are then subjected to a simulated thermal quench by artificially enhancing the perpendicular transport in the ion and electron channels. The enhanced transport coefficients flows the Rechester and Rosenbluth model (1978) for ergodic transport in a tokamak with destroyed flux surfaces, i. e. χ, D∼const. xT''5/2 where the constants differ by the square root of the mass ratio for ions and electrons. By varying the steady-state neutral puffing rate we can modify the divertor conditions in terms of plasma temperature and density. Our numerical findings indicate that the disruption characteristics depend on the pre disruptive divertor conditions. We study the timescales and the spatial distribution of the divertor power fluxes. The simulated disruptions show rise and decay timescales in the range observed at ASDEX Upgrade. The decay timescale for the central electron temperature of ∼800 μs is typical for non-ITB disruptions. Varying the divertor conditions we find a distinct transition from a regime with symmetric power fluxes to inboard and outboard divertors to a regime where the bulk of the power flux goes to the outboard divertor. This asymmetry in the divertor peak fluxes for the higher puffing case is accompanied by a time delay between the outboard and inboard divertor heat pulses. Our simulations show a strong variation of the width of the power deposition profile with the pre-disruptive divertor conditions. In the experiment, power deposition profiles broaden significantly during plasma disruptions. Within the scope our transport model in B2, this behaviour is found for the cases with low densities, while profile widths change only little for the cases with strong neutral puffing. Further investigation, both experimental and theoretical, is needed for a better understanding of the thermal quench phase of disruptions. (Author)

[en] According to our present knowledge, a disruption in ITER is going to convert a large fraction of the plasma current into a beam of highly energetic runaway electrons. In order to prevent the generation of the runaway electrons, the electron density must be increased up to 5 x 10²² m^-3, which is a factor of 500 larger than the nominal one. Such density is difficult to reach, it has a strong impact on the design of the pumping system and on the operation of the tokamak. Therefore a significant effort must be addressed in understanding which processes control the fueling efficiency in order to maximize it. This work illustrates the progress made at ASDEX Upgrade in (1) developing and installing a valve close to the plasma with a relatively large fueling efficiency and (2) understanding the experimental results. The experimental data collected up to now indicate that the fueling efficiency ranges between 25 and 40% for the in-vessel valve and is independent of the type of gas injected (experiments with helium have not been done yet). In addition it does not depend clearly on the quantity of gas injected and is not a strong function of plasma energy. The code package SOLPS is used to model gas injection in an ASDEX Upgrade plasma; the experimental measurements are directly compared with the code prediction. (author)

No abstract available

[en] The presence of an extended region of open flux surfaces (halo). during the current quench phase of the disruption of elongated plasmas, is supported by measurements of halo currents and by numerical simulations. The halo, in addition to providing a poloidal current path between the plasma and the first-wall components, allows rapid conduction and convection of energy along field lines, and therefore a mechanism for the localized deposition of energy onto the wall. The heat load to the region of the plasma-first-wall interaction is higher than in the scenario in which the magnetic energy is mostly dissipated by radiative processes. (author). Letter-to-the-editor. 8 refs, 4 figs

[en] Scrape-off layer (SOL) currents are measured by means of Langmuir probes and shunts in the divertor of ASDEX Upgrade. They consist of the overlayed contributions of thermoelectric and Pfirsch-Schlueter (PS) currents. The SOL currents exhibit a drastic decrease when the line-averaged density approaches the Greenwald density in the H-mode. An analytical model is presented which reproduces the measured thermoelectric current quantitatively. Matching of the analytical model with the measured current scaling yields information about divertor temperatures and SOL e-folding lengths