Table of contents

Ray tracing simulations based on experimental PBX-M equilibria show a limited range of the parallel wavenumber (n_||=k_||c/ω) along a ray trajectory. The range of n_|| accessible to the excited wave is shown to have both a lower and an upper bound. The ray's phase volume is projected into (n_||, R, Z) space to define a domain of wave accessibility for each toroidal mode number excited by the launcher. A comparison of circular and bean shaped plasmas with a high aspect ratio (A ~ 5.5) shows that the rays fill a substantially larger portion of the accessible domain in bean shaped plasmas. Furthermore, in the bean shaped case the accessible domain tends to extend to higher n_||

Experiments in which the divertor plates are biased with respect to the grounded vacuum chamber were conducted in the Tokamak de Varennes, with ohmic discharges and boronized walls. A comprehensive set of diagnostics was used and those effects relating to plasma-surface interactions and their consequences, namely recycling and impurity production, are reported here. The main conclusion is that the bias voltage actively controls the balance between the particle fluxes to the walls and the diverters, with a weak influence on the confinement. The quantities associated with plasma surface interactions, i.e. the impurity influx, the plasma contamination and the global recycling coefficient, increase as a function of the voltage applied between the plates and the wall. Typically, the quantities of interest vary by a factor of two or more in the range from -240 to 290 V. The poloidal fluxes to the upper or lower divertor are enhanced by negative or positive biasing, respectively, in agreement with the sign of the E*B drift. For negative biasing the reduction in the wall flux and increase in the poloidal flux combine to improve greatly the divertor efficiency: the divertor pressure and the impurity retention time are increased by factors of up to 5 and 8, respectively. The above effects, together with the SOL density profiles and the biasing current-voltage characteristic, are consistent with a model in which the radial transport in the SOL is limited by the ion mobility; a mobility of 0.04 m²/V.s is found. Besides allowing active density and impurity control, it is suggested that biasing could facilitate the attainment of a radiative divertor by injection of well retained impurities in the divertor

A negative current in the scrape-off region beyond the separatrix of a diverted Tokamak can improve its MHD stability. By utilizing a strong thermionic emitter a current of up to 25% of the plasma current can be driven in the scrape-off region. If the thermionic current is emitted at only one end of the open field lines, a current will flow in the absence of a potential bias. The heat to maintain the electron emission that drives the scrape-off current would be supplied by the plasma outflow and no additional power is needed

A fluid model that connects the regions of weak and strong collisionality of trapped electrons has been derived by taking moments of the kinetic equation. The fluid model also includes magnetic curvature effects to sufficiently high order to recover the isothermal limit for the perpendicular motion. The model reproduces both the collision dominated trapped electron mode and the effects of rare collisions

Predictions for TFTR current scaling experiments are investigated for four recent theoretically based models, to find support for empirical current scaling used for planning future fusion devices. The profile-consistent drift wave model is found to predict current scaling. None of three multimode models, including the Bateman Version 5.10 model, previously claimed to show current scaling for TFTR, succeeds in simulating measured temperature and density profiles for either high or low current experiments. Standard edge resistive ballooning mode models do not lead to I_p scaling of confinement, underestimating losses at low current by factors of 2 to 4

Improved confinement is achieved on TEXTOR under high power conditions (up to 4 MW of additional heating with NBI-co+ICRH, NBI-co+counter or NBI-co+counter+ICRH) with edge radiative cooling employing silicon or neon as the radiating impurities. It is shown that in quasi-stationary conditions up to 85% of the input power can be radiated. Such high power fractions offer the possibility of utilizing these techniques to facilitate the power exhaust problem for a Tokamak reactor. Discharges with edge radiative cooling exhibit enhanced confinement properties at high densities, e.g. at a central line averaged electron density of 7.5*10¹³ cm^-3, an enhancement factor of 1.7 over ITER L89-P confinement scaling is obtained with an edge q value as low as 2.7. Stable discharges have been obtained even with the q=2 surface located inside the radiating zone. Furthermore, for radiatively cooled discharges heated with balanced NBI-co+counter with or without ICRH, supershot-like peaked electron density profiles, with central density values above 1.0*10¹⁴ cm^-3 are observed. The present results show that there is no impurity accumulation in the centre and the Ne and/or Si concentration is so low that the reactivity of the plasma remains unaffected

A dynamic computational model of lower hybrid current drive in the presence of an electric field is described and some results are given. Details of geometry, plasma profiles and circuit equations are treated carefully. Two dimensional velocity space effects are approximated in a one dimensional Fokker-Planck treatment. The model is unable to approximate experimental results in some cases characterized by low density, low current, high aspect ratio and a launched spectrum at high phase velocity relative to the thermal velocity. In other cases, qualitative agreement with measurements is found. A simple formula already in the literature appears to determine whether agreement of model and experiment will be good or poor. Application to an experimental discharge in which q(0) is raised above unity shows an appropriate time-scale. Computation of a planned effort for high beta suggests potential success

Radiation losses from light impurities from tokamak boundary plasmas within the last closed magnetic surface, where radial transport is of great importance for ion distribution over charge states, are estimated. "Radiation strengths", characterizing the integral intensity of radiation, are calculated for different ions of carbon, oxygen, neon and silicon. Characteristic plasma parameters in the regions, where these particles mainly radiate, are determined. A simple zero dimensional model of the heat transport at the plasma edge is proposed and the effect of the particle transfer on the detachment conditions is examined. The results are compared with experimental data from ohmic and auxiliary heated discharges in TEXTOR. Conditions necessary for the creation of a "radiating periphery" in a reactor are discussed

Experimental determinations of the radial plasma flow velocity, of Z_eff and of the thermal energy content, made on the Frascati Tokamak Upgrade (FTU), have been compared with the predictions of the conventional neoclassical theory, of the extended neoclassical theory and of the pseudoneoclassical theory. In the comparison only the momentum balance and continuity equations were used. The best overall agreement is found for the second of these theories, thus confirming the result of previous analyses

The effects of the wave accessibility condition on lower hybrid current drive (LHCD) are investigated in JT-60U. The density dependence of the detectable maximum photon energy of hard X rays in LHCD plasmas is consistent with that of the maximum electron energy which is limited by the accessibility condition. The current drive efficiency and the driven current profile are affected by the accessibility condition. Inaccessible wave power enhances impurity content, particle recycling and main plasma radiation loss, and in extreme cases triggers a MARFE. The wave intensity on the plasma periphery, which is detected by a Langmuir probe located on the divertor plate, is also consistent with an inaccessible power fraction going into the plasma interior. A ray trace analysis taking account of toroidal effect indicates that the inaccessible wave power is localized on the plasma periphery and finally propagates into the divertor plasma region

A helical coil winding law has been found for a stellarator with many of the desirable properties of the Helias. High performance computer codes make it possible to optimize the configuration over a parameter space of large dimension. For reactor applications an ample gap is provided between the coils and the plasma. Line tracing shows that the magnetic surfaces are robust. Non-linear stability at high beta is verified by equilibrium calculations in three dimensions. A neoclassical theory of transport taking quasi-neutrality into account predicts that energy confinement times sufficient for ignition can be achieved economically. For power plant studies a more attractive modular version of this concept called the MHH stellarator is under investigation

The flexibility provided by a variety of ion cyclotron resonance heating (ICRH) and current drive (CD) scenarios for ITER requires that the antenna should operate at a number of frequencies spread over a wide frequency range (20-85 MHz). A short circuited strip-line antenna in which a long strap is connected in parallel with a very short section is shown to provide the frequency range of operation required for ITER. The short section acts as a matching element located within the antenna and improves the power coupling capability especially at low frequencies where the plasma coupling is generally poor. The short section also supports the feeder-line central conductor. This means there is no longer a need for a ceramic support in the immediate vicinity of the antenna that will be subjected to a harsh neutron environment in a reactor