Results 1 - 10 of 6039
Results 1 - 10 of 6039. Search took: 0.032 seconds
|Sort by: date | relevance|
[en] The basic in–out divertor asymmetry with respect to ion B × ∇B direction has been examined in EAST by changing the divertor configuration from upper single null (USN) to lower single null (LSN) during the same discharge without reversing BT. It is remarkable that the in–out asymmetry is reversed when moving from USN to LSN. However, modeling with SOLPS, taking into account classical drifts, shows little difference. The divertor magnetic configuration also affects the access to H-modes, favoring DN or near-DN divertor configurations on EAST. ELMs further enhance the in–out divertor asymmetry, leading to greater particle and heat deposition on the outer target with a broader footprint, presumably arising from enhanced ELM transport in the outboard region
[en] Turbulence, and turbulence-driven transport are ubiquitous in magnetically confined plasmas, where there is an intimate relationship between turbulence, transport, instability driving mechanisms (such as gradients), plasma flows, and flow shear. Though many of the detailed physics of the interrelationship between turbulence, transport, drive mechanisms, and flow remain unclear, there have been many demonstrations that transport and/or turbulence can be suppressed or reduced via manipulations of plasma flow profiles. This is well known in magnetic fusion plasmas [e.g., high confinement mode (H-mode) and internal transport barriers (ITB's)], and has also been demonstrated in laboratory plasmas. However, it may be that the levels of particle transport obtained in such cases [e.g. H-mode, ITB's] are actually lower than is desirable for a practical fusion device. Ideally, one would be able to actively feedback control the turbulent transport, via manipulation of the flow profiles. The purpose of this research was to investigate the feasibility of using both advanced model-based control algorithms, as well as non-model-based algorithms, to control cross-field turbulence-driven particle transport through appropriate manipulation of radial plasma flow profiles. The University of New Mexico was responsible for the experimental portion of the project, while our collaborators at the University of Montana provided plasma transport modeling, and collaborators at Lehigh University developed and explored control methods.
[en] A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a diverter. Significant progress has been made in all of these areas and the purpose of this paper is to summarize the major results obtained during the last two years. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped diverter chamber has controlled the density in H-mode discharges and a radiative diverter is a promising technique for controlling the heat flux from the main plasma
[en] After a general discussion of the experimental characteristics of the L-H transition and consideration of basic theoretical principles underlying models for it, this paper reviews the various theories of the L-H transition available in the literature, providing some background information on each theory and expressing the transition criteria in forms suitable for comparison with experiment. Some conclusions on the relevance of these models for explaining the experimental data on the transition are drawn. (author)
[en] In recent H-mode experiments at JET with giant ELMs a lateral deflection of hot tokamak plasma leaving the scrape-off layer and striking the divertor plate has been observed. This deflection can effect the divertor erosion caused by the hot plasma irradiation, because of enlarging the irradiated area. A simplified MHD model of the vapor shield plasma and of the hot plasma initially formed at time t → -∞ is analyzed. At t = -∞ both plasmas are assumed to stay on rest and to be separated by a boundary, which is parallel to the plate surface. The interaction between plasmas is assumed to develop gradually ('adiabatically') as exp(t/t0) with t0 ∝ 102 μs the ELM duration time. Electrical insulation of the core tokamak plasma is assumed everywhere except for the contact with the divertor. Electric currents are flowing only in the toroidal direction. These currents developing in the interaction zone of the hot plasma and the rather cold target plasma are calculated for inclined impact of the magnetized hot plasma. At such conditions the J x B force in the lateral direction accelerates the interacting plasmas. The motion of the cold plasma and the gradual increase of the plasma interaction intensity are shown to be important for the appropriate deflection magnitude. Adiabatically responding against the increase of the interaction intensity the cold plasma motion compensates significantly the currents thus decreasing the deflection compared to motionless approach. The calculated magnitude of the hot plasma deflection is comparable to the observed one. The results of the modeling are discussed in relation to the experiments. It is shown that sudden switching on of the interaction produces Alfven oscillations of large amplitudes causing much larger amplitudes of the magnetic field induced by the currents than in the adiabatic case. (orig.)
[en] We develop and test a model, EPED1.6, for the H-mode pedestal height and width based upon two fundamental and calculable constraints: (1) onset of non-local peeling-ballooning modes at low to intermediate mode number, (2) onset of nearly local kinetic ballooning modes at high mode number. Calculation of these two constraints allows a unique, predictive determination of both pedestal height and width. The present version of the model is first principles, in that no parameters are fit to observations, and includes important non-ideal effects. Extensive successful comparisons with existing experiments on multiple tokamaks, including experiments where predictions were made prior to the experiment, are presented, and predictions for ITER are discussed.
[en] DIII-D is making significant contributions to a scientific basis for sustained burning plasma operation. These include explorations of increasingly reactor-relevant scenarios, studies of key issues for projecting performance, development of techniques for handling heat and particle efflux, and assessment of key issues for the ITER research plan. Advanced scenarios are being optimized in DIII-D via experiments to empirically determine the relationship between transport and the current profile, which in turn can provide essential input to inform improvement of the theory-based models that do not currently capture the observed behaviour. Joint DIII-D/JET ρ* scans in the hybrid regime imply Bohm-like confinement scaling. Startup and shutdown techniques were developed for the restrictive environment of future devices while retaining compatibility with advanced scenarios. Towards the goal of a fully predictive capability, the DIII-D program emphasizes validation of physics-based models, facilitated by a number of new and upgraded diagnostics. Specific areas include transport, rotation, energetic particles and the H-mode pedestal, but this approach permeates the entire research programme. Concerns for heat and particle efflux in future devices are addressed through studies of ELM control, disruption avoidance and mitigation, and hydrogenic retention in DIII-D's carbon wall. DIII-D continues to respond to specific needs for ITER. Recent studies have compared H-mode access in several different ion species, identifying not only isotopic, but density, rotation and geometrical dependences that may guide access to H-mode during ITER's non-activated early operation. DIII-D used an insertable module to simulate the magnetic perturbations introduced by one of ITER's three test blanket module sets, demonstrating that little impact on performance is seen at ITER equivalent levels of magnetic perturbation.
[en] Results are presented from analysis of the database for open-quotes quiescentclose quotes (i.e., without ELM) H mode of confinement. It is shown that the only parameter (aside from a constant factor) in terms of which the scaling law must be modified for the H mode in comparison with that of the L mode is the dependence on the plasma elongation k = b/a. As a result of the analysis a power-law scaling is found for the quiescent H mode which can be written in the simple form τEHE = 1.5 x τEIT-89-L x k0.63, where τEIT-89-L is the ITER-89-L scaling for L mode. This new scaling law describes the experimental data better than previously obtained scaling laws for the H mode and can be recommended for estimating the confinement time in future tokamaks (ITER, DEMO). 7 refs., 11 figs., 3 tabs
[en] The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept
[en] An upgrade to the TEXT tokamak will include a poloidal divertor, located on the inside of the plasma radius. Combined with an increase in the available auxilliary heating, this upgrade willextend the present program to include the study of confinement in H-mode plasmas. The design details of the upgrade are presented, with the results of the equilibrium and stability studies. (author). 12 refs.; 5 figs.; 1 tab