Results 1 - 10 of 321
Results 1 - 10 of 321. Search took: 0.021 seconds
|Sort by: date | relevance|
[en] In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma x drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.
[en] A helical MHD perturbation in a finite-conductivity tokamak plasma has been considered in the straight-cylinder model in a situation where there is no resonance surface q = m/n in plasma. The radial eigenfunction of the helical mode, in addition to the large-scale component described at σ||→ ∞ by the ideal MHD equation, contains a small-scale component localized near the wall and near discontinuities in the radial profiles of the unperturbed quantities. At smooth profiles, the small-scale component is attached to the wall and is smaller in magnitude than the large-scale component. Therefore, beyond a thin near-wall plasma layer, the mode is close to the large-scale ideal MHD mode. The presence of the small-scale component is necessary to satisfy the boundary conditions for the perturbed field on the wall.
[en] Full text: The SST-1 Tokamak has 16 toroidal field (TF) coils and 9 superconducting poloidal field (PF) coils. They have been assembled and the positioning inaccuracy of the coils is measured. The deviations in coil positions will generate error field and this will degrade the plasma performance. The error field produced by the TF coil misalignment can impact the plasma startup and it is necessary to quantify that error. The averaged toroidal field error can be measured and detected using an electron beam source inside tokamak vessel. In SST-1 tokamak (major radius R = 1.1 m and minor radius a = 0.2 m), a low voltage electron beam source is mounted on the radial port at the midplane and can be moved to any point between R = 0.95-1.35 m. Cameras are mounted in radial port as well as top port to capture the deviation of the electron beam lines. The vacuum vessel is filled with helium gas which creates a luminescent trace of the electron beam due to impact excitation, creating a visual toroidal electron beam inside the vacuum vessel when the toroidal field coils are energized. This paper present the experimental observation of beam deviation with respect to various TF currents and then, the estimation of measured error field in SST-1 tokamak. An attempt is also made to explain the electron beam deviations measured through these experiments. The deviations in R, φ and Z are incorporated into the numerical model of the SST-1 TF coils for the error-field estimation. Effect of other coils on this error field would also be analyzed. Error field profiles in both R and Z direction would be quantified, which would be a useful information for the plasma operation. (author)
[en] In the sixties soviet scientists developed a new fusion machine called tokamak. Since then a few hundreds of tokamaks have been constructed in the world. Today most of them contribute to the ITER project but other projects exist too. In the United kingdom the Joint European Torus (JET) at Culham was commissioned in 1983. JET achieved a world first by producing a relevant amount of energy coming from fusion. In France the tokamak WEST (the new name of the Tore Supra tokamak) holds the world record for the longest duration for a plasma: 6.5 mn. In Japan the JT-60 tokamak holds the record for the highest value for the triple fusion product (density, temperature, confinement time). Aditya is the first tokamak designed and built in India. Member of ITER, India has the hope of controlling fusion thanks to Helium 3 fusion. Russia is the native country for the tokamak concept. In the Kurchatov institute, the first plasma reaching a temperature of 10 million degrees for 20 milliseconds was realized in 1964 in the T-4 tokamak. An hybrid machine is also studied in Russia, combining fission and fusion with the aim of burning nuclear waste. In the United States, the DIII-D tokamak is operated by General Atomics but financed by DOE. Experiments performed in the DIII-D tokamak have helped to design ITER. In the United-States there are 13 experimental facilities dedicated to thermonuclear fusion following 2 research axes: magnetic confinement and inertial fusion. (A.C.)
[en] Full text: It is getting increasingly clear that many tokamak plasma phenomena which have traditionally been investigated separately, are actually intrinsically linked. One outstanding example along these lines— which is investigated in the present contribution — is the the interaction between Alfvén modes (AM), turbulence, and zonal structures (ZS), like zonal flows and geodesic acoustic modes. Recently, a strong interest was raised in the fusion community by the possibility of generating ZS via nonlinear interaction with global modes like Alfvén instabilities. In this work, the interaction of AM, turbulence and ZS is studied with the code ORB5. This model treats ions and electrons respectively as gyrokinetic and drift-kinetic. ORB5 is a nonlinear global particle-in-cell code, developed for turbulence studies and extended to its electromagnetic multispecies version for the investigation of Alfvén dynamics. Recently, the importance of the kinetic electron effects in the ZS dynamics has also been emphasized with ORB5. ORB5 has also accomplished a verification/benchmark phase for AMs and has been used for the study of the nonlinear wave-particle interaction. The competition between the different excitation mechanisms of ZS is the main focus of this work. When an EP population is added to the electromagnetic turbulence, the perturbed saturated field is observed to be modified by the presence of AMs. The effect of the different players are described separately, and in particular: wave-particle nonlinearity, wave-wave nonlinearity, effect of turbulence on AMs, effect of AMs on turbulence, for example via ZS generation, and bulk plasma ω* effects on the AM growth rate and saturation. Comparisons with analytical theory and other models like the gyrokinetic Eulerian code GENE are also done. (author)
[en] Full text: Steady-State Superconducting Tokamak-1 (SST-1) has 16 toroidal field (TF) and 9 superconducting poloidal field (PF) coils rated for 10 kA DC. The TF coils are connected in series and operated in DC condition, whereas the PF coils are operated independently in pulse mode. The SST-1 current feeder system (CFS) houses nine pairs of PF superconducting current leads and one pair of TF superconducting current leads. The SST-1 CFS had observed arcing incidences during OT discharge in past SST-1 campaigns. Similar arcing incidences have also been observed in other tokamaks devices also like KSTAR, W7-X, and EAST. The conditions which led to the electrical arcing in SST-1 CFS, thereby resulting in severe damages to the PF current leads and helium hydraulic lines will be presented in this paper. As an important preventive measure to avoid such arcing at PF current leads during SST-1 operation, insulation strengthening processes of the PF current leads have been initiated to increase the voltage isolation capability of the PF current leads. In the view of same, development of an insulation scheme using combination of polyimide and GFRP along with DGEBA epoxy resin and its validation at lab scale has been carried out. It involves study of chemical kinetics of resin towards curing cycle, electrical and mechanical characterizations of insulation samples at room temperature as well as at LN2 temperature. A breakdown voltage of > 25 kV DC has been successfully achieved with ∼ 1.2 mm of insulation thickness at lab scale insulation samples. In order to validate the proposed insulation system under specified helium Paschen conditions, a lab scale setup considering SST-1 operational requirements has been developed. The operation, salient features of test setup and results will also be presented in this paper. The progressive development of insulation system and validation from prototype scale to half-dummy current lead scale and thereafter implementation on actual PF current leads will also be presented in this paper. (author)
[en] The advantages of thermonuclear fusion are -) an inexhaustible source of energy that is present everywhere, -) an inherently safe physics principle, -) a limited impact on the environment, -) no production of high-level radioactive waste with long half-life, and -) no risk of proliferation. Throughout the world, 3 different technologies have emerged: the tokamak in which plasma confinement is achieved thanks to the magnetic field circulating in the plasma itself, the stellarator in which plasma confinement is achieved thanks to the magnetic field generated by a combination of coils around the fusion chamber, and the inertial fusion in which the fusion of light nuclei is achieved by using high power lasers to compress matter. (A.C.)
[en] Full text: The development of fusion as a viable power source is moving from the science driven design of experimental devices to the engineering design considerations required to develop a feasible power plant. An effective maintenance plan is essential because the time in maintenance is potentially very large. To be effective, a maintenance plan must be outlined early in the plant design because it has two key requirements that must be embedded from the outset of the plant layout. First it requires the efficient transport of components and equipment around the plant through corridors, shield doors and contamination control systems using the most appropriate transport system. Second it requires maintenance oriented strategies to reduce the maintenance burden and to achieve the maintenance in a shorter time, with lower risk of failure and with simpler recovery scenarios. Work is therefore required at the preconcept design stage to define the maintenance plan so that the design driving factors required to enable the plan can be embedded in the plant design from the outset. This paper will describe this work, including the key transport system that has been proposed for the transfer of components and equipment to and from the tokamak using ceiling mounted cranes and dexterous manipulator systems. A qualitative comparison will be made between the proposed system and an alternative cask-based system will be made. The paper will also briefly describe some of the proposed maintenance-oriented strategies and development and testing work that is being carried out to mitigate the technical risks associated with the proposed maintenance plan. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No. 633053 and from the RCUK Energy Programme, grant number EP/P012450/1. The views and opinions expressed herein do not necessarily reflect those of the European Commission. (author)
[en] Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less important than those at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level. (authors)
[en] Full text: In the design of next step tokamak devices, it will be of key importance to carefully optimize the plasma magnetic configuration, in particular its elongation, triangularity and aspect ratio. Indeed, the sharp dependence of the safety factor q upon plasma elongation and, as a consequence, of the maximum achievable plasma current in disruption-free operation mode is of paramount importance to achieve high plasma performances and high-Q. The aim of this paper is to study the effects that plasma aspect ratio, plasma-wall normalized distance and plasma current profile have on passive plasma vertical stabilization, this being quantified by the plasma stability margin and instability growth time. Following this first step, the key aspects and figures of merit of plasma active vertical stabilization will be reviewed, with specific emphasis on the maximum disturbances that can be recovered by the plasma active control system with fixed current and voltage ratings. Both in-vessel as well as ex-vessel optimally placed stabilization coils will be considered. State of the art plasma models (as based on linear; MHD theory) and metallic structures (fully 3D structures) will be deployed in study. The final goal is to derive the dependence of plasma elongation upon the parameters described above in order to achieve robust passive and active stabilization with and without internal coils. (author)