International Nuclear Information System (INIS)
The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration
Axisymmetric control in tokamaks
International Nuclear Information System (INIS)
Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration
International Nuclear Information System (INIS)
The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration. (author) 19 figs., 2 tabs., 30 refs
Energy Technology Data Exchange (ETDEWEB)
Kim, Kimin [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Korea Advanced Institute of Science and Technology, Daejeon 305-701, Korea; Ahn, J-W [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scotti, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Park, J-K [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Menard, J. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
2015-09-03
Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifies the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.
HECTOR: a code for the study of high energy charged particles in axisymmetric tokamak plasmas
International Nuclear Information System (INIS)
A code for the study of high energy charged particles resulting primarily from thermonuclear reactions within the confining magnetic fields of non-circular axisymmetric tokamak plasmas is described. The trajectories of the particles are traced in the (C.O.M.) space using a new, fast, and efficient hybrid orbit following scheme based upon the drift equations in the guiding centre approximation and the constants of motion. The code includes the important Coulomb scattering processes of dynamical friction and pitch angle scattering. The code is specifically designed to operate within the experimental environment or in a predictive mode. (author)
Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Jong-kyu Park, Allen H. Boozer, Jonathan E. Menard, Andrea M. Garofalo, Michael J. Schaffer, Richard J. Hawryluk, Stanley M. Kaye, Stefan P. Gerhardt, Steve A. Sabbagh, and the NSTX Team
2009-04-22
Tokamaks are sensitive to deviations from axisymmetry as small as δB=B0 ~ 10-4. These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equiva- lently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not suffciently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents.
Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback
International Nuclear Information System (INIS)
A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs
Frantz, Eric Randall
Elongation and shaping of the tokamak plasma cross -section can allow increased beta and other favorable improvements. As the cross-section is made non-circular, however, the plasma can become unstable against axisymmetric motions, the most predominant one being a nearly uniform displacement in the direction of elongation. Without additional stabilizing mechanisms, this instability has growth rates typically (TURN)10('6)sec('-1). With passive and active feedback from external conductors, the plasma can be significantly slowed down and controlled. In this work, a mathematical formulism for analyzing the vertical instability is developed in which the external conductors are treated (or broken -up) as discrete coils. The circuit equations for the plasma induced currents can be included within the same mathematical framework. The plasma equation of motion and the circuit equations are combined and manipulated into a diagonalized form that can be graphically analyzed to determine the growth rate. An effective mode approximation (EMA) to the dispersion relation in introduced to simplify and approximate the growth rate of the more exact case. Controller voltage equations for active feedback are generalized to include position and velocity feedback and time delay. A position cut-off displacement is added to model finite spatial resolution of the position detectors or a dead-band voltage level. Stability criteria are studied for EMA and the more exact case. The time dependent responses for plasma position controller voltages, and currents are determined from the Laplace transformations. Slow responses are separated from the fast ones (dependent on plasma inertia) using a typical tokamak ordering approximation. The methods developed are applied in numerous examples for the machine geometry and plasma of TNS, an inside-D configuration plasma resembling JET, INTOR, or FED.
Energy Technology Data Exchange (ETDEWEB)
Stacey, W. M. [Georgia Institute of Technology, Atlanta, Georgia 30332 (United States); Bae, C. [National Fusion Research Institute, Daejoen (Korea, Republic of)
2015-06-15
A systematic formalism for the calculation of rotation in non-axisymmetric tokamaks with 3D magnetic fields is described. The Braginskii Ωτ-ordered viscous stress tensor formalism, generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry, and the resulting fluid moment equations provide a systematic formalism for the calculation of toroidal and poloidal rotation and radial ion flow in tokamaks in the presence of various non-axisymmetric “neoclassical toroidal viscosity” mechanisms. The relation among rotation velocities, radial ion particle flux, ion orbit loss, and radial electric field is discussed, and the possibility of controlling these quantities by producing externally controllable toroidal and/or poloidal currents in the edge plasma for this purpose is suggested for future investigation.
International Nuclear Information System (INIS)
A systematic formalism for the calculation of rotation in non-axisymmetric tokamaks with 3D magnetic fields is described. The Braginskii Ωτ-ordered viscous stress tensor formalism, generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry, and the resulting fluid moment equations provide a systematic formalism for the calculation of toroidal and poloidal rotation and radial ion flow in tokamaks in the presence of various non-axisymmetric “neoclassical toroidal viscosity” mechanisms. The relation among rotation velocities, radial ion particle flux, ion orbit loss, and radial electric field is discussed, and the possibility of controlling these quantities by producing externally controllable toroidal and/or poloidal currents in the edge plasma for this purpose is suggested for future investigation
Radial electrical field in non axi-symmetrical tokamak plasmas - study through doppler reflectometry
International Nuclear Information System (INIS)
Nuclear fusion research aims at producing plasmas mainly heated by fusion reactions between Tritium and Deuterium ions. This work deals with the problem of turbulent transport, which is one of the main limiting factors in the performance of tokamak operation. It is focused on the radial electric field (Er, pointing outwards/inwards from the plasma), which can generate transport barriers when its shearing rate is sufficient to cause a turbulence de-correlation. We have investigated the mechanisms causing the spontaneous generation of the radial electric field inside the last closed magnetic surface. In the Tore Supra tokamak, a Doppler reflectometer allows a quasi-direct measurement of the electric drift velocity due to Er. The effect of ripple (a periodic variation of the magnetic field between two coils, in the toroidal direction) is shown by comparing the measurements with predictions from various models, corresponding to different diffusion regimes (ripple-plateau, local trapping). In some special experimental conditions, a locally positive radial electric field has been measured inside the last closed flux surface in Tore Supra, which contrasts with the usual negative Er in this region. This suggests the presence of other non-ambipolar mechanisms. A discussion on the possible role of MHD activity and islands based on the Doppler reflectometry measurements is made. (author)
Energy Technology Data Exchange (ETDEWEB)
Strait, E. J. [General Atomics, San Diego, CA (United States); Park, J. -K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Marmar, E. S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ahn, J. -W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Berkery, J. W. [Columbia Univ., New York, NY (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Canik, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delgado-Aparicio, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. M. [General Atomics, San Diego, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Greenwald, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kim, K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); King, J. D. [General Atomics, San Diego, CA (United States); Lanctot, M. J. [General Atomics, San Diego, CA (United States); Lazerson, S. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, Y. Q. [Culham Science Centre, Abingdon (United Kingdom). Euratom/CCFE Association; Logan, N. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lore, J. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Menard, J. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Nazikian, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Shafer, M. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Paz-Soldan, C. [General Atomics, San Diego, CA (United States); Reiman, A. H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Rice, J. E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Sabbagh, S. A. [Columbia Univ., New York, NY (United States); Sugiyama, L. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Turnbull, A. D. [General Atomics, San Diego, CA (United States); Volpe, F. [Columbia Univ., New York, NY (United States); Wang, Z. R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Wolfe, S. M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
2014-09-30
The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10^{-4} of the main axisymmetric field, such “3D” fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data, in
Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry
International Nuclear Information System (INIS)
If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.
Fusion-product transport in axisymmetric tokamaks: losses and thermalization
International Nuclear Information System (INIS)
High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-β, non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated
Computational studies of tokamak plasmas
International Nuclear Information System (INIS)
Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)
Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks
International Nuclear Information System (INIS)
Experiments and axisymmetric MHD simulations on tokamak disruptions have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an advantageous vertical plasma position to avoiding VDEs during the plasma current quench, is shown to be fairly insensitive to plasma shape and current profile parameters. Secondly, a rapid flattening of the plasma current profile frequently seen at thermal quench is newly clarified to play a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments. Together with the attractive force that arises from passive shell currents and essentially vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)
Modeling non-stationary, non-axisymmetric heat patterns in DIII-D tokamak
Ciro, D; Caldas, I L
2016-01-01
Non-axisymmetric stationary magnetic perturbations lead to the formation of homoclinic tangles near the divertor magnetic saddle in tokamak discharges. These tangles intersect the divertor plates in static helical structures that delimit the regions reached by open magnetic field lines reaching the plasma column and leading the charged particles to the strike surfaces by parallel transport. In this article we introduce a non-axisymmetric rotating magnetic perturbation to model the time development of the three-dimensional magnetic field of a single-null DIII-D tokamak discharge developing a rotating tearing mode. The stable and unstable manifolds of the asymmetric magnetic saddle are calculated through an adaptive method providing the manifold cuts at a given poloidal plane and the strike surfaces. For the modeled shot, the experimental heat pattern and its time development are well described by the rotating unstable manifold, indicating the emergence of homoclinic lobes in a rotating frame due to the plasma ...
Non-Axisymmetric Equilibrium Reconstruction for Stellarators, Reversed Field Pinches and Tokamaks
International Nuclear Information System (INIS)
Full text: Equilibrium reconstruction is the process of minimizing the mismatch between modeled and observed signals by changing the parameters that specify the equilibrium. While stellarator equilibria are inherently non-axisymmetric, non-axisymmetric effects are also crucial for understanding stability and confinement of high-performance reversed field pinch and tokamak plasmas. Therefore, two-dimensional reconstruction tools are not adequate for fully exploring 3D plasmas. The V3FIT and STELLOPT codes are 3D equilibrium reconstruction codes, both based on the VMEC 3D equilibrium code. VMEC models field-period symmetric 3D flux surface geometry but does not treat magnetic islands and chaotic regions. VMEC requires the specification of the pressure and either rotational transform or toroidal current profiles, as functions of either the toroidal or poloidal flux. VMEC can treat both axisymmetric and non-axisymmetric configurations, both free- and fixed-boundary equilibria, and both stellarator-symmetric and non-stellarator-symmetric equilibria. Both V3FIT and STELLOPT can utilize signals from magnetic diagnostics, soft X-rays (SXR), Thomson scattering, and geometrical information from plasma limiters. STELLOPT can also utilize Motional Stark Effect (MSE) signals. Both calculate a finite difference approximation to a Jacobian for the signal-mismatch minimization. V3FIT and STELLOPT differ in the details of their minimization algorithms, their utilization of auxiliary profiles (like electron density and soft x-ray emissivity), and in their computation of model signals. V3FIT is currently in use on stellarators (HSX, CTH), reversed field pinches (RFX-mod) and tokamaks (DIII-D) for a wide variety of studies: interpretation of Pfirsch-Schliiter and bootstrap currents, design of new magnetic diagnostics, magnetic island generation, vertical instabilities, density-limit disruption activity, conformance of multiple data sources to a single set of flux surfaces, quasi
Polarization spectroscopy of tokamak plasmas
International Nuclear Information System (INIS)
Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs
Axisymmetric equilibria with pressure anisotropy and plasma flow
Evangelias, A
2016-01-01
A generalised Grad-Shafranov equation that governs the equilibrium of an axisymmetric toroidal plasma with anisotropic pressure and incompressible flow of arbitrary direction is derived. This equation includes six free surface functions and recovers known Grad-Shafranov-like equations in the literature as well as the usual static, isotropic one. The form of the generalised equation indicates that pressure anisotropy and flow act additively on equilibrium. In addition, two sets of analytical solutions, an extended Solovev one with a plasma reaching the separatrix and an extended Hernegger-Maschke one for a plasma surrounded by a fixed boundary possessing an X-point, are constructed, particularly in relevance to the ITER and NSTX tokamaks. Furthermore, the impacts both of pressure anisotropy and plasma flow on these equilibria are examined. It turns out that depending on the maximum value and the shape of an anisotropy function, the anisotropy can act either paramagnetically or diamagnetically. Also, in most of...
Axisymmetric equilibria with pressure anisotropy and plasma flow
Evangelias, Achilleas
2016-01-01
In this Master thesis we investigate the influence of pressure anisotropy and incompressible flow of arbitrary direction on the equilibrium properties of magnetically confined, axisymmetric toroidal plasmas. The main novel contribution is the derivation of a pertinent generalised Grad-Shafranov equation. This equation includes six free surface functions and recovers known Grad-Shafranov-like equations in the literature as well as the usual static, isotropic one. The form of the generalised equation indicates that pressure anisotropy and flow act additively on equilibrium. In addition, two sets of analytical solutions, an extended Solovev one with a plasma reaching the separatrix and an extended Hernegger-Maschke one for a plasma surrounded by a fixed boundary possessing an X-point, are constructed, particularly in relevance to the ITER and NSTX tokamaks. Furthermore, the impacts both of pressure anisotropy, through an anisotropy function assumed to be uniform on the magnetic surfaces, and plasma flow, via the...
Axisymmetric disruption dynamics including current profile changes in the ASDEX-Upgrade tokamak
International Nuclear Information System (INIS)
Axisymmetric MHD simulations have revealed a new driving mechanism that governs the vertical displacement event (VDE) dynamics in tokamak disruptions. A rapid flattening of the plasma current profile during the disruption plays a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect, due to an abrupt change in the current profile, is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the plasma current quench, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)
Equilibrium and ballooning mode stability of an axisymmetric tensor pressure tokamak
International Nuclear Information System (INIS)
A force balance relation, a representation for the poloidal beta (β/sub p/), and expressions for the current densities are derived from the MHD equilibrium relations for an axisymmetric tensor pressure tokamak. Perpendicular and parallel beam pressure components are evaluated from a distribution function that models high energy neutral particle injection. A double adiabatic energy principle is derived from that of Kruskal and Oberman, with correction terms added. The energy principle is then applied to an arbitrary cross-section axisymmetric tokamak to examine ballooning instabilities of large toroidal mode number. The resulting Euler equation is remarkably similar to that of ideal MHD. Although the field-bending term is virtually unaltered, the driving term is modified because the pressures are no longer constant on a flux surface. Either a necessary or a sufficient marginal stability criterion for a guiding center plasma can be derived from this equation whenever an additional stabilizing element unique to the double adiabatic theory is either kept or neglected, respectively
Nonlinear electromagnetic gyrokinetic equations for rotating axisymmetric plasmas
International Nuclear Information System (INIS)
The influence of sheared equilibrium flows on the confinement properties of tokamak plasmas is a topic of much current interest. A proper theoretical foundation for the systematic kinetic analysis of this important problem has been provided here by presented the derivation of a set of nonlinear electromagnetic gyrokinetic equations applicable to low frequency microinstabilities in a rotating axisymmetric plasma. The subsonic rotation velocity considered is in the direction of symmetry with the angular rotation frequency being a function of the equilibrium magnetic flux surface. In accordance with experimental observations, the rotation profile is chosen to scale with the ion temperature. The results obtained represent the shear flow generalization of the earlier analysis by Frieman and Chen where such flows were not taken into account. In order to make it readily applicable to gyrokinetic particle simulations, this set of equations is cast in a phase-space-conserving continuity equation form
Chu, M. S.; Guo, Wenfeng
2016-06-01
The frequency spectrum and mode structure of axisymmetric electrostatic oscillations [the zonal flow (ZF), sound waves (SW), geodesic acoustic modes (GAM), and electrostatic mean flows (EMF)] in tokamaks with general cross-sections and toroidal flows are studied analytically using the electrostatic approximation for magnetohydrodynamic modes. These modes constitute the "electrostatic continua." Starting from the energy principle for a tokamak plasma with toroidal rotation, we showed that these modes are completely stable. The ZF, the SW, and the EMF could all be viewed as special cases of the general GAM. The Euler equations for the general GAM are obtained and are solved analytically for both the low and high range of Mach numbers. The solution consists of the usual countable infinite set of eigen-modes with discrete eigen-frequencies, and two modes with lower frequencies. The countable infinite set is identified with the regular GAM. The lower frequency mode, which is also divergence free as the plasma rotation tends to zero, is identified as the ZF. The other lower (zero) frequency mode is a pure geodesic E×B flow and not divergence free is identified as the EMF. The frequency of the EMF is shown to be exactly 0 independent of plasma cross-section or its flow Mach number. We also show that in general, sound waves with no geodesic components are (almost) completely lost in tokamaks with a general cross-sectional shape. The exception is the special case of strict up-down symmetry. In this case, half of the GAMs would have no geodesic displacements. They are identified as the SW. Present day tokamaks, although not strictly up-down symmetric, usually are only slightly up-down asymmetric. They are expected to share the property with the up-down symmetric tokamak in that half of the GAMs would be more sound wave-like, i.e., have much weaker coupling to the geodesic components than the other half of non-sound-wave-like modes with stronger coupling to the geodesic
International Nuclear Information System (INIS)
The neoclassical theory of transport is used to calculate the current efficiency of electronic cyclotron impulsion (ECCD) in an axisymmetric tokamak in the few collisions regime. The standard parameter of the tokamak is used to obtain a system of equations that describe the hydrodynamic of the plasma, where the ponderomotive force (PM) due to high power radio frequency waves is taken in account. The PM force is produced in the proximity of electron cyclotron resonance surface in a specific poloidal localization. The efficiency ECCD is analyzed in the cases of first and second harmonic (for different angles of injection of radio frequency waves) and it is validated using the experimental values of the TCV and T-10 tokamaks. The results are according to those obtained by means of the techniques of the Green functions. (Author)
Edge plasma diagnostics in tokamaks
Czech Academy of Sciences Publication Activity Database
Stöckel, Jan; Brotánková, Jana; Hron, Martin; Adámek, Jiří; Ďuran, Ivan; Van Oost, G.; Peleman, P.; Gunn, J.; Devynck, P.; Martines, E.; Schrittwieser, R.; Kocan, M.
Kudowa Zdrój : -, 2006, s. 910-935. [Sixth International Workshop and Summer School Towards Fusion Energy - Plasma Physics, Diagnostics, Spin-offs. Kudowa Zdrój (PL), 18.09.2006-22.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * diagnostics * heating Subject RIV: BL - Plasma and Gas Discharge Physics
Plasma boundary phenomena in tokamaks
International Nuclear Information System (INIS)
The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (ne and Te) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)
Canonical transformation for trapped/passing guiding-center orbits in axisymmetric tokamak geometry
Energy Technology Data Exchange (ETDEWEB)
Brizard, Alain J. [Department of Physics, Saint Michael' s College, Colchester, Vermont 05439 (United States); Duthoit, François-Xavier [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); SNU Division of Graduate Education for Sustainabilization of Foundation Energy, Seoul National University, Seoul 151-742 (Korea, Republic of)
2014-05-15
The generating function for the canonical transformation from the parallel canonical coordinates (s,p{sub ||}) to the action-angle coordinates (ζ, J) for trapped/passing guiding-center orbits in axisymmetric tokamak geometry is presented. Drawing on the analogy between the phase-space portraits of the librating/rotating pendulum and the trapped/passing guiding-center orbits, the generating function is expressed in terms of the Jacobi zeta function, which can then readily be used to obtain an explicit expression for the bounce-center transformation for trapped/passing-particle guiding-center orbits in axisymmetric tokamak geometry.
Bootstrap current for tokamak plasma with anisotropic electron temperature
International Nuclear Information System (INIS)
The neoclassical bootstrap current for an anisotropic plasma has been studied in a large aspect-ratio tokamak. The enhancement factor due to the temperature anisotropy in the equilibrium electron distribution function is explicitly calculated, and is shown to reach to about 1.5 when the perpendicular temperature is twice as large as the parallel temperature. This bootstrap current is also predicted to have the component proportional to the radial electric field even in an axisymmetric magnetic field. (author)
Axisymmetric equilibria with pressure anisotropy and plasma flow
Evangelias, A.; Throumoulopoulos, G. N.
2016-04-01
A generalised Grad-Shafranov equation that governs the equilibrium of an axisymmetric toroidal plasma with anisotropic pressure and incompressible flow of arbitrary direction is derived. This equation includes six free surface functions and recovers known Grad-Shafranov-like equations in the literature as well as the usual static, isotropic one. The form of the generalised equation indicates that pressure anisotropy and flow act additively on equilibrium. In addition, two sets of analytical solutions, an extended Solovev one with a plasma reaching the separatrix and an extended Hernegger-Maschke one for a plasma surrounded by a fixed boundary possessing an X-point, are constructed, particularly in relevance to the ITER and NSTX tokamaks. Furthermore, the impacts both of pressure anisotropy and plasma flow on these equilibria are examined. It turns out that depending on the maximum value and the shape of an anisotropy function, the anisotropy can act either paramagnetically or diamagnetically. Also, in most of the cases considered both the anisotropy and the flow have stronger effects on NSTX equilibria than on ITER ones.
Plasma equilibria and stationary flows in axisymmetric systems. Pt. 1
International Nuclear Information System (INIS)
During discharges within a tokamak device such as JET fluctuations are observed in the plasma, of plasma density, temperature, electric potential and of the magnetic field. These fluctuations have complicated structure and are linked with different kinds of instabilities. However, it is not clear which instabilities are most important in determining the behaviour of the plasma. A comprehensive numerical theory which can predict the effect of the instabilities on the transport of plasma in axisymmetric systems has been sought using the static Grad-Shafranov-Schlueter (SGSS) equation as a basis. However, the static equation was over simplified for the situation in JET with additional heating giving rise to large toroidal flows, and an extended equation (EGSS) was developed. The results of the study include the discovery of algebraic branches of solutions to the EGSS equation even for very small poloidal flows, solutions to the inverse problem for the SGSS and EGSS equations using Fourier decomposition, classification of the boundary condition at the magnetic axis, demonstration of a visible effect of the poloidal flow on the separation of the density surface and the magnetic surface an indication of the existence of multiple branches of solutions to the EGSS and SGSS equations and their relation to stability properties. (U.K.)
Axisymmetric equilibria with pressure anisotropy and plasma flow
Throumoulopoulos, George; Evangelias, Achilleas
2015-11-01
A generalised Grad-Shafranov equation that governs the equilibrium of an axisymmetric toroidal plasma with anisotropic pressure and incompressible flow of arbitrary direction is derived. This equation includes six free surface functions and recovers known Grad-Shafranov-like equations in the literature as well as the usual static, isotropic one. The form of the generalised equation indicates that pressure anisotropy and flow act additively on equilibrium. In addition, two sets of analytical solutions, an extended Solovev one with a free boundary and an extended Hernegger-Maschke one for a plasma surrounded by a fixed boundary possessing an X-point, are constructed, particularly in relevance to the ITER and NSTX tokamaks. Furthermore, the impacts both of pressure anisotropy and plasma flow on these equilibria are examined. It turns out that depending on the maximum value and the shape of an anisotropy function, the anisotropy can act either paramagnetically or diamagnetically. Also, in most of the cases considered both the anisotropy and the flow have stronger effects on NSTX equilibria than on ITER ones. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from (a) the National Programme for the Controlled Thermonuclear Fusion, Hellenic Republic, (b) Euratom research and training programme 2014-2018.
Atomic physics in tokamak plasmas
International Nuclear Information System (INIS)
Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)
Toroidicity and shape dependence of peeling mode growth rates in axisymmetric toroidal plasmas
International Nuclear Information System (INIS)
The growth rate of the peeling mode instability with large toroidal mode number is calculated for general axisymmetric toroidal plasmas, including tokamaks and the spherical torus (ST) equilibia by using formalism presented by Connor et al. Analytic equilibia with non-zero edge current density and quasi-uniform current profiles are assumed. It is found that in sharp D-shape tokamak plasma, the derivative of the safety factor with respect to the poloidal flux becomes very large, making the perturbed poloidal motion very large, in turn making a significant reduction of the growth rate of the peeling mode, similar to the X-point effect in diverted plasma. The large aspect ratio effect is also studied, which reduces the growth rate further. (physics of gases, plasmas, and electric discharges)
Compact formulas for bounce/transit averaging in axisymmetric tokamak geometry
International Nuclear Information System (INIS)
Compact formulas for bounce and transit orbit averaging of the fluctuation-amplitude eikonal factor in axisymmetric tokamak geometry, which is frequently encountered in bounce-gyrokinetic description of microturbulence, are given in terms of the Jacobi elliptic functions and elliptic integrals. These formulas are readily applicable to the calculation of the neoclassical susceptibility in the framework of modern bounce-gyrokinetic theory. In the long-wavelength limit for axisymmetric electrostatic perturbations, we recover the expression for the Rosenbluth-Hinton residual zonal flow [M. N. Rosenbluth and F. L. Hinton, Phys. Rev. Lett. 80, 724 (1998)] accurately
Potential turbulence in tokamak plasmas
International Nuclear Information System (INIS)
Microscopic potential turbulence in tokamak plasmas are investigated by a multi-sample-volume heavy ion beam probe. The wavenumber/frequency spectra S(k,ω) of the plasmas potential fluctuation as well as density fluctuation are obtained for the first time. The instantaneous turbulence-driven particle flux, calculated from potential and density turbulence has oscillations of which amplitude is about 100 times larger than the steady-state outwards flux, showing sporadic behaviours. We also observed large-scale coherent potential oscillations with the frequency around 10-40 kHz. (author)
Continuum kinetic modeling of the tokamak plasma edge
Dorf, M. A.; Dorr, M. R.; Hittinger, J. A.; Cohen, R. H.; Rognlien, T. D.
2016-05-01
The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.
Tokamak plasma position dynamics and feedback control
International Nuclear Information System (INIS)
The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form
The disruptive instability in Tokamak plasmas
Salzedas, F.J.B.
2001-01-01
Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te
Anisotropic plasma with flows in tokamak: Steady state and stability
International Nuclear Information System (INIS)
An adequate description of equilibrium and stability of anisotropic plasma with macroscopic flows in tokamaks is presented. The Chew-Goldberger-Low (CGL) approximation is consistently used to analyze anisotropic plasma dynamics. The admissible structure of a stationary flow is found to be the same as in the ideal magnetohydrodynamics with isotropic pressure (MHD), which means an allowance for the same relabeling symmetry as in ideal MHD systems with toroidally nested magnetic surfaces. A generalization of the Grad-Shafranov equation for the case of anisotropic plasma with flows confined in the axisymmetric magnetic field is derived. A variational principle was obtained, which allows for a stability analysis of anisotropic pressure plasma with flows, and takes into account the conservation laws resulting from the relabeling symmetry. This principle covers the previous stability criteria for static CGL plasma and for ideal MHD flows in isotropic plasma as well. copyright 1996 American Institute of Physics
Tokamak plasma interaction with limiters
International Nuclear Information System (INIS)
The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict
Stability analysis of tokamak plasmas
International Nuclear Information System (INIS)
In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)
Elements of Neoclassical Theory and Plasma Rotation in a Tokamak
Smolyakov, A.
2015-12-01
The following sections are included: * Introduction * Quasineutrality condition * Diffusion in fully ionized magnetized plasma and automatic ambipolarity * Toroidal geometry and neoclassical diffusion * Diffusion and ambipolarity in toroidal plasmas * Ambipolarity and equilibrium poloidal rotation * Ambipolarity paradox and damping of poloidal rotation * Neoclassical plasma inertia * Oscillatory modes of poloidal plasma rotation * Dynamics of the toroidal momentum * Momentum diffusion in strongly collisional, short mean free path regime * Diffusion of toroidal momentum in the weak collision (banana) regime * Toroidal momentum diffusion and momentum damping from drift-kinetic theory and fluid moment equations * Comments on non-axisymmetric effects * Summary * Acknowledgments * Appendix: Trapped (banana) particles and collisionality regimes in a tokamak * Appendix: Hierarchy of moment equations * Appendix: Plasma viscosity tensor in the magnetic field: parallel viscosity, gyroviscosity, and perpendicular viscosity * Appendix: Closure relations for the flux surface averaged parallel viscosity in neoclassical (banana and plateau) regimes * References
Simulation of burning tokamak plasmas
International Nuclear Information System (INIS)
To simulate dynamical behaviour of tokamak fusion reactors, a zero-dimensional time-dependent particle and power balance code has been developed. The zero-dimensional plasma model is based on particle and power balance equations that have been integrated over the plasma volume using prescribed profiles for plasma parameters. Therefore, the zero-dimensional model describes the global dynamics of a fusion reactor. The zero-dimensional model has been applied to study reactor start-up, and plasma responses to changes in the plasma confinement, fuelling rate, and impurity concentration, as well as to study burn control via fuelling modulation. Predictions from the zero-dimensional code have been compared with experimental data and with transport calculations of a higher dimensionality. In all cases, a good agreement was found. The advantage of the zero-dimensional code, as compared to higher-dimensional transport codes, is the possibility to quickly scan the interdependencies between reactor parameters. (88 refs., 58 figs., 6 tabs.)
Edge plasma studies on the CASTOR tokamak
Czech Academy of Sciences Publication Activity Database
Hron, Martin; Peleman, P.; Spolaore, M.; Martines, E.; Hronová-Bilyková, Olena; Dejarnac, Renaud; Devynck, P.; Brotánková, Jana; Sentkerestiová, Jana; Ďuran, Ivan; Gunn, J.; Stöckel, Jan; Van Oost, G.; Adámek, Jiří; van de Peppel, L.; Štěpán, Michal
Krakow : Euratom - IPPLM Association, 2006 - (Zagorski, R.), - [IEA Large Tokamak IA Workshop on Edge Transport in Fusion plasmas. Kraków (PL), 11.09.2006-13.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://www.etfp2006.ifpilm.waw.pl/presentations.html
Tokamak plasma self-organization-synergetics of magnetic trap plasmas
Razumova, K. A.; Andreev, V. F.; Eliseev, L. G.; Kislov, A. Y.; La Haye, R. J.; Lysenko, S. E.; Melnikov, A. V.; Notkin, G. E.; Pavlov, Y. D.; Kantor, M. Y.
2011-01-01
Analysis of a wide range of experimental results in plasma magnetic confinement investigations shows that in most cases, plasmas are self-organized. In the tokamak case, it is realized in the self-consistent pressure profile, which permits the tokamak plasma to be macroscopically MHD stable. Existin
King, J D; Strait, E J; Boivin, R L; Taussig, D; Watkins, M G; Hanson, J M; Logan, N C; Paz-Soldan, C; Pace, D C; Shiraki, D; Lanctot, M J; La Haye, R J; Lao, L L; Battaglia, D J; Sontag, A C; Haskey, S R; Bak, J G
2014-08-01
The DIII-D tokamak magnetic diagnostic system [E. J. Strait, Rev. Sci. Instrum. 77, 023502 (2006)] has been upgraded to significantly expand the measurement of the plasma response to intrinsic and applied non-axisymmetric "3D" fields. The placement and design of 101 additional sensors allow resolution of toroidal mode numbers 1 ≤ n ≤ 3, and poloidal wavelengths smaller than MARS-F, IPEC, and VMEC magnetohydrodynamic model predictions. Small 3D perturbations, relative to the equilibrium field (10(-5) precision is achieved using electrical discharge machined components, and alignment techniques employing rotary laser levels and a coordinate measurement machine. A 16-bit data acquisition system is used in conjunction with analog signal-processing to recover non-axisymmetric perturbations. Co-located radial and poloidal field measurements allow up to 14.2 cm spatial resolution of poloidal structures (plasma poloidal circumference is ~500 cm). The function of the new system is verified by comparing the rotating tearing mode structure, measured by 14 BP fluctuation sensors, with that measured by the upgraded B(R) saddle loop sensors after the mode locks to the vessel wall. The result is a nearly identical 2/1 helical eigenstructure in both cases. PMID:25173265
Plasma equilibrium and instabilities in tokamaks
International Nuclear Information System (INIS)
A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.)
Synchrotron radiation in inhomogeneous tokamak plasmas
International Nuclear Information System (INIS)
Synchrotron emission in a tokamak configuration with inhomogeneous plasma parameters is considered to investigate the effects of the temperature profile and vertical elongation on the radiation loss. Using the numerical solution of the transfer equation for ITER-like plasma parameters, several new results on the radiated energy in a Maxwellian plasma have been derived. In particular: (i) synchrotron loss is profile dependent, namely, at constant average thermal energy, the emitted radiation increases with the peak temperature, (ii) an analytical formula of the global loss in inhomogeneous tokamak plasmas with arbitrary vertical elongation is established, (iii) the maximum of the frequency emission spectrum is a linear function of the volume average temperature, (iiii) high frequency synchrotron radiation is entirely due to electrons with energy much greater than the thermal energy. The need for experimental investigations on synchrotron emission in present-day large tokamaks to determine the effect of reflections of the complex tokamak first wall is stressed
Control of a burning tokamak plasma
Energy Technology Data Exchange (ETDEWEB)
Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.
1993-03-01
This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.
Impact of magnetic perturbation fields on tokamak plasmas
International Nuclear Information System (INIS)
Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.
Energy Technology Data Exchange (ETDEWEB)
Gutierrez T, C.; Beltran P, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)
2004-07-01
The neoclassical theory of transport is used to calculate the current efficiency of electronic cyclotron impulsion (ECCD) in an axisymmetric tokamak in the few collisions regime. The standard parameter of the tokamak is used to obtain a system of equations that describe the hydrodynamic of the plasma, where the ponderomotive force (PM) due to high power radio frequency waves is taken in account. The PM force is produced in the proximity of electron cyclotron resonance surface in a specific poloidal localization. The efficiency ECCD is analyzed in the cases of first and second harmonic (for different angles of injection of radio frequency waves) and it is validated using the experimental values of the TCV and T-10 tokamaks. The results are according to those obtained by means of the techniques of the Green functions. (Author)
Three novel tokamak plasma regimes in TFTR
International Nuclear Information System (INIS)
Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region
Three novel tokamak plasma regimes in TFTR
Energy Technology Data Exchange (ETDEWEB)
Furth, H.P.
1985-10-01
Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region.
Confinement of charged fusion products in reversed shear tokamak plasmas
International Nuclear Information System (INIS)
Full text: Recent tokamak studies indicate the attraction of operational scenarios with internal transport barriers (ITBs) that provide improved energy confinement with reversed shear (RS) in the plasma core. Whereas the presence of ITBs is beneficial to the energy confinement of the bulk plasma, RS is expected to deteriorate the confinement of fusion alphas (FA) in tokamaks with moderate plasma current, ∼2-3MA, due to enhanced first orbit and collisional loss. Experimentally, the influence of RS on the relaxation of the FA distribution function after NBI tritium blips into deuterium plasma has been observed recently in Trace Tritium Experiments on JET. In discharges with relatively high monotonic currents (>2MA) the observed FA density decay, was consistent with classical slowing down, while in 2.5MA strong RS discharges with a current hole ∼1/3 of the plasma radius the measured decay time was much shorter than the classical slowing down time, indicating a FA confinement degradation similar to that seen at 1MA current. Axisymmetric 3D Fokker-Planck modelling results presented confirm the confinement deterioration and the decay time decrease of FA distribution observed in RS JET discharges. (author)
Particle and energy balances in tokamak plasmas
International Nuclear Information System (INIS)
Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)
Turbulent and neoclassical toroidal momentum transport in tokamak plasmas
International Nuclear Information System (INIS)
The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are
International Nuclear Information System (INIS)
In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented
Plasma diagnostics using synchrotron radiation in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Fidone, I.; Giruzzi, G.; Granata, G.
1995-09-01
This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs.
Plasma-gun fueling for tokamak reactors
International Nuclear Information System (INIS)
In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment
Electron thermal transport in tokamak plasmas
International Nuclear Information System (INIS)
The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (108 K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called 'tokamak' this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high 'fusion' temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This 'anomalous transport' of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL)
Triangularity effects on the collisional diffusion for elliptic tokamak plasma
International Nuclear Information System (INIS)
In this conference the effect of ellipticity and triangularity will be analyzed for axisymmetric tokamak in the collisional regime. Analytic forms for the magnetic field cross sections are taken from those derived recently by other authors [1,2]. Analytical results can be obtained in elliptic plasmas with triangularity by using an special system of tokamak coordinates recently published [3-5]. Our results show that triangularities smaller than 0.6, increases confinement for ellipticities in the range 1.2 to 2. This behavior happens for negative and positive triangularities; however this effect is stronger for positive than for negative triangularities. The maximum diffusion velocity is not obtained for zero triangularity, but for small negative triangularities. Ellipticity is also very important in confinement, but the effect of triangularity seems to be more important. High electric inductive field increases confinement, though this field is difficult to modify once the tokamak has been built. The analytic form of the current produced by this field is like that of a weak Ware pinch with an additional factor, which weakens the effect by an order of magnitude. The dependence of the triangularity effect with the Shafranov shift is also analyzed. References 1. - L. L. Lao, S. P. Hirshman, and R. M. Wieland, Phys. Fluids 24, 1431 (1981) 2. - G. O. Ludwig, Plasma Physics Controlled Fusion 37, 633 (1995) 3. - P. Martin, Phys. Plasmas 7, 2915 (2000) 4. - P. Martin, M. G. Haines and E. Castro, Phys. Plasmas 12, 082506 (2005) 5. - P. Martin, E. Castro and M. G. Haines, Phys. Plasmas 12, 102505 (2005)
Boundary Plasma Turbulence Simulations for Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Xu, X; Umansky, M; Dudson, B; Snyder, P
2008-05-15
The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.
Energy Technology Data Exchange (ETDEWEB)
Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Institute of Plasma Physics National Science Center “Kharkov Institute of Physics and Technology” ul. Akademicheskaya 1, 61108 Kharkov (Ukraine); Kernbichler, Winfried; Martitsch, Andreas F.; Heyn, Martin F. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Maassberg, Henning [Max-Planck Institut für Plasmaphysik, D-17491 Greifswald (Germany)
2014-09-15
The toroidal torque driven by external non-resonant magnetic perturbations (neoclassical toroidal viscosity) is an important momentum source affecting the toroidal plasma rotation in tokamaks. The well-known force-flux relation directly links this torque to the non-ambipolar neoclassical particle fluxes arising due to the violation of the toroidal symmetry of the magnetic field. Here, a quasilinear approach for the numerical computation of these fluxes is described, which reduces the dimension of a standard neoclassical transport problem by one without model simplifications of the linearized drift kinetic equation. The only limiting condition is that the non-axisymmetric perturbation field is small enough such that the effect of the perturbation field on particle motion within the flux surface is negligible. Therefore, in addition to most of the transport regimes described by the banana (bounce averaged) kinetic equation also such regimes as, e.g., ripple-plateau and resonant diffusion regimes are naturally included in this approach. Based on this approach, a quasilinear version of the code NEO-2 [W. Kernbichler et al., Plasma Fusion Res. 3, S1061 (2008).] has been developed and benchmarked against a few analytical and numerical models. Results from NEO-2 stay in good agreement with results from these models in their pertinent range of validity.
Advanced tokamak burning plasma experiment
International Nuclear Information System (INIS)
A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
International Nuclear Information System (INIS)
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
Energy Technology Data Exchange (ETDEWEB)
HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M
2003-10-01
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance.
Dissipative nonlinear structures in tokamak plasmas
Directory of Open Access Journals (Sweden)
K. A. Razumova
2001-01-01
Full Text Available A lot of different kinds of instabilities may be developed in high temperature plasma located in a strong toroidal magnetic field (tokamak plasma. Nonlinear effects in the instability development result in plasma self-organization. Such plasma has a geometrically complicated configuration, consisting of the magnetic surfaces imbedded into each other and split into islands with various characteristic numbers of helical twisting. The self-consistency of the processes means that the transport coefficients in plasma do not depend just on the local parameters, being a function of the whole plasma configuration and of the forces affecting it. By disrupting the bonds between separate magnetic surfaces filled with islands, one can produce zones of reduced transport in the plasma, i.e. Ã¢Â€Âœinternal thermal barriersÃ¢Â€Â, allowing one essentially to increase the plasma temperature and density.
Modular Coils and Plasma Configurations for Quasi-axisymmetric Stellarators
Energy Technology Data Exchange (ETDEWEB)
L.P. Ku and A.H. Boozer
2010-09-10
Characteristics of modular coils for quasi-axisymmetric stellarators that are related to the plasma aspect ratio, number of field periods and rotational transform have been examined systematically. It is observed that, for a given plasma aspect ratio, the coil complexity tends to increase with the increased number of field periods. For a given number of field periods, the toroidal excursion of coil winding is reduced as the plasma aspect ratio is increased. It is also clear that the larger the coil-plasma separation is, the more complex the coils become. It is further demonstrated that it is possible to use other types of coils to complement modular coils to improve both the physics and the modular coil characteristics.
Electromagnetic effects of plasma disruptions in tokamaks
International Nuclear Information System (INIS)
The tokamak is modeled as typically 100 mutually-coupled toroidal circuits. The self and mutual inductances and the currents and voltages are calculated. Using the calculated currents, the poloidal magnetic field and the electromagnetic forces as functions of space and time are calculated. The major conclusion of the analysis is that the torus sectors should be electrically connected to each other near the plasma. Such connections reduce the structural loads, eliminate arcing, and reduce the induced potentials in the poloidal field coils
Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA
Indian Academy of Sciences (India)
D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team
2000-11-01
The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is ﬁrst tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.
Electron cyclotron emission from tokamak plasmas
International Nuclear Information System (INIS)
Emitted electron radiation can be used as a diagnostic signal to measure the electron temperature of a thermonuclear plasma. This type of diagnostics is well established in tokamak physics. In ch. 2 of this thesis the development, calibration and special design features are treated of a six-channel prototype of a twelve-channel grating spectrometer which is built for JET at Culham for electron cyclotron emission (ECE) measurements. In order to test this prototype measurements have been performed with the T-10 tokamak at the Kurchatov Institute in Moscow. With this prototype nearly half of the temperature profile of the T-10 could be measured. Detailed observations of sawteeth instabilities have been performed. Plasma heating by electron cyclotron resonance heating experiments was studied. A detailed description of these measurements and results is given in ch. 3. Often ECE spectra from tokamaks showed non-thermal features. In order to interprete them a computer code Notec has been developed. This code that calculates the ECE radiation emerging from the plasma for a 3-D configuration, is described in ch. 4. Some preliminary results and applications are presented. (Auth.)
Neoclassical plasma viscosity and transport processes in non-axisymmetric tori
Shaing, K. C.; Ida, K.; Sabbagh, S. A.
2015-11-01
Neoclassical transport processes are important to the understanding of plasma confinement physics in doubly periodic magnetized toroidal plasmas, especially, after the impact of the momentum confinement on the particle and energy confinement is recognized. Real doubly periodic tori in general are non-axisymmetric, with symmetric tori as a special case. An eight-moment approach to transport theory with plasma density N, plasma pressure p, mass flow velocity V and heat flow q as independent variables is adopted. Transport processes are dictated by the solutions of the momentum and heat flux balance equations. For toroidal plasma confinement devices, the first order (in the gyro-radius ordering) plasma flows are on the magnetic surface to guarantee good plasma confinement and are thus two-dimensional. Two linearly independent components of the momentum equation are required to determine the flows completely. Once this two-dimensional flow is relaxed, i.e. the momentum equation reaches a steady state, plasmas become ambipolar, and all the transport fluxes are determined through the flux-force relation. The flux-force relation is derived both from the kinetic definitions for the transport fluxes and from the manipulation of the momentum and heat flux balance equations to illustrate the nature of the transport fluxes by examining their corresponding driven forces and their roles in the momentum and heat flux balance equations. Steady-state plasma flows are determined by the components of the stress and heat stress tensors in the momentum and heat flux balance equations. This approach emphasizes the pivotal role of the momentum equation in the transport processes and is particularly useful in modelling plasma flows in experiments. The methodology for neoclassical transport theory is applied to fluctuation-driven transport fluxes in the quasilinear theory to unify these two theories. Experimental observations in tokamaks and stellarators for the physics discussed are
Plasma Physics Regimes in Tokamaks with Li Walls
International Nuclear Information System (INIS)
Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors
Spontaneous generation of rotation in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Parra Diaz, Felix [Oxford University
2013-12-24
Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.
Magnetic microtearing coherence in tokamak plasmas
International Nuclear Information System (INIS)
The analyses of the microtearing-modes coherence is effected. The tokamak characteristics, concerning fusion, electromagnetic confinement and turbulence are reviewed. The nature of the tearing modes, the variational principle of linear mode studies, a linear study in collisional and non-collisional plasma conditions are summarized, before studying the microtearing-mode coherence. The flux line configuration in the presence of a magnetic turbulence, the plasma response to a microtearing perturbation and instability, in the presence of a radial-electrons diffusion, is described. The autocoherence of microtearing modes in non-linear conditions are analyzed
Electron cyclotron emission imaging in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Munsat, Tobin; Domier, Calvin W.; Kong, Xiangyu; Liang, Tianran; Luhmann, Jr.; Neville C.; Tobias, Benjamin J.; Lee, Woochang; Park, Hyeon K.; Yun, Gunsu; Classen, Ivo. G. J.; Donne, Anthony J. H.
2010-07-01
We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the large-aperture optical systems and the linear detector arrays sensitive to millimeter-wavelength radiation. We present the status and recent progress on existing instruments as well as new systems under development for future experiments. We also discuss data analysis techniques relevant to plasma imaging diagnostics and present recent temperature fluctuation results from the tokamak experiment for technology oriented research (TEXTOR).
Linear wave propagation in a hot axisymmetric toroidal plasma
International Nuclear Information System (INIS)
Kinetic effects on the propagation of the Alfven wave are studied for the first time in a toroidal plasma relevant for experiments. This requires the resolution of a set of coupled partial differential equations whose coefficients depend locally on the plasma parameters. For this purpose, a numerical wave propagation code called PENN has been developed using either a bilinear or a bicubic Hermite finite element discretization. It solves Maxwell's equations in toroidal geometry, with a dielectric tensor operator that takes into account the linear response of the plasma. Two different models have been implemented and can be used comparatively to describe the same physical case: the first treats the plasma as resistive fluids and gives results which are in good agreement with toroidal fluid codes. The second is a kinetic model and takes into account the finite size of the Larmor radii; it has successfully been tested against a kinetic plasma model in cylindrical geometry. New results have been obtained when studying kinetic effects in toroidal geometry. Two different conversion mechanisms to the kinetic Alfven wave have been described: one occurs at toroidally coupled resonant surfaces and is the kinetic counterpart of the fluid models' resonance absorption. The other has no such correspondence and results directly from the toroidal coupling between the kinetic Alfven wave and the global wavefield. An analysis of a heating scenario suggests that it might be difficult to heat a plasma with Alfven waves up to temperatures that are relevant for a tokamak reactor. Kinetic effects are studied for three types of global Alfven modes (GAE, TAE, BAE) and a new class of kinetic eigenmodes is described which appear inside the fluid gap: it could be related to recent observations in the JET (Joint European Torus) tokamak. (author) 56 figs., 6 tabs., 58 refs
Tokamak Plasmas : Plasma position control in SST1 tokamak
Indian Academy of Sciences (India)
I Bandyopadhyay; S P Deshpande
2000-11-01
For long duration steady state operation of SST1, it would be very crucial to maintain the plasma radial and vertical positions accurately. For designing the position controller in SST1 we have adopted the simple linear RZIP control model. While the vertical position instability is slowed down by a set of passive stabilizers placed closed to the plasma edge, a pair of in-vessel active feedback coils can adequately control vertical position perturbations of up to 1 cm. The shifts in radial position arising due to minor disruptions would be controlled by a separate pair of poloidal ﬁeld (PF) coils also placed inside the vessel, however the controller would ignore fast but insigniﬁcant changes in radius arising due to edge localised modes. The parameters of both vertical and radial position control coils and their power supplies are determined based on the RZIP simulations.
Relativistic runaway electrons in tokamak plasmas
International Nuclear Information System (INIS)
Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)
Argonne Plasma Engineering Experiment (APEX) Tokamak
International Nuclear Information System (INIS)
The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials
Mathematical modeling plasma transport in tokamaks
International Nuclear Information System (INIS)
In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 1020/m3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%
Equilibrium reconstruction of plasma discharges in the Aditya Tokamak
International Nuclear Information System (INIS)
External magnetic measurements with flux loops and magnetic pick-up coils in tokamaks have provided vital information on the shape of the plasma column and also global current profile parameters, such as the sum of the poloidal beta (βp) and the internal inductance (ℓi). Such a reconstruction needs to be fast and sufficiently accurate such that it can be used routinely as a complementary input with other experimentally measured parameters for any sort of physics analysis of the plasma discharges. Here we present a method which can be used to proficiently reconstruct the current profile parameters, the plasma shapes, and a current density profile satisfying the MHD equilibrium constraint, reasonably conserving the external magnetic measurements. A Grad-Shafranov (GS) equation solver, named as IPREQ, has been developed in IPR to search for the best-fit current density profile. GS equation is a nonlinear elliptical differential equation describing axisymmetric toroidal equilibria. Ohmic transformer current (OT), vertical field coil current (BV) along with the plasma pressure (p) and current (Ip) profiles are used as inputs to the IPREQ code to reconstruct the equilibrium and the poloidal flux, plasma shape, βp and the safety factor (q) are inferred. At the four corners of the square cross-section vacuum vessel of Aditya, there are four magnetic pick-up coils aligned to measure the poloidal magnetic field (Bθ) during a plasma discharge. Further, there are two toroidal flux loops at the shadow of the limiter on the high field side to measure the loop voltage inside the vacuum vessel. Vacuum shots with OT and BV and no fill gas are used to calibrate these coils and loops. Measurement from these coils and flux loops are used to reconstruct the equilibrium consistently with the peak density and temperature measurements. Finally, the reconstructed equilibria are validated against the visible images from the fast visible imaging diagnostic on Aditya. (author)
Stabilization of axisymmetric mirror plasmas by energetic ion injection
International Nuclear Information System (INIS)
Plasmas in axisymmetric mirror devices can be made stable to MHD interchange modes by injecting energetic ions which contribute significantly to the pressure and spend a sufficiently large fraction of a bounce time in regions of favorable curvature. Pitch-angle scattering adversely affects the method by reducing this fraction. The ions must be sufficiently energetic that pitch-angle scattering is not detrimental for that part of a slowing-down time during which they contribute significantly to the pressure. We have solved the bounce-averaged Fokker-Planck equation, including drag and pitch-angle scattering, and calculated the energetic ion contribution to the stability integral. With specially tailored magnetic fields, the required injection energy and power drain are found to be reasonable
X-ray diagnostics of tokamak plasmas
International Nuclear Information System (INIS)
In this review, the authors venture into a new arena for work in atomic X-ray spectroscopy which can be dubbed tokamak X-ray spectroscopy diagnostics (TOXRASD). Not only is the experimental development exciting, but the measurements explore areas of atomic and plasma physics which have been inaccessible until just recently. Even though much of the present experimental effort is oriented towards obtaining a TOXRASD for fusion conditions, the new results touch upon some very basic atomic physics questions as well. Much effort has gone into the study of few electron systems, i.e., H-, He-, and Li-like ions with the purpose of utilizing the characteristic X-ray emission for diagnosing laboratory and astronomical plasmas or with the aim of developing the diagnostics, i.e., extending our understanding of the relationship between X-ray line emission and the plasma conditions under which the ions are formed and their ground states excited. However, the emission from highly charged heavy ions, such as neon-like molybdenum, has also attracted interest. Here the focus of the discussion is around results of the latest vintage from the TOXRASD project at the Alcator C tokamak at MIT. 38 references, 14 figures
Comparisons of linear and nonlinear plasma response models for non-axisymmetric perturbations
Energy Technology Data Exchange (ETDEWEB)
Turnbull, A. D.; Ferraro, N. M.; Lao, L. L.; Lanctot, M. J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Izzo, V. A. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Lazarus, E. A.; Hirshman, S. P. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, Tennessee 37831 (United States); Park, J.-K.; Lazerson, S.; Reiman, A. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States); Cooper, W. A. [Association Euratom-Confederation Suisse, Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Liu, Y. Q. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Turco, F. [Columbia University, 116th St and Broadway, New York, New York 10027 (United States)
2013-05-15
With the installation of non-axisymmetric coil systems on major tokamaks for the purpose of studying the prospects of ELM-free operation, understanding the plasma response to the applied fields is a crucial issue. Application of different response models, using standard tools, to DIII-D discharges with applied non-axisymmetric fields from internal coils, is shown to yield qualitatively different results. The plasma response can be treated as an initial value problem, following the system dynamically from an initial unperturbed state, or from a nearby perturbed equilibrium approach, and using both linear and nonlinear models [A. D. Turnbull, Nucl. Fusion 52, 054016 (2012)]. Criteria are discussed under which each of the approaches can yield a valid response. In the DIII-D cases studied, these criteria show a breakdown in the linear theory despite the small 10{sup −3} relative magnitude of the applied magnetic field perturbations in this case. For nonlinear dynamical evolution simulations to reach a saturated nonlinear steady state, appropriate damping mechanisms need to be provided for each normal mode comprising the response. Other issues arise in the technical construction of perturbed flux surfaces from a displacement and from the presence of near nullspace normal modes. For the nearby equilibrium approach, in the absence of a full 3D equilibrium reconstruction with a controlled comparison, constraints relating the 2D system profiles to the final profiles in the 3D system also need to be imposed to assure accessibility. The magnetic helicity profile has been proposed as an appropriate input to a 3D equilibrium calculation and tests of this show the anticipated qualitative behavior.
Investigation of Aditya Tokamak plasmas with lithiumized wall
International Nuclear Information System (INIS)
The lithium coating on plasma facing components of tokamak leads to better plasma properties through the reduction in impurities and controlling the hydrogen recycling. In Aditya tokamak, lithiumization of vacuum vessel wall is regularly carried out prior to its daily operation using lithium rod exposed to overnight glow discharge-cleaning plasma. Spectroscopic studies of Aditya tokamak plasmas shows the reduction of hydrogen (Hα at 656.3 nm) and oxygen (O II at 441.6 nm) as compared to discharges without the lithium coated walls. This clearly indicates reduction of recycling and impurity influxes from the wall, respectively. After Li coating, plasma stored energy increases significantly and plasmas with higher electron densities are obtained. Estimation of energy confinement time shows that it increases after lithimization and becomes comparable to the values predicated by Neo-Alcator scaling for ohmically heated tokamak plasma. Further analysis indicates that recycling must be low to achieve better plasma confinement. (author)
Theoretical issues on the spontaneous rotation of axisymmetric plasmas
International Nuclear Information System (INIS)
An extensive series of experiments have confirmed that the observed ‘spontaneous rotation’ phenomenon in axisymmetric plasmas is related to the confinement properties of these plasmas and connected to the excitation of collective modes associated with these properties (Coppi 2000 18th IAEA Fusion Energy Conf. (Sorrento, Italy, 2000) THP 1/17, www-pub.iaea.org/MTCD/publications/PDF/csp_008c/html/node343.htm and Coppi 2002 Nucl. Fusion 42 1). In particular, radially localized modes can extract angular momentum from the plasma column from which they grow while the background plasma has to recoil in the direction opposite to that of the mode phase velocity. In the case of the excitation of the plasma modes at the edge, the loss of their angular momentum can be connected to the directed particle ejection to the surrounding medium. The recoil angular momentum is then redistributed inside the plasma column mainly by the combination of an effective viscous diffusion and an inward angular momentum transport velocity that is connected, for instance, to ion temperature gradient (ITG) driven modes. The linear and quasi-linear theories of the collisionless trapped electron modes and of the toroidal ITG driven modes are re-examined in the context of their influence on angular momentum transport. Internal modes that produce magnetic reconnection and are electromagnetic in nature, acquire characteristic phase velocity directions in high temperature regimes and become relevant to the ‘generation’ of angular momentum. The drift-tearing mode, the ‘complex’ reconnecting mode and the m0 = 1 internal mode belong to this category, the last mode acquiring different features depending on the strength of its driving factor. Toroidal velocity profiles that reproduce the experimental observations are obtained considering a global angular momentum balance equation that includes the localized sources associated with the excited internal electrostatic and electromagnetic modes
Technique for plasma filament stabilization in a tokamak
International Nuclear Information System (INIS)
The invention is related to the field of automatic control of thermonuclear device processes and can be used in control systems of plasma filament stabilization by large radius in tokamak type thermolnuclear devices. The economic effect of the suggested technique is caused by improvement of stabilization of optimum (from the viewpoint of the decrease of plasma energy losses) plasma filament position in the tokamak-reactor which results in the decrease of power of additional plasma heating systems
Tokamak Plasmas : Internal magnetic ﬁeld measurement in tokamak plasmas using a Zeeman polarimeter
Indian Academy of Sciences (India)
M Jagadeeshwari; J Govindarajan
2000-11-01
In a tokamak plasma, the poloidal magnetic ﬁeld proﬁle closely depends on the current density proﬁle. We can deduce the internal magnetic ﬁeld from the analysis of circular polarization of the spectral lines emitted by the plasma. The theory of the measurement and a detailed design of the Zeeman polarimeter constructed to measure the poloidal ﬁeld proﬁle in the ADITYA tokamak are presented. The Fabry-Perot which we have employed in our design, with photodiode arrays followed by lock-in detection of the polarization signal, allows the measurement of the fractional circular polarization. In this system He-II line with wavelength 4686 Å is adopted as the monitoring spectral line. The line emission used in the present measurement is not well localized in the plasma, necessiating the use of a spatial inversion procedure to obtain the local values of the ﬁeld.
Sensitivity of transient synchrotron radiation to tokamak plasma parameters
Energy Technology Data Exchange (ETDEWEB)
Fisch, N.J.; Kritz, A.H.
1988-12-01
Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs.
Ion cyclotron emission in tokamak plasmas
International Nuclear Information System (INIS)
Detection of α(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, α particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. α particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central α density in a reactor. (author)
Tokamak simulation code manual
Energy Technology Data Exchange (ETDEWEB)
Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-01-01
The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.
Tokamak simulation code manual
International Nuclear Information System (INIS)
The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs
Neoclassical transport of impurtities in tokamak plasmas
International Nuclear Information System (INIS)
Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/2/n/sub H/e2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included
Integrated plasma control for high performance tokamaks
International Nuclear Information System (INIS)
Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)
Plasma transport in a Compact Ignition Tokamak
International Nuclear Information System (INIS)
Nominal predicted plasma conditions in a Compact Ignition Tokamak (CIT) are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models that have given almost equally good fits to experimental data. Using a transport model that best fits the data, thermonuclear ignition occurs in a CIT design with a major radius of 1.32 m, plasma half-width of 0.43 mn, elongation of 2.0, and toroidal field and plasma current ramped in 6 s from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the /sup 3/He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates have a large effect on ignition and on the maximum beta that can be achieved
Zeeman spectroscopy of tokamak edge plasmas
International Nuclear Information System (INIS)
Zeeman spectroscopy is a valuable tool both for diagnostic purposes, and for more fundamental studies of atomic and molecular processes in the boundary region of magnetically confined fusion plasmas (B ≅ 1 to 10 T). The method works well when the Zeeman (Paschen-Back) effect plays an important, or dominant, role in relation to other broadening mechanisms (Doppler, Stark, resonant excitation transfer) in determining the spectral line shape. For impurity species identification and temperature determination, Zeeman spectroscopy has advantages over charge-exchange recombination spectroscopy from highly excited radiator states, since spectral features practically unique to the species under investigation are analysed. It also provides useful information on probable mechanisms of line production (e.g. sputtering mechanisms, electron impact-induced dissociative excitation from molecules in the edge plasma), and on the temperature evolution of lower charge states in the process of convection inwards or diffusion outwards from the hotter plasma interior. Where different physical processes are responsible for different sections of the line profile -- especially in the case of hydrogen isotopes -- Zeeman spectroscopy can provide a set of characteristic temperatures for each section. The method is introduced in both passive and active spectroscopy, and general principles of the Zeeman effect are discussed with special reference to regimes of interest for the tokamak. Relevant physical processes (sputtering mechanisms, electron impact-induced dissociative excitation from molecules in the edge plasma, and ion-atom collisional heating mechanisms) are illustrated by sample spectra
Experimental studies of tokamak plasma in IPP Prague
International Nuclear Information System (INIS)
A short survey is given of the experimental activities at the small Prague tokamak CASTOR during recent years. At present, investigation is primarily aimed at the anomalous transport and plasma-wall interaction in the tokamak under conditions of combined OH/LHCD regimes. Moreover, some New diagnostic methods were also developed and certain improvements in the CASTOR performance were achieved. (author). 41 refs
Thomson Scattering on COMPASS Tokamak – Plasma Edge Profile
Czech Academy of Sciences Publication Activity Database
Böhm, Petr; Bílková, Petra; Aftanas, Milan; Štefániková, Estera; Mikulín, Ondřej; Melich, Radek; Janky, Filip; Havlíček, Josef; Šesták, David; Weinzettl, Vladimír; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Frassinetti, L.; Fassina, A.; Naylor, G.; Walsh, M.J.
Madison : UW Madison, 2013. [International Symposium Laser Aided Plasma Diagnostics/16./. Madison (US), 22.09.2013-26.09.2013] Institutional support: RVO:61389021 Keywords : tokamak * Thomson Scattering * H-mode * plasma profile pedestal * alignment Subject RIV: BL - Plasma and Gas Discharge Physics http://plasma.physics.wisc.edu/conferences/LAPD-2013/presentations/BohmLAPD16.pdf
Pumped limiter results on TFR Tokamak plasma
International Nuclear Information System (INIS)
Pump limiter experiments are carried out in the TFR Tokamak. The pump limiter is located in the outer part of the torus, its double- throat head is made of graphite tiles and it is pumped by a 2000 ls-1 titanium sublimation pump. The first attempts showed that the exhaust efficiency of this pump limiter was low (ε = 1.5% of the total plasma particle efflux). To improve these results, a new limiter head with a single longer throat has been built; particles were better trapped and the pumping provided an important decrease of the recycling coefficient. Geometric features mainly explain the increase by a factor 3.5 of the exhaust efficiency (ε = 5.5%). Ion temperature of the order of a few eV has been deduced from Doppler broadening measurements at the neutralizer plate of the pump limiter
Interactions of tokamak plasma with solid walls
International Nuclear Information System (INIS)
The interactions of tokamak fusion plasmas with solid walls of the devices were investigated on special model systems. The elastic recoil detection method was used for the determination of absolute hydrogen concentration. For the calibration of the method the scattering cross sections were measured in large ranges of scattering angle and energy. The erosion and deformation of wall surfaces were investigated by reemission of accelerated He ions. Theoretical models were developed to describe the surface undulation discovered earlier, caused by large dose He irradiation. The surface sputtering and segregation were investigated by nuclear methods and the mechanism of sputtering was simulated by computer. The surface deformation and gas reemission of Al surfaces were analyzed by Ar implementation and heat treatment. (D.Gy.) 6 figs
International Nuclear Information System (INIS)
In evaluating neoclassical transport by radially local simulations, the magnetic drift tangential to a flux surface is usually ignored in order to keep the phase-space volume conservation. In this paper, effect of the tangential magnetic drift on the local neoclassical transport is investigated. To retain the effect of the tangential magnetic drift in the local treatment of neoclassical transport, a new local formulation for the drift kinetic simulation is developed. The compressibility of the phase-space volume caused by the tangential magnetic drift is regarded as a source term for the drift kinetic equation, which is solved by using a two-weight δf Monte Carlo method for non-Hamiltonian system [G. Hu and J. A. Krommes, Phys. Plasmas 1, 863 (1994)]. It is demonstrated that the effect of the drift is negligible for the neoclassical transport in tokamaks. In non-axisymmetric systems, however, the tangential magnetic drift substantially changes the dependence of the neoclassical transport on the radial electric field Er. The peaked behavior of the neoclassical radial fluxes around Er = 0 observed in conventional local neoclassical transport simulations is removed by taking the tangential magnetic drift into account
Tokamak Plasmas : Observation of ﬂoating potential asymmetry in the edge plasma of the SINP tokamak
Indian Academy of Sciences (India)
Krishnendu Bhattacharyya; N R Ray
2000-11-01
Edge plasma properties in a tokamak is an interesting subject of study from the view point of conﬁnement and stability of tokamak plasma. The edge plasma of SINP-tokamak has been investigated using specially designed Langmuir probes. We have observed a poloidal asymmetry of ﬂoating potentials, particularly the top-bottom ﬂoating potential differences are quite noticeable, which in turn produces a vertical electric ﬁeld (v). This v remains throughout the discharge but changes its direction at certain point of time which seems to depend on applied vertical magnetic ﬁeld v).
Two-dimensional transport of tokamak plasmas
International Nuclear Information System (INIS)
A reduced set of two-fluid transport equations is obtained from the conservation equations describing the time evolution of the differential particle number, entropy, and magnetic fluxes in an axisymmetric toroidal plasma with nested magnetic surfaces. Expanding in the small ratio of perpendicular to parallel mobilities and thermal conductivities yields as solubility constraints one-dimensional equations for the surface-averaged thermodynamic variables and magnetic fluxes. Since Ohm's law E +u x B =R', where R' accounts for any nonideal effects, only determines the particle flow relative to the diffusing magnetic surfaces, it is necessary to solve a single two-dimensional generalized differential equation, (partial/partialt) delpsi. (delp - J x B) =0, to find the absolute velocity of a magnetic surface enclosing a fixed toroidal flux. This equation is linear but nonstandard in that it involves flux surface averages of the unknown velocity. Specification of R' and the cross-field ion and electron heat fluxes provides a closed system of equations. A time-dependent coordinate transformation is used to describe the diffusion of plasma quantities through magnetic surfaces of changing shape
TFTR/JET INTOR workshop on plasma transport tokamaks
International Nuclear Information System (INIS)
This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included
A control approach for plasma density in tokamak machines
International Nuclear Information System (INIS)
Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1
A control approach for plasma density in tokamak machines
Energy Technology Data Exchange (ETDEWEB)
Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)
2013-10-15
Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].
Experimental observations related to the thermodynamic properties of tokamak plasmas
International Nuclear Information System (INIS)
The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)
Kinetic extensions of magnetohydrodynamic models for axisymmetric toroidal plasmas
International Nuclear Information System (INIS)
A nonvariational kinetic-MHD stability code (NOVA-K) has been developed to integrate a set of non-Hermitian integro-differential eigenmode equations due to energetic particles for axisymmetric toroidal plasmas in a general flux coordinate system with an arbitrary Jacobian. The NOVA-K code employs the Galerkin method involving Fourier expansions in the generalized poloidal angle θ and generalized toroidal angle /zeta/ directions, and cubic-B spline finite elements in the radial /Psi/ direction. Extensive comparisons with the existing variational ideal MHD codes show that the ideal MHD version of the NOVA-K code converges faster and gives more accurate results. The NOVA-K code is employed to study the effects of energetic particles on MHD-type modes: the stabilization of ideal MHD internal kink modes and the excitation of ''fishbone'' internal kink modes; and the alpha particle destabilization of toroidicity-induced Alfven eigenmodes (TAE) via transit resonances. Analytical theories are also presented to help explain the NOVA-K results. For energetic trapped particles generated by neutral beam injection (NBI) or ion cyclotron resonant heating (ICRH), a stability window for the n = 1 internal kink mode in the hot particle beta space exists even in the absence of the core ion finite Larmor radius effect. On the other hand, the trapped alpha particles are found to have negligible effects on the stability of the n = 1 internal kink mode, but the circulating alpha particles can strongly destabilize TAE modes via inverse Landau damping associated with the spatial gradient of the alpha particle pressure. 60 refs., 24 figs., 1 tab
Measurements of plasma position in TJ-I Tokamak
Energy Technology Data Exchange (ETDEWEB)
Quin, J.; Ascasibar, E.; Navarro, A. P.; Ochando, M. A.; Pastor, I.; Pedrosa, M. A.; Rodriguez, L.; Sanchez, J.
1994-07-01
This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs.
Measurements of plasma position in TJ-I Tokamak
International Nuclear Information System (INIS)
This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs
Evaluation of the average ion approximation for a tokamak plasma
International Nuclear Information System (INIS)
The average ion approximation, sometimes used to calculated atomic processes in plasmas, is assessed by computing deviations in various rates over a set of conditions representative of tokamak edge plasmas. Conditions are identified under which the rates are primarily a function of the average ion charge and plasma parameters, as assumed in the average ion approximation. (Author) 19 refs., tab., 5 figs
JOINT EXPERIMENTS ON SMALL TOKAMAKS: EDGE PLASMA STUDIES ON CASTOR
Czech Academy of Sciences Publication Activity Database
Van Oost, G.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Del Bosco, E.; Dufková, Edita; Ďuran, Ivan; Gryaznevich, M.P.; Horáček, Jan; Hron, Martin; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferrera, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ossyannikov, A.; Singh, A.; Sokholov, M.; Talebitaher, A.
2007-01-01
Roč. 47, č. 5 (2007), s. 378-386. ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * edge plasma * turbulence * Langmuir probe * plasma radiation * Hall probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.278, year: 2007
Turbulent ion heating in TCV Tokamak plasmas
International Nuclear Information System (INIS)
The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active
Generation of plasma rotation by ICRH in tokamaks
International Nuclear Information System (INIS)
A physical mechanism to generate plasma rotation by ICRH is presented in a tokamak geometry. By breaking the omnigenity of resonant ion orbits, ICRH can induce a non-ambipolar minor-radial flow of resonant ions. This induces a return current jpr in the plasma, which then drives plasma rotation through the jprxB force. It is estimated that the fast-wave power in the present-day tokamak experiments can be strong enough to give a significant modification to plasma rotation. (author)
Intrinsic Plasma Rotation Determined by Neoclassical Toroidal Plasma Viscosity in Tokamaks
International Nuclear Information System (INIS)
Full text: Intrinsic steady state plasma rotation is important for plasma confinement in ITER, since the momentum input is expected to be small. It is well known that the intrinsic plasma rotation in stellarators is determined by non-ambipolar diffusion due to helical ripple. The non-ambipolar diffusion due to small 3D magnetic perturbation described by the Neoclassical Toroidal plasma Viscosity (NTV) theory may be important in determining the intrinsic plasma rotation in tokamaks, because the Non-Axisymmetric Magnetic Perturbations (NAMP) always exist. The NTV theory in different collisionality regimes has been well developed in the last few years, and it has been summarized in Ref. [1]. The numerical results showed a good agreement with the analytic solutions in different asymptotic limits. The intrinsic toroidal plasma rotation determined by the NTV effect in tokamaks is investigated in this paper. It is found by searching the root of the ambipolarity constraint Σj=i.e ejΓjΩΦ = 0, where ΩΦ is the toroidal rotation, and ej and Γj are the electric charge and the particle flux of the j (i = ions, and e = electrons) species. It is found that the result strongly depends on the plasma collisionality. In high collisionality case, the ion flux is dominant and the intrinsic steady state flow is in counter-current direction. It corresponds to the 'ion root' named in stellarators. In low collisionality case, there are three roots. One corresponds to the 'ion root' in counter-current direction. The second stable one corresponds to the 'electron root' in co-current direction, near which the electron flux is dominant. The third one is an unstable root. The NTV torque drives the plasma rotation towards one of the stable roots. This means that the intrinsic toroidal rotation in low collisionality case can also be possible in co-current direction. Both of these two roots scale like the diamagnetic frequency. The prediction of intrinsic rotation due to NTV on ITER will
Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bourdelle, C
2000-10-01
In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)
Lower hybrid current drive in tokamak plasmas
International Nuclear Information System (INIS)
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bare - 1020m-3, ALCATOR-C) and the highest current drive efficiency (ηCD = 3.5x1019 m-2A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
MHD stability analyses of a tokamak plasma by time-dependent codes
International Nuclear Information System (INIS)
The MHD properties of a tokamak plasma are investigated by using time evolutional codes. As for the ideal MHD modes we have analyzed the external modes including the positional instability. Linear and nonlinear ideal MHD codes have been developed. Effects of the toroidicity and conducting shell on the external kink mode are studied minutely by the linear code. A new rezoning algorithm is devised and it is successfully applied to express numerically the axisymmetric plasma perturbation in a cylindrical geometry. As for the resistive MHD modes we have developed nonlinear codes on the basis of the reduced set of the resistive MHD equations. By using the codes we have studied the major disruption processes and properties of the low n resistive modes. We have found that the effects of toroidicity and finite poloidal beta are very important. Considering the above conclusion we propose a new scenario of the initiation of the major disruption. (author)
A Midsize Tokamak As Fast Track To Burning Plasmas
International Nuclear Information System (INIS)
This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain ((ge) 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
A Midsize Tokamak As Fast Track To Burning Plasmas
Energy Technology Data Exchange (ETDEWEB)
E. Mazzucato
2010-07-14
This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
Energy Technology Data Exchange (ETDEWEB)
Fraboulet, D.
1996-09-17
Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.
Imaging System and Plasma Imaging on HL-2A Tokamak
Institute of Scientific and Technical Information of China (English)
郑银甲; 冯震; 罗萃文; 刘莉; 李伟; 严龙文; 杨青巍; 刘永
2004-01-01
As a new diagnostic means, plasma-imaging system has been developed on the HL2A tokamak, with a basic understanding of plasma discharge scenario of the entire torus, checking the plasma position and the clearance between the plasma and the first wall during discharge. The plasma imaging system consists of (1) color video camera, (2) observation window and turn mirror,(3) viewing & collecting optics, (4) video cable, (5) Video capture card as well as PC. This paper mainly describes the experimental arrangement, plasma imaging system and detailed part in the system, along with the experimental results. Real-time monitoring of plasma discharge process,particularly distinguishing limitor and divertor configuration, the imaging system has become key diagnostic means and laid the foundation for further physical experiment on the HL-2A tokamak.
Anomalous diffusion and transport beta limits in dense tokamak plasmas
International Nuclear Information System (INIS)
For tokamak plasmas which are sufficiently large and/or dense that the ionization source on axis may be neglected, particle balance is achieved by the inward diffusion due to the Ware pinch compensating the outward flow due to both neoclassical and anomalous diffusion. Insertion of measured data into the particle flux balance relation determines the anomalous particle diffusion coefficient Dsub(A); comparison of the results from a variety of tokamaks implies that the dominant dependence on machine and/or plasma parameters is Dsub(A) proportional to 1/n. Particle flux balance also implies an upper bound on the central value of βsub(e), the limiting value being obtained when the plasma parameters are chosen such that Dsub(A)<< Dsub(NEO). This limit has been computed for circular-cross-section tokamaks, and the results so obtained are of the same order as magnetohydrodynamic beta limits. (author)
Edge plasma studies on the CASTOR tokamak
International Nuclear Information System (INIS)
This paper will present an overview of recent edge plasma experiments on the CASTOR tokamak (R = 0.4 m, a = 0.085 m, BT = 1.3 T, Ip ∼ 10 kA, ne ∼ 2 x 1019 m-3). A poloidal array of 96 Langmuir probes, 16 magnetic coils, and 16 Hall sensors surrounding the full poloidal circumference monitors poloidal profiles of electric field, density, and magnetic field with high temporal resolution. A radial array of Langmuir probes measures the radial profiles of floating potential, poloidal electric field, and ion saturation current. A Gundestrup probe measures the parallel and perpendicular flows, a segmented tunnel probe measures the electron and ion temperatures. All data are acquired with 1 MHz sampling rate. First, an overview of the edge turbulence measurements will be given. Correlation analysis of the poloidal structure of the edge turbulence indicates that the behaviour of the scrape-off layer (SOL) turbulence is strongly linked to the magnetic field configuration in this region. The fluctuation measurements can be interpreted in a straightforward way assuming a single long structure, aligned with the magnetic field lines, which intersects a given poloidal cross section several times, giving rise to an apparent m = q mode. Next, behaviour of the plasma is presented for biasing experiments performed using a graphite electrode, immersed into the SOL and edge plasma, respectively. In the SOL, the measurements have shown presence of a region with long connection length (several toroidal turns), in which a biased flux tube is created. Thus, not only the radial field, but also a rather strong poloidal electric field is formed at the magnetic surface associated with the biased electrode. This poloidal field changes its sign periodically along the poloidal circumference and, as a consequence of the Epol x BT drift, the density is also poloidally modulated. The overall result is a creation of a pattern of particle flux with a strong poloidal modulation, which can be seen
Poloidal and parallel plasma viscosities in tokamak geometry
International Nuclear Information System (INIS)
The poloidal and parallel plasma viscosities in tokamak geometry in Hamada coordinates are calculated from the drift kinetic equation, including a large mass flow velocity without imposing the usual constraint that VpB/(vtiBp) be small. Here, Vp is the poloidal plasma flow velocity, vti is the ion thermal speed, B is the magnetic field strength, and Bp is the poloidal magnetic field strength. With this extended validity, the poloidal and parallel viscosities are useful in modeling the radial electric field in the edge region of a tokamak in the enhanced confinement regime
Influence of magnetic field ripple on plasma rotation in tokamaks
International Nuclear Information System (INIS)
The problem of plasma rotation in tokamaks with injection is considered. It is shown that the experimental results obtained in the PLT and ISX-B tokamaks can be explained by the presence of the magnetic-field ripple, provided that localized particles exist. The analysis carried out is based on the general expression for the magnetic-field ripple. This expression, which is derived in the present paper, describes the ripple distribution over the plasma cross-section for both the ordinary case and the case of shunted or damaged coils. (author)
Mechanisms of plasma disruption and runaway electron losses in tokamaks
Abdullaev, S. S.; Finken, K. H.; Wongrach, K.; Tokar, M.; Koslowski, H. R.; Willi, O.; Zeng, L.; Team, the TEXTOR
2015-01-01
Based on the analysis of data from the numerous dedicated experiments on plasma disruptions in the TEXTOR tokamak the mechanisms of the formation of runaway electron beams and their losses are proposed. The plasma disruption is caused by strong stochastic magnetic field formed due to nonlinearly excited low-mode number magnetohydrodynamic (MHD) modes. It is hypothesized that the runaway electron beam is formed in the central plasma region confined by an intact magnetic surface due to the acce...
Rotational soft x-ray tomography of noncircular tokamak plasmas
International Nuclear Information System (INIS)
A rotational tomography technique for noncircular tokamak plasmas has been developed. Using a linear transformation from an elliptic coordinate system to the circular one, and compensating for the Shafranov shift, the elliptic plasma shape is transformed to the concentric circular shape. Fitting the data of a quarter rotation to the Fourier--Bessel expansions, the tomography is performed. This technique is applied to the snake oscillation, to the slow sawtooth crash, and to the post-cursor oscillations of noncircular plasmas on JET
Nonlinear fusion dynamics in d-t tokamak plasmas
International Nuclear Information System (INIS)
Global fusion dynamics are examined for a d-t plasma confined in an ITER-like tokamak reactor. The analysis is based on the solution of a set of coupled nonlinear first-order differential equations describing the evolution of plasma density and plsma temperature as affected by the several particle and energy gain/loss mechanisms occurring in the burning d-t fusion plasma. (orig.)
Plasma edge biasing on CASTOR tokamak using LHCD
Czech Academy of Sciences Publication Activity Database
Žáček, František; Petržílka, Václav; Jakubka, Karel; Stöckel, Jan; Gunn, J.; Goniche, M.; Devynck, P.; Podesta, M.; Nanobashvili, S.
2001-01-01
Roč. 51, č. 10 (2001), s. 1129-1138. ISSN 0011-4626. [Europhysics Workshop on Role of Electric Fields in Plasma Confinement and Exhaust/4th./. Funchal, Madeira, 24.06.2001-25.06.2001] R&D Projects: GA AV ČR IAA1043101 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.345, year: 2001
Profile formation and sustainment of autonomous tokamak plasma with current hole configuration
International Nuclear Information System (INIS)
We have investigated the profile formation and sustainment of tokamak plasmas with the current hole (CH) configuration by using 1.5 dimensional time-dependent transport simulations. A model of the current limit inside the CH on the basis of the Axisymmetric Tri-Magnetic-Islands equilibrium is introduced into the transport simulation. We found that a transport model with the sharp reduction of anomalous transport in the reversed-shear (RS) region can reproduce the time evolution of profiles observed in JT-60U experiments. The transport becomes neoclassical-level in the RS region, which results in the formation of profiles with internal transport barrier (ITB) and CH. The CH plasma has an autonomous property because of the strong interaction between a pressure profile and a current profile through the large bootstrap current. The ITB width determined by the neoclassical-level transport agrees well with that measured in JT-60U. The energy confinement inside the ITB agrees with the scaling based on the JT-60U data. The scaling means that the core plasma inside ITB is governed by the MHD equilibrium limit, i.e. the autonomous limitation of energy confinement in the CH plasma. The plasma with the large CH is sustained with the full current drive by the bootstrap current. The plasma with the small CH and the small bootstrap current fraction shrinks due to the penetration of inductive current. This shrink is prevented and the CH size can be controlled by the appropriate external current drive (CD). The CH plasma is found to respond autonomically to the external CD. The application of the CH plasma to the tokamak reactor is discussed
Geodesic Acoustic Modes in Rotating Large Aspect Ratio Tokamak Plasmas
International Nuclear Information System (INIS)
Full text: Analytical theory of Geodesic Acoustic Modes (GAM's) is modified for a general case of rotating tokamak plasma. Both toroidal and poloidal components of steady-state plasma rotation are taken into account. For large aspect ratio tokamaks, the dispersion relation of electrostatic perturbations is derived analytically in the frame of one-fluid ideal magneto-hydrodynamics. In the case of small (compared to the sound frequency) angular rotation velocity, two solutions of dispersion relation are found. The first one is the standard GAM modified by the rotation effects. The second mode has a frequency close to the frequency of acoustic mode. The new GAM is induced by poloidal plasma rotation. This mode appears as a consequence of the Doppler frequency shift in the side-band components of plasma density, pressure and parallel velocity perturbations. The side-bands arise as the curvature driven response to the electrostatic potential perturbation with m =0(m is the poloidal wavenumber). The Doppler frequency shift is caused by poloidal rotation and has opposite signs for the m = 1 and m = -1 side-bands. Unlike the case of tokamak equilibrium with isothermal magnetic flux surfaces, no new low-frequency GAM arises in the case of purely toroidal plasma rotation in tokamak with isentropic magnetic surfaces. The pure toroidal flow results only in the up-shift of GAM frequency. (author)
International Nuclear Information System (INIS)
A set of reduced linear equations for the description of low-frequency perturbations in toroidally rotating plasma in axisymmetric tokamak is derived in the framework of ideal magnetohydrodynamics. The model suitable for the study of global geodesic acoustic modes (GGAMs) is designed. An example of the use of the developed model for derivation of the integral conditions for GGAM existence and of the corresponding dispersion relation is presented. The paper is dedicated to the memory of academician V.D. Shafranov
Energy Technology Data Exchange (ETDEWEB)
Lakhin, V. P.; Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com, E-mail: vilkiae@gmail.com; Ilgisonis, V. I. [National Research Centre Kurchatov Institute (Russian Federation); Konovaltseva, L. V. [Peoples’ Friendship University of Russia (Russian Federation)
2015-12-15
A set of reduced linear equations for the description of low-frequency perturbations in toroidally rotating plasma in axisymmetric tokamak is derived in the framework of ideal magnetohydrodynamics. The model suitable for the study of global geodesic acoustic modes (GGAMs) is designed. An example of the use of the developed model for derivation of the integral conditions for GGAM existence and of the corresponding dispersion relation is presented. The paper is dedicated to the memory of academician V.D. Shafranov.
Use of field-reversed compact plasma toroids for tokamak plasma make-up
International Nuclear Information System (INIS)
Main requirements to the parameters of compact plasma toroids, injection of which into tokamak plasma can be used for fuel make-up, are considered. The numeric modelling results attest that minimum disturbances of tokamak magnetic configuration can be expected when the injection direction is close to torus tangent line. In addition, an experimental device SAPFIR used for studying the formation of toroids of high value at plasma accelerators is described briefly. 7 refs.; 3 figs
Studies on fundamental technologies for producing tokamak-plasma
International Nuclear Information System (INIS)
The report describes studies on fundamental technologies to produce tokamak-plasma of the JFT-2 and JFT-2M tokamaks. (1) In order to measure the particle number of residual gases, calibration methods of vacuum gauges have been developed. (2) Devices for a Taylor-type discharge cleaning (TDC), a glow discharge cleaning (GDC) and ECR discharge cleaning (ECR-DC) have been made and the cleaning effects have been investigated. In TDC the most effective plasma for cleaning is obtained in the plasma with 5 eV of electron temperature. GDC is effective in removing carbon impurities, but is less effective for removing oxygen impurities. ECR-DC has nearly the similar effect as TDC. The cleaning effect of these three types were studied by comparing the properties of resulting tokamak plasmas in the JFT-2M tokamak. (3) Experimental studies of pre-ionization showed as following results; A simple pre-ionization equipment as a hot-electron-gun and a J x B gun was effective in reducing breakdown voltage. An ordinary mode wave of the electron cyclotron frequency was very effective for pre-ionization. The RF power whose density is 3.6 x 10-2 W/cm3 produced plasma of an electron density of 5 x 1011 cm-3. In this case, it is possible to start up with negligible consumption of the magnetic flux caused by the plasma resistance. (4) Concerning to studies on plasma control, the following results were obtained; In order to obtain constant plasma current, a pulse forming network was constructed and sufficient constant plasma current was achieved. In applying an iso-flux method for measuring the plasma position, it is no problem practically to use only one loop-coil and one magnetic probe. (author)
Energy Technology Data Exchange (ETDEWEB)
Alladio, F; Mancuso, A; Micozzi, P [Associazione Euratom-ENEA sulla Fusione, CP 65 Frascati, Roma (Italy)], E-mail: alladio@frascati.enea.it
2008-12-15
The filament state of a magnetic field is the usual way for plasmas to avoid magnetic inhibition of convective overturning. However, it requires dynamo conversion of kinetic into magnetic energy and is therefore often associated with a plasma velocity shear layer. In the sun, isolated current carrying magnetic filaments (twisted flux tubes) are produced by the solar dynamo from a continuous strong toroidal field, sitting just below the radiative-convective transition, on the Sun rotation shear layer (tachocline, r{sub Tach} {approx} 2 {center_dot} R{sub o-dot}/3 in terms of the solar radius, R{sub o-dot}). The twisted flux tubes, become buoyant; some of them fall back into the tachocline adding up to the continuous toroidal field; some emerge from the photosphere kinked and twisted, reconnect and produce flares. In the mode of high magnetic confinement (H-mode), when a magnetic separatrix bounds the axisymmetric tokamak discharge and a sheared plasma rotation is present, magnetic filaments with concentrated internal currents (edge localized modes) are produced near the velocity shear layer (pressure pedestal, at r{sub Ped} {>=} 0.94a{sub Sep} in terms of the minor radius of the plasma boundary, a{sub Sep}): again a dynamo conversion of kinetic into magnetic energy is required in order to filament the current density at the pedestal. The current carrying filaments break the unperturbed axisymmetric tokamak equilibrium, producing ergodicity in the edge plasma. The faster loss of energy from the ergodic plasma makes the rotating magnetic filaments outboards anti-buoyant: therefore they convect outboards from the pedestal. The anti-buoyancy and motion model for the tokamak case is compared with the buoyancy model for the Sun.
Calculation of plasma position and shape in KT-1 Tokamak
Energy Technology Data Exchange (ETDEWEB)
Chang, Doo Hee; Chung Kyu Sun [Hanyang Univ., Seoul (Korea, Republic of); Oh, Byung Hoon; Jeong, Seung Ho; Hong, Bong Guen; Lee, Kwang Won [KAERI, Taejon (Korea, Republic of)
2000-05-01
For the first time, the full-time scale variations of plasma position and shape from the outmost poloidal isoflux surfaces during the buildup, flat-top, and decay period of plasma current are calculated from the measurement of magnetic fields at the plasma-boundary surface in a KT-1 Tokamak (major radius of 27 cm and minor radius of 4.2 cm). Three kinds of magnetic probes arranged poloidally in 30 degrees outside the torus are used in determining the plasma boundary. Without the information of internal plasma current profiles, the calculations are only performed by a linear combination of measured fields.
Spatially resolved soft x-ray spectroscopy of tokamak plasmas
International Nuclear Information System (INIS)
We describe the space-resolved soft x-ray (1-33nm) instrumentation developed for the Tore Supra tokamak. By using a programmable hydraulic jack to move the spectrometer, several spatial profiles (up to ten) of many impurity lines are obtained during a single plasma discharge, with a time resolution which can be as short as 600 ms. (author)
Modelling multi-ion plasma gun simulations of Tokamak disruptions
International Nuclear Information System (INIS)
The effect of impurity ions in plasma gun ablation tests of various targets is considered. Inclusion of reasonable amounts of impurity (∼10%) is adequate to explain observed energy transmission and erosion measurements. The gun tests and the computer code calculations are relevant to the parameter range expected for major disruptions on large tokamaks
Profile formation and sustainment of autonomous tokamak plasma with current hole configuration
International Nuclear Information System (INIS)
We have investigated the profile formation and sustainment of tokamak plasmas with the current hole (CH) configuration by using 1.5D time-dependent transport simulations. A model of the current limit inside the CH on the basis of the Axisymmetric Tri-Magnetic-Islands equilibrium is introduced into the transport simulation. We found that a transport model with the sharp reduction of anomalous transport in the reversed-shear (RS) region can reproduce the time evolution of profiles observed in JT-60U experiments. The transport becomes neoclassical-level in the RS region, which results in the formation of profiles with internal transport barrier (ITB) and CH. The CH plasma has an autonomous property because of the strong interaction between a pressure profile and a current profile through the large bootstrap current fraction. The ITB width determined by the neoclassical-level transport agrees well with that measured in JT-60U. The energy confinement inside the ITB agrees with the scaling based on the JT-60U data. The scaling means the autonomous limitation of energy confinement in the CH plasma. The plasma with the large CH is sustained with the full current drive by the bootstrap current. The plasma with the small CH and the small bootstrap current fraction shrinks due to the penetration of inductive current. This shrink is prevented and the CH size can be controlled by the appropriate external current drive (CD). The CH plasma is found to respond autonomically to the external CD. (author)
A penalization technique to model plasma facing components in a tokamak with temperature variations
Energy Technology Data Exchange (ETDEWEB)
Paredes, A.; Bufferand, H.; Ciraolo, G.; Schwander, F. [Aix Marseille Universite, CNRS, Centrale Marseille, M2P2 UMR 7340, 13451 Marseille (France); Serre, E., E-mail: eric.serre@L3m.univ-mrs.fr [Aix Marseille Universite, CNRS, Centrale Marseille, M2P2 UMR 7340, 13451 Marseille (France); Ghendrih, P.; Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)
2014-10-01
To properly address turbulent transport in the edge plasma region of a tokamak, it is mandatory to describe the particle and heat outflow on wall components, using an accurate representation of the wall geometry. This is challenging for many plasma transport codes, which use a structured mesh with one coordinate aligned with magnetic surfaces. We propose here a penalization technique that allows modeling of particle and heat transport using such structured mesh, while also accounting for geometrically complex plasma-facing components. Solid obstacles are considered as particle and momentum sinks whereas ionic and electronic temperature gradients are imposed on both sides of the obstacles along the magnetic field direction using delta functions (Dirac). Solutions exhibit plasma velocities (M=1) and temperatures fluxes at the plasma–wall boundaries that match with boundary conditions usually implemented in fluid codes. Grid convergence and error estimates are found to be in agreement with theoretical results obtained for neutral fluid conservation equations. The capability of the penalization technique is illustrated by introducing the non-collisional plasma region expected by the kinetic theory in the immediate vicinity of the interface, that is impossible when considering fluid boundary conditions. Axisymmetric numerical simulations show the efficiency of the method to investigate the large-scale transport at the plasma edge including the separatrix and in realistic complex geometries while keeping a simple structured grid.
Design of plasma facing components for the SST-1 tokamak
International Nuclear Information System (INIS)
Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m2. (author)
Dust-Particle Transport in Tokamak Edge Plasmas
Energy Technology Data Exchange (ETDEWEB)
Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D
2005-09-12
Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.
Thermal Event Recognition Applied to Tokamak Protection during Plasma Operation
Martin, Vincent; Bremond, François; Travere, Jean-Marcel; Moncada, Victor; Dunand, Gwenaël
2009-01-01
Magnetic confinement fusion reactors are complex devices where a large amount of energy is required to make the fusion reactions happen. In such experimental conditions, the Plasma Facing Components (PFC) are subjected to high heat fluxes. In current tokamaks like Tore Supra, infrared thermographic diagnostics based on image analysis and feedback control are used to measure and monitor the heating of the PFC during plasma operation. The system consists in detecting high increase of the IR lum...
Measurement of the internal magnetic field structure of tokamak plasmas
International Nuclear Information System (INIS)
The first part of this article deals with the physical fundaments and technical aspects of this polarimetric measuring method, with its diagnostic capability, but also with its limitations. The second part summarizes the essential experimental results and their feedback on the theoretical description of Tokamak plasmas, which caused a revision of the accepted ideas of the magnetic field structure and its magnetohydrodynamic stability, in particular in the area of the hot plasma core. (orig.)
Optimization of tokamak plasma equilibrium control in TOKASTAR-2
International Nuclear Information System (INIS)
To improve capability of tokamak plasma equilibrium control in TOKASTAR-2, a pair of pulsed vertical field coils (PVF coils) was installed in 2012. Optimization of PVF coil current waveform was made with re-calculation using TOSCA code in varying the conditions of tokamak discharge. Then, simulation of the electric circuit of PVF coil system with LTspice code was done to find the optimum capacitance and charging voltage of capacitor for PVF coils consistent to the optimal current waveform. Eventually capacitor parameters for PVF coils to be used as base for experiment were obtained. (author)
Negative edge plasma currents in the SINP tokamak
Indian Academy of Sciences (India)
Ramesh Narayanan; A N Sekar Iyengar
2011-12-01
A tokamak plasma discharge having an increase in duration accompanied with enhanced runaway electron ﬂux has been experimentally studied in this paper. The discharges have been obtained by controlling the applied vertical magnetic ﬁeld ($B^{\\text{appl}}_v$) to below a critical value. Such discharges have been observed to have ‘negative edge plasma currents’, detected using an internal Rogowskii coil (IRC). We have tried to correlate the runaway behaviour with the negative edge plasma currents and have explained that these observations are a result of beam plasma instabilities.
Energy Technology Data Exchange (ETDEWEB)
Sarazin, Y
2004-03-01
This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.
Solenoid-free plasma start-up in spherical tokamaks
Raman, R.; Shevchenko, V. F.
2014-10-01
The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.
Solenoid-free plasma start-up in spherical tokamaks
International Nuclear Information System (INIS)
The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid. (topical review)
Emission spectroscopy of ECRH non-axisymmetric helium mirror plasmas
International Nuclear Information System (INIS)
In this experiment emission spectroscopy in the 3000 to 5000 Angstrom range has been utilized to determine the electron temperature of helium plasmas produced by the Michigan Mirror Machine (MIMI). The plasma is generated and heated by whistler-mode electron-cyclotron resonance waves at 7.43 GHz and 500 Watts power in 400 μs pulses. Gas is puffed into the mid-plane region where a quartz window is used to observe the plasma. The plasma is viewed in a direction perpendicular to the mirror axis
Heating of plasmas in tokamaks by current-driven turbulence
International Nuclear Information System (INIS)
Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued
New techniques for inducing plasma current in large tokamak devices
International Nuclear Information System (INIS)
Up to 70% of the plasma current in large tokamak devices can be induced by the flux swing of the equilibrium field (EF), if the plasma pressure is increased very early during start-up and maintained at the MHD-limiting value throughout the current rise. In an illustration of this technique, a 4.4-MA plasma current is induced by the EF coils assisted by a set of ''leaky OH'' coils and a small central solenoid. In a second example, current induction by the EF is supplemented by a 2.0-MA neutral-beam-driven current, and the central transformer is eliminated entirely. A hybrid coil is described, wherein the TF and OH coils are merged into a single winding with greatly reduced net forces. These techniques can alleviate space restriction near the major axis in small-R tokamak reactors
Transition to subcritical turbulence in a tokamak plasma
van Wyk, F; Schekochihin, A A; Roach, C M; Field, A R; Dorland, W
2016-01-01
Unstable perturbations driven by the pressure gradient and other sources of free energy in tokamak plasmas can grow exponentially and eventually saturate nonlinearly, leading to turbulence. Recent work has shown that in the presence of sheared flows, such systems can be subcritical. This means that all perturbations are linearly stable and a transition to a turbulent state only occurs if large enough initial perturbations undergo sufficient transient growth to allow nonlinear interaction. There is, however, currently very little known about a subcritical transition to turbulence in fusion-relevant plasmas. Here we use first-principles gyrokinetic simulations of a turbulent plasma in the outer core of the Mega-Ampere Spherical Tokamak (MAST) to demonstrate that the experimentally observed state is near the transition threshold, that the turbulence in this state is subcritical, and that transition to turbulence occurs via accumulation of very long-lived, intense, finite-amplitude coherent structures, which domi...
Formulation of two-dimensional transport modeling in tokamak plasmas
International Nuclear Information System (INIS)
A two-dimensional transport modeling applicable to a whole tokamak plasma is proposed. The model is derived from the multi-fluid equations and Maxwell's equations and the moment approach of neoclassical transport is employed as fluid closures. The multi-fluid equations consist of the equations for particle density, momentum, energy and total heat flux transport for each plasma species. The expressions of the parallel viscosity and heat viscosity are extended in order to be applicable to both inside and outside of the last closed flux surface. It is confirmed that our neoclassical transport model is consistent with the ordinary flux-surface-averaged one-dimensional neoclassical transport model. Our transport equations are coupled with the electromagnetic equations in order to describe the time evolution of tokamak plasmas. The procedure for coupling a transport solver based on our transport model with an equilibrium solver is also briefly described. (author)
Instrumentation for plasma diagnosis in TN (Novillo Tokamak)
International Nuclear Information System (INIS)
In the Plasma Physics Laboratory of National Institute of Nuclear Research it has been utilized different devices for to determine electromagnetic parameters of Novillo Tokamak such as: magnetic fields, plasma currents, plasma column position and hoop voltage. For these measurements it was designed, constructed and calibrated magnetic soundings such as: magnetic field soundings, Rogowsky coil, coils of the type called sine/cosine and spires type riding saddle; as well as the electronic instrumentation associated with these devices. This electronics to be clear of instrumentation amplifiers for the detection of the soundings signals and differentiators utilized for the elimination of spurious induced currents in the soundings by the different Novillo electromagnetic fields. In this work is presented the methodology for the construction of this instruments, as well as the results of measurements effectuated in the two operation regimens of Tokamak: Cleaning discharge and Main discharge. (Author)
Extremely shaped plasmas to improve the Tokamak concept
International Nuclear Information System (INIS)
experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is
Plasma Shape and Current Control Simulation of HT-7U Tokamak
Institute of Scientific and Technical Information of China (English)
吴斌; 张澄
2003-01-01
This paper describes the discharge simulation of HT-7U tokamak plasma equilibriumand plasma current by solving MHD equations and surface average transport equations using anequilibrium evolution code. The simulated result shows the evolution of plasma parameter versustime .The simulated result can play an important role in the design of the plasma equilibrium andcontrol system of a tokamak.
ECRH current drive in tokamak plasmas
International Nuclear Information System (INIS)
The current drive by electron cyclotron resonance heating (ECRH) is investigated in a typical magnetic field of tokamak with circular cross section. The trapped electrons and the modification of electron-cyclotron resonance condition by the relativistic mass increase are shown to have significant effects on the efficiency of this current drive. The efficiency strongly depends on the values of the parallel velocity u0 of resonant electrons, the inverse aspect ratio ε, the poloidal angle θ0 of absorption point, and the relativistic parameter S, which represents the strength of the relativistic correction to the resonance condition. (author)
Inward energy transport in tokamak plasmas
International Nuclear Information System (INIS)
Peaked electron temperature profiles are observed in the DIII-D tokamak during electron cyclotron heating despite the fact that >75% of the input power is deposited significantly off axis. Power balance analysis indicates a net inward flow of energy for electrons. An inward energy flow is not compatible with diffusive or critical gradient models. A time-dependent perturbation technique is employed to estimate the conductive loss and the nondiffusive part of the energy transport. The nondiffusive component of the transport appears only at radii smaller than that of the heating location
Low-n shear Alfven spectra in axisymmetric toroidal plasmas
International Nuclear Information System (INIS)
In toroidal plasmas, the toroidal magnetic field is nonuniform over a magnetic surface and causes coupling of different poloidal harmonics. It is shown both analytically and numerically that the toroidicity not only breaks up the shear Alfven continuous spectrum, but also creates new, discrete, toroidicity-induced shear Alfven eigenmodes with frequencies inside the continuum gaps. Potential applications of the low-n toroidicity-induced shear Alfven eigenmodes on plasma heating and instabilities are addressed. 17 refs., 4 figs
Plasma radiation in tokamak disruption simulation experiments
International Nuclear Information System (INIS)
Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 106 cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation
Characterization of the Tokamak de Varennes ohmic plasma in equilibrium
International Nuclear Information System (INIS)
Experimental results obtained on the Tokamak de Varennes during a series of reproducible ohmic discharges are presented. Profiles of basic plasma parameters are constructed and compared with theoretical predictions. In particular, the measured plasma resistivity agrees with the neoclassical scaling rather than with Spitzer resistivity. The study of electron-density fluctuations indicates a linear dispersion relation with a propagation velocity of 3.0 x 104 cm s-1. Particle transport investigations are initiated, giving experimental diffusion and convection coefficients across the plasma radius for electrons and impurity ions. (Author) 32 refs., 10 figs
Tokamak plasma shape identification based on the boundary integral equations
International Nuclear Information System (INIS)
A necessary condition for tokamak plasma shape identification is discussed and a new identification method is proposed in this article. This method is based on the boundary integral equations governing a vacuum region around a plasma with only the measurement of either magnetic fluxes or magnetic flux intensities. It can identify various plasmas with low to high ellipticities with the precision determined by the number of the magnetic sensors. This method is applicable to real-time control and visualization using a 'table-look-up' procedure. (author)
A midsize tokamak as a fast track to burning plasmas
Directory of Open Access Journals (Sweden)
E. Mazzucato
2011-03-01
Full Text Available This paper describes the conceptual design of a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥ 10 with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER. This can be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a different magnetic divertor from those currently employed in present experiments is discussed.
Sawtooth driven particle transport in tokamak plasmas
International Nuclear Information System (INIS)
The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author)
Control strategy for plasma equilibrium in a tokamak
International Nuclear Information System (INIS)
The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)
Study of a disruption mitigation method for tokamak plasmas
International Nuclear Information System (INIS)
Disruptions are a sudden loss of confinement of a tokamak plasma which take place in around 20 ms. They may lead to severe damaging of the tokamak structure, through heat deposition on Plasma Facing Components, electromagnetic stresses and relativistic runaway electrons. On future reactors, disruption mitigation will be critical. Massive gas injection is one of the methods proposed to mitigate disruptions. It was studied both experimentally and numerically in the thesis. Experiments on the Tore Supra and JET tokamaks showed that light gases (helium) were able to suppress runaway electrons. They induce a large density build-up which is large enough to suppress runaway production. On the contrary, heavier gases should be able to radiate more of the plasma thermal energy, but generate runaway electrons. All gases reduce electromagnetic forces. Gas mixtures have also been tested successfully to combine the advantages of the two types of gas. The gas jet penetration is linked to MHD instabilities enhancing the radial transport of the ionized gas, but preventing the neutrals from penetrating further inside a critical MHD surface. Massive gas injection simulations have been carried out using the 3D MHD code Jorek, by adding a neutral fluid model to the code. Results show that MHD instabilities are triggered more rapidly with high amounts of gas, and that successive rational surfaces are ergodized by the penetration of the density front in the plasma, in agreement with experimental observations. (author)
Field simulation of axisymmetric plasma screw pinches by alternating-direction-implicit methods
International Nuclear Information System (INIS)
An axisymmetric plasma screw pinch is an axisymmetric column of ionized gaseous plasma radially confined by forces from axial and azimuthal currents driven in the plasma and its surroundings. This dissertation is a contribution to detailed, high resolution computer simulation of dynamic plasma screw pinches in 2-d rz-coordinates. The simulation algorithm combines electron fluid and particle-in-cell (PIC) ion models to represent the plasma in a hybrid fashion. The plasma is assumed to be quasineutral; along with the Darwin approximation to the Maxwell equations, this implies application of Ampere's law without displacement current. Electron inertia is assumed negligible so that advective terms in the electron momentum equation are ignored. Electrons and ions have separate scalar temperatures, and a scalar plasma electrical resistivity is assumed. Altemating-direction-implicit (ADI) methods are used to advance the electron fluid drift velocity and the magnetic fields in the simulation. The ADI methods allow time steps larger than allowed by explicit methods. Spatial regions where vacuum field equations have validity are determined by a cutoff density that invokes the quasineutral vacuum Maxwell equations (Darwin approximation). In this dissertation, the algorithm was first checked against ideal MM stability theory, and agreement was nicely demonstrated. However, such agreement is not a new contribution to the research field. Contributions to the research field include new treatments of the fields in vacuum regions of the pinch simulation. The new treatments predict a level of magnetohydrodynamic turbulence near the bulk plasma surface that is higher than predicted by other methods
Impurity transport in a collision-dominated rotating tokamak plasma
International Nuclear Information System (INIS)
The flux of heavy impurities is an axisymmetric, toroidal plasma with all particles in the collision-dominated regime is considered. Plasma rotation and charge-exchange with neutrals are taken into account. A hydrodynamic model employing Braginskii's transport equations is used. The theorry is extended to higher collision freqencies as compared to previous treatments. It is found that the Pfirsch-Schlueter flux is significantly reduced as compared to the value given by Rutherford and that it is of the same order of magnitude, or less, than the classical flux in all regimes considered. It is also shown that the impurity flux can be influenced by charge-exchange with neutrals. (author)
Influence of plasma surface interactions on tokamak startup
International Nuclear Information System (INIS)
The startup phase of a tokamak is a complex phenomenon involving burnthrough of the low-Z impurities and rampup of Ip, the plasma current. The design considerations of a tokamak are closely connected with the startup modeling. Plasma evolution is analysed using a zero-dimensional model. The particle and energy balance is considered of two subclasses of plasmas which are penetrable by neutral gas, together with another component, neutrals trapped in the wall. The first subclass includes plasmas being penetrated by slow neutrals of (∼few eV) temperature. The second includes plasmas being penetrated only by fast neutrals having a temperature comparable to that of the ions. The impact of impurities on energy balance is considered through their generation by ion induced desorption of adsorbed oxygen on the first wall and physical and chemical sputtering of carbon. The paper demonstrates self-consistently that the evolution of initial phase of the discharge is intimately linked to the condition of the plasma facing components (PFCs) and the resultant plasma surface interactions
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Castracane, J.
2001-01-04
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
International Nuclear Information System (INIS)
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies
Numerical studies of transport processes in Tokamak plasma
International Nuclear Information System (INIS)
The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included
Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions
International Nuclear Information System (INIS)
Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)
Efficient trap of a coaxial gun plasma in an axisymmetric mirror with an internal hoop
International Nuclear Information System (INIS)
A method to trap a high temperature and high density plasma from a coaxial gun in a mirror machine is described. The method is to inject plasma parallel to the axis from a coaxial gun located off the axis. The validity of the method is experimentally demonstrated with an MHD-stabilized axisymmetric mirror with an internal hoop. Density, electron and ion temperatures and their time behaviors were measured and it was made clear that a high density high temperature plasma was well trapped in the mirror by the parallel off-axis injection while the plasma was little trapped by on-axis injection. The plasma parameters obtained were also compared with those of a conventional titanium washer gun plasma. The causes to restrict the maximum ion temperature and of its quick decay are discussed. (author)
Ray refraction in a fully-ionized axisymmetrical plasma
International Nuclear Information System (INIS)
Analytical and numerical methods are used to obtain the approximate formulas for ray refraction angle when passing through plasma with exponential dip of density profile. In the region of small values of the refraction angle its dependence on electron density in the point of ray turn is described by exponential function with index, close to one, and the value is inversely proportional to the square of frequency of incident radiation. 6 refs.; 5 figs.; 2 tabs
Time-resonant tokamak plasma edge instabilities?
Webster, A. J.; Dendy, R. O.; Calderon, F. A.; Chapman, S. C.; Delabie, E.; Dodt, D.; Felton, R.; Todd, T. N.; Maviglia, F.; Morris, J.; Riccardo, V.; Alper, B.; Brezinsek, S.; Coad, P.; Likonen, J.; Rubel, M.; JET-EFDA Contributors,
2014-01-01
For a two week period during the Joint European Torus 2012 experimental campaign, the same high confinement plasma was repeated 151 times. The dataset was analysed to produce a probability density function (pdf) for the waiting times between edge-localized plasma instabilities (ELMs). The result was
Plasma diagnostics for tokamaks and stellarators
International Nuclear Information System (INIS)
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Sattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma
Pellet-plasma interactions in tokamaks
DEFF Research Database (Denmark)
Chang, C.T.
1991-01-01
The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experiment...
Plasma diagnostics for tokamaks and stellarators
Energy Technology Data Exchange (ETDEWEB)
Stott, P. E.; Sanchez, J.
1994-07-01
A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs.
International Nuclear Information System (INIS)
The first direct observation of the internal structure of driven global Alfven eigenmodes in a tokamak plasma is presented. A carbon dioxide laser scattering/interferometer has been designed, built, and installed on the PRETEXT tokamak. By using this diagnostic system in the interferometer configuration, we have for the first time, thoroughly investigated the resonance conditions required for, and the spatial wave field structure of, driven plasma eigenmodes at frequencies below the ion cyclotron frequency in a confined, high temperature, tokamak plasma
Surface temperature measurement of plasma facing components in tokamaks
International Nuclear Information System (INIS)
During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author)
Influence of error fields on the plasma confining field and the plasma confinement in tokamak
International Nuclear Information System (INIS)
Influence of error fields on the plasma confining field and the plasma confinement is treated in the standpoint of design. In the initial breakdown phase before formation of the closed magnetic surfaces, the vertical field properly applied is the most important. Once the magnetic surfaces are formed, the non-axisymmetric error field is important. Effect of the shell gap associated with iron core and with pulsed vertical coils is thus studied. The formation of magnetic islands due to the external non-axisymmetric error field is studied with a simple model. A method of suppressing the islands by choosing the minor periodicity is proposed. (auth.)
To the scenario choosing of plasma density limit reaching in the CIT tokamak
International Nuclear Information System (INIS)
Results of numerical simulation of reaching of maximum permissible plasma density in CIT tokamak are given. Scenario optimization principles that are not connected with specific plasma confinement model are formulated; it is important if real conditions (heat transfer, diffusion, convection etc) in CIT tokamak plasma would be different from one that are considered in the simulation. 10 refs.; 3 figs
Liquid gallium jet-plasma interaction studies in ISTTOK tokamak
International Nuclear Information System (INIS)
Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma contamination by liquid metal. Additionally the response of an infrared sensor, intended to measure the jet surface temperature increase during its interaction with the plasma, has been studied. The jet power extraction capability is extrapolated from the heat flux profiles measured in ISTTOK plasmas.
Offline Development of Plasma Boundary Controllers for the KSTAR Tokamak
Ballinger, S.; Eidietis, N. W.; Humphreys, D. A.; Hyatt, A. W.; Welander, A. S.; Hahn, S. H.
2014-10-01
The KSTAR TokSys tokamak simulator, implemented in Matlab®/Simulink, has been extended to include a plasma boundary control system to allow automated offline tuning of shape control feedback loops. Offline control development minimizes resources expended tuning controllers during actual run time, and automated tuning is desirable in order to optimize the large number of shape control gains. The new simulation includes simplified versions of the rtEFIT/Isoflux controller used in the KSTAR plasma control system, allowing full-closed-loop analysis of the plasma shape control. Results presented include application of robust design methods to optimizing control of KSTAR's plasma boundary, and analysis to understand observed differences in boundary control between KSTAR and other superconducting devices. Work supported in part by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US Department of Energy under DE-FC02-04ER54698.
International Nuclear Information System (INIS)
Burning plasma simulation in tokamak (TR), spherical tokamak (ST) and helical (HR) reactors were carried out focusing on Internal Transport Barrier (ITB) plasma operations using the TOTAL-T (Toroidal Transport Analysis Linkage - Tokamak) code coupled with GLF23 turbulent transport code and NCLASS neoclassical transport codes, and TOTAL-H (Helical) code with multi-helicity helical ripple transport analysis code. The effectiveness of these ITB transport coefficients is checked using experimental data of JT-60U and LHD. It clarified the requirement of deep penetration of high-field-side (HFS) pellet injection fueling to realize steady-state advanced burning operation in TR and ST. The neoclassical ripple transport plays an important role on the ITB operation in HR. Moreover, economical and environmental assessments were performed for these three type reactors by the PEC (Physics Engineering and Cost) system code in the case of four blanket designs (Li/V, Flibe/FS(Ferritic Steel), LiPb/SiC, FF(Fission- Fusion) Hybrid). In the present analysis, maximum field of superconducting coil is assumed 13 T, instead of maximum normal conductor strength of 8T in ST reactor. The tolerable neutron wall fluence is assumed 20 MW.Yr/m"2 in the case of LiPb/SiC blanket system, which determines the replacement cycle of blanket modules. As for cost analysis, the fusion island (FI) cost of ST-1 is lowest. However, its fusion thermal power is largest and the TR is superior in cost of electricity (COE). Among four blanket designs Flibe/FS is superior in cost, because ferritic steel (FS) is much cheaper than vanadium (V). The life-cycle CO2 emission amount per output electric power and the energy payback ratio are also evaluated. The ST reactor is favorable in CO2 emission reduction, because rather compact and simple normal conducting coil system is adopted here. The ST and TR need more frequent blanket exchanges than HR with lower neutron wall load. However, HR is still expensive and has
Electron temperature gradient driven instability in the tokamak boundary plasma
International Nuclear Information System (INIS)
A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t-1/2eγmt
Plasma Boundary Reconstruction using Fast Camera on the COMPASS Tokamak
Czech Academy of Sciences Publication Activity Database
Háček, Pavel; Berta, Miklós; Stöckel, Jan; Weinzettl, Vladimír; Budai, C.; Szabolics, T.; Bencze, A.
Prague: MATFYZPRESS, 2014 - (Šafránková, J.; Pavlů, J.), s. 221-226. (WDS). ISBN 978-80-7378-276-4. [Annual Conference of Doctoral Students – WDS 2014 /23./. Prague (CZ), 03.06.2014-05.06.2014] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics http://www.mff.cuni.cz/veda/konference/wds/proc/pdf14/WDS14_39_f2_Hacek.pdf
Plasma facing components design of KT-2 tokamak
International Nuclear Information System (INIS)
The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs
Eikonal waves, caustics and mode conversion in tokamak plasmas
Jaun, A.; Tracy, E. R.; Kaufman, A. N.
2007-01-01
Ray optics is used to model the propagation of short electromagnetic plasma waves in toroidal geometry. The new RAYCON code evolves each ray independently in phase space, together with its amplitude, phase and focusing tensor to describe the transport of power along the ray. Particular emphasis is laid on caustics and mode conversion layers, where a linear phenomenon splits a single incoming ray into two. The complete mode conversion algorithm is described and tested for the first time, using the two space dimensions that are relevant in a tokamak. Applications are shown, using a cold plasma model to account for mode conversion at the ion-hybrid resonance in the Joint European Torus.
Spectra of heliumlike krypton from Tokamak Fusion Test Reactor plasmas
International Nuclear Information System (INIS)
Experiments were conducted on TFTR to study the radiation of krypton which will be important for future tokamaks, such as ITER, for the diagnostic of the central ion temperature and for the control of the energy release from the plasma by radiative cooling. The total krypton radiation was monitored, and satellite spectra of Kr XXXV were recorded with a high-resolution crystal spectrometer. Radiative cooling and reduced particle recycling at the plasma edge region were observed, in reasonable agreement with modeling calculations which included radial transport
Evaluation of the plasma parameters in COMPASS tokamak divertor area
Czech Academy of Sciences Publication Activity Database
Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud
2012-01-01
Roč. 356, č. 1 (2012), s. 012007. ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf
Role of ECRH in SST-1 tokamak plasma
International Nuclear Information System (INIS)
In SST-1, the Electron Cyclotron resonance Heating (ECRH) system has been used extensively to carry out various experiments related to fundamental and second harmonic ECRH assisted breakdown and start-up of tokamak. The ECRH further contributes to plasma current during long pulse operation. The 42 GHz ECRH system delivers 500 kW power for 500 ms duration and corrugated waveguide (∅ 63.5mm) based transmission line at normal atmospheric pressure is used to launch power in HE11 mode. The mirror based launcher is used to launch focused beam in SST-1 plasma. Since the loop voltage of SST-1 is low (∼3.0 V), the ECRH assisted start-up is mandatory for reliable plasma discharges. In the beginning of each plasma campaign, it is observed that impurity dominates results in small discharges. In such cases ECRH is used at higher power for long duration to overcome the impurity burn-through and get the good plasma discharges. In SST-1, the ECRH is also used to drive some current to support plasma current. The ECRH power from 150 kW to 350 kW has been launched in fundamental O-mode and second harmonic X-mode. The ECRH is used for short pulse ∼ 80 to 120 ms (for breakdown and start-up) and long pulse duration up to 430ms (for start-up as well as support plasma current with electron cyclotron current drive ECCD). As the first pass absorption of ECRH is not good in breakdown phase (at low density and temperature), the ECRH power transmits up to inboard side wall of tokamak. At the inboard side, a profiled reflector is installed at an angle to launch focused beam from high field side with an angle to toroidal magnetic field (BT). This is similar to co-injection launch of ECRH power in X-mode from high field side to support plasma current with ECCD. The experiments show that the plasma current profile is different in two cases (ECRH short pulse and long pulse). In the long pulse ECRH, the plasma current profile is smooth with some increase (∼ 5 to 10%) in plasma current, which
Pellet-plasma interactions in tokamaks
DEFF Research Database (Denmark)
Chang, C.T.
1991-01-01
The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental...... data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy...... confinement time, offset by the accumulation of impurities at the plasma core is brought into focus. A possible remedy is suggested to diminish the effect of the impurities. Plausible arguments are presented to explain the apparent controversial observations on the propagation of a fast cooling front ahead of...
Protection of tokamak plasma facing components by a capillary porous system with lithium
Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.
2015-08-01
Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.
Real-time optical plasma boundary reconstruction for plasma position control at the TCV Tokamak
Hommen, G.; de M. Baar,; Duval, B. P.; Andrebe, Y.; Le, H. B.; Klop, M. A.; Doelman, N. J.; Witvoet, G.; Steinbuch, M.; TCV team,
2014-01-01
A dual, high speed, real-time visible light camera setup was installed on the TCV tokamak to reconstruct optically and in real-time the plasma boundary shape. Localized light emission from the plasma boundary in tangential view, broadband visible images results in clearly resolved boundary edge-feat
Emissive probe measurements of plasma potential fluctuations in the edge plasma regions of tokamaks
Czech Academy of Sciences Publication Activity Database
Balan, P.; Schrittweiser, R.; Ionita, C.; Cabral, J. A.; Figueiredo, H. F. C.; Fernandes, H.; Varandas, C.; Adámek, Jiří; Hron, Martin; Stöckel, Jan; Martines, E.; Tichý, M.; Van Oost, G.
2003-01-01
Roč. 74, č. 3 (2003), s. 1583-1587. ISSN 0034-6748 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma physics, tokamaks, probes Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.343, year: 2003
Experimental studies on an axisymmetric divertor in DIVA(JFT-2a)
International Nuclear Information System (INIS)
DIVA(JFT-2a) is the first tokamak with an axisymmetric divertor in the world. Objectives of the experiments were i) Plasma production and confinement in a tokamak with a separatrix magnetic surface, and ii) divertor effects on radiation loss and plasma confinement. The results so far are as follows: i) The equilibrium with a separatrix magnetic surface is stable during the discharge. ii) There is an ergodic region near the separatrix magnetic surface due to non-axisymmetric magnetic perturbations. iii) The divertor reduces radiation loss and increases energy confinement time. iv) The divertor does not affect the transport process in the main plasma. (author)
A quasi-linear gyrokinetic transport model for tokamak plasmas
International Nuclear Information System (INIS)
After a presentation of some basics around nuclear fusion, this research thesis introduces the framework of the tokamak strategy to deal with confinement, hence the main plasma instabilities which are responsible for turbulent transport of energy and matter in such a system. The author also briefly introduces the two principal plasma representations, the fluid and the kinetic ones. He explains why the gyro-kinetic approach has been preferred. A tokamak relevant case is presented in order to highlight the relevance of a correct accounting of the kinetic wave-particle resonance. He discusses the issue of the quasi-linear response. Firstly, the derivation of the model, called QuaLiKiz, and its underlying hypotheses to get the energy and the particle turbulent flux are presented. Secondly, the validity of the quasi-linear response is verified against the nonlinear gyro-kinetic simulations. The saturation model that is assumed in QuaLiKiz, is presented and discussed. Then, the author qualifies the global outcomes of QuaLiKiz. Both the quasi-linear energy and the particle flux are compared to the expectations from the nonlinear simulations, across a wide scan of tokamak relevant parameters. Therefore, the coupling of QuaLiKiz within the integrated transport solver CRONOS is presented: this procedure allows the time-dependent transport problem to be solved, hence the direct application of the model to the experiment. The first preliminary results regarding the experimental analysis are finally discussed
International Nuclear Information System (INIS)
The tokamak edge plasma region begins beyond the middle plasma, limited by a diaphragm and spread to torus vacuum chamber wall. Parameters of edge plasma have been measured; several disgnostic type have been used. Numerical simulation code is used for result interpretarion and to display important phenomena in this region. Simulation results give a relation between the plasma parameters at the limiter radius; these parameters can be used as limit conditions for inner plasma transport codes. Edge plasma measurements have been examined with care during lower hybrid frequency heating. Study of plasma parameter modifications can help to a better comprehension of phenomena related to heating
The COMPASS Tokamak Plasma Control Software Performance
Czech Academy of Sciences Publication Activity Database
Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír
2011-01-01
Roč. 58, č. 4 (2011), s. 1490-1496. ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726
Equilibrium calculations for plasma control in CIT [Compact Ignition Tokamak
International Nuclear Information System (INIS)
The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs
Molecular emission in the edge plasma of T-10 tokamak
International Nuclear Information System (INIS)
The experiments on recording molecular emission in the edge plasma of the T-10 tokamak are described. To obtain reliable spectra with sufficient spectral, temporal, and spatial resolution, the optical circuit is optimized for various experimental conditions. Typical spectra measured in two sections of the tokamak are shown. It is shown that, upon varying the parameters of the discharge, the molecular spectrum not only changes significantly in intensity but also undergoes a qualitative change in the rotational and vibrational structure. For a detailed analysis, we use the Fulcher-α system (d3Πu–a3Σg+) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X1Σg+ and upper excited state d3Πu are estimated by the measured spectra
Molecular emission in the edge plasma of T-10 tokamak
Energy Technology Data Exchange (ETDEWEB)
Zimin, A. M., E-mail: zimin@power.bmstu.ru [Bauman Moscow State Technical University (Russian Federation); Krupin, V. A. [National Research Centre Kurchatov Institute (Russian Federation); Troynov, V. I. [Bauman Moscow State Technical University (Russian Federation); Klyuchnikov, L. A. [National Research Centre Kurchatov Institute (Russian Federation)
2015-12-15
The experiments on recording molecular emission in the edge plasma of the T-10 tokamak are described. To obtain reliable spectra with sufficient spectral, temporal, and spatial resolution, the optical circuit is optimized for various experimental conditions. Typical spectra measured in two sections of the tokamak are shown. It is shown that, upon varying the parameters of the discharge, the molecular spectrum not only changes significantly in intensity but also undergoes a qualitative change in the rotational and vibrational structure. For a detailed analysis, we use the Fulcher-α system (d{sup 3}Π{sub u}–a{sup 3}Σ{sub g}{sup +}) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X{sup 1}Σ{sub g}{sup +} and upper excited state d{sup 3}Π{sub u} are estimated by the measured spectra.
Helium distribution functions in tokamak plasmas
International Nuclear Information System (INIS)
Two different methods are used to obtain information on the helium distribution. The first method is a machine that measures the velocity distribution of neutral helium particles escaping from the plasma (NPA). The second method is charge exchange spectroscopy that measures the helium density and temperature as a function of time and place from the Doppler broadened intensity of He lines. (orig./HP)
Tokamak plasma current disruption infrared control system
Kugel, H.W.; Ulrickson, M.
1984-04-16
This invention is directed to the diagnosis and detection of gross or macroinstabilities in a magnetically-confined fusion plasma device. Detection is performed in real time, and is prompt such that correction of the instability can be initiated in a timely fashion.
Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas
Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira
2010-11-01
As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.
Measurement of the effective plasma ion mass in large tokamaks
International Nuclear Information System (INIS)
There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs
Analysis of fast ion induced instabilities in tokamak plasmas
Horváth, László
2015-01-01
In magnetic confinement fusion devices like tokamaks, it is crucial to confine the high energy fusion-born helium nuclei ($\\alpha$-particles) to maintain the energy equilibrium of the plasma. However, energetic ions can excite various instabilities which can lead to their enhanced radial transport. Consequently, these instabilities may degrade the heating efficiency and they can also cause harmful power loads on the plasma-facing components of the device. Therefore, the understanding of these modes is a key issue regarding future burning plasma experiments. One of the main open questions concerning energetic particle (EP) driven instabilities is the non-linear evolution of the mode structure. In this thesis, I present my results on the investigation of $\\beta$-induced Alfv\\'{e}n eigenmodes (BAEs) and EP-driven geodesic acoustic modes (EGAMs) observed in the ramp-up phase of off-axis NBI heated plasmas in the ASDEX Upgrade tokamak. These modes were well visible on several line-of-sights (LOSs) of the soft X-ra...
Field simulation of axisymmetric plasma screw pinches by alternating-direction-implicit methods
Energy Technology Data Exchange (ETDEWEB)
Lambert, M.A.
1996-06-01
An axisymmetric plasma screw pinch is an axisymmetric column of ionized gaseous plasma radially confined by forces from axial and azimuthal currents driven in the plasma and its surroundings. This dissertation is a contribution to detailed, high resolution computer simulation of dynamic plasma screw pinches in 2-d {ital rz}-coordinates. The simulation algorithm combines electron fluid and particle-in-cell (PIC) ion models to represent the plasma in a hybrid fashion. The plasma is assumed to be quasineutral; along with the Darwin approximation to the Maxwell equations, this implies application of Ampere`s law without displacement current. Electron inertia is assumed negligible so that advective terms in the electron momentum equation are ignored. Electrons and ions have separate scalar temperatures, and a scalar plasma electrical resistivity is assumed. Altemating-direction-implicit (ADI) methods are used to advance the electron fluid drift velocity and the magnetic fields in the simulation. The ADI methods allow time steps larger than allowed by explicit methods. Spatial regions where vacuum field equations have validity are determined by a cutoff density that invokes the quasineutral vacuum Maxwell equations (Darwin approximation). In this dissertation, the algorithm was first checked against ideal MM stability theory, and agreement was nicely demonstrated. However, such agreement is not a new contribution to the research field. Contributions to the research field include new treatments of the fields in vacuum regions of the pinch simulation. The new treatments predict a level of magnetohydrodynamic turbulence near the bulk plasma surface that is higher than predicted by other methods.
Oscillating field current drive and nonlinear plasma response in bootstrapped tokamaks
International Nuclear Information System (INIS)
The nonlinear response of bootstrapped tokamaks to oscillating field current drive (OFCD) is studied with the aid of a formal mathematical solution. The tokamak plasma response to OFCD is shown to be similar to that of a driven inductor--resistor (LR) circuit, but the effective plasma resistance can be negative. A physical picture is constructed in which the tokamak bootstrap effect is viewed as a nonlinear amplification of the plasma magnetic helicity. The bootstrap amplifier can rectify the toroidal plasma current. The importance of the tokamak thermal instability for peaking the plasma current and creating a plasma dynamo is stressed. Estimates are made which suggest that steady-state tokamak fusion reactors operating with OFCD may be achievable with moderate amplitude toroidal magnetic flux and loop voltage oscillations and oscillation driving frequencies significantly below the audible frequency range. copyright 1995 American Institute of Physics
Self-organized ignition of a tokamak plasma
International Nuclear Information System (INIS)
The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product nτET, where τE represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual fusion power deposition [4,5] and its effect of reducing τE are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself - towards an ignited stable
Modeling plasma/material interactions during a tokamak disruption
International Nuclear Information System (INIS)
Disruptions in tokamak reactors are still of serious concern and present a potential obstacle for successful operation and reliable design. Erosion of plasma-facing materials due to thermal energy dump during a disruption can severely limit the lifetime of these components, therefore diminishing the economic feasibility of the reactor. A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The initial burst of energy delivered to facing-material surfaces from direct impact of plasma particles causes sudden ablation of these materials. As a result, a vapor cloud is formed in front of the incident plasma particles. Shortly thereafter, the plasma particles are stopped in the vapor cloud, heating and ionizing it. The energy transmitted to the material surfaces is then dominated by photon radiation. It is the dynamics and the evolution of this vapor cloud that finally determines the net erosion rate and, consequently, the component lifetime. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics, and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed
Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER
Razumova, K. A.; Andreev, V. F.; Kislov, A. Y.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Y. D.; Shafranov, T. V.; Donne, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; R. Jaspers,; Kantor, M.; Walsh, M.
2009-01-01
The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile ha
Generation and diagnostics of fast electrons within Tokamak plasmas
International Nuclear Information System (INIS)
The first part of this invited paper is devoted to mechanisms of the production of fast electrons in plasma experiments involving magnetic traps of the tokamak type. The phenomenon of generation of the so-called runaway electrons - which may reach energies up to several dozen MeV - is considered and basic characteristics of such electrons are described. In particular, the orbits of the runaway electrons and their energy limits are presented. Problems related to the cross-field transport in tokamak plasmas and interactions of the relativistic electrons with plasma oscillations are also considered. Production of the so-called ripple-born electrons, which may be observed in the energy range from approximately 50 keV to several hundreds keV, is analyzed separately. In the second part of this paper various diagnostic methods used for investigation of the runaway and ripple-born electrons are presented. Various techniques are described, which are based on different reactions induced by the runaway electrons, e.g., the emission of X-rays or neutrons, or the synchrotron radiation. Finally, a modern technique of electron measurements is described, which was developed by the author's team at the Andrzej Soltan Institute for Nuclear Studies (IPJ) in Swierk (Poland), and which is based on the use of Cherenkov-type detectors. Examples of applications of the discussed techniques in different tokamak experiments are described. Particular attention is paid to the Cherenkov detectors, which have already been used in experiments at the small-size CASTOR device in Prague (Czech Republic), the ISTTOK machine in Lisbon (Portugal), as well as in the larger TORE-SUPRA facility in Cadarache (France). (author)
The evolution of the plasma current during tokamak disruptions
International Nuclear Information System (INIS)
In a tokamak disruption, the ohmic plasma current is partly replaced by a current carried by runaway electrons. This process is analysed by combining the equations for runaway electron generation with Maxwell's equations for the evolution of the electric field. This allows a quantitative understanding to be gained of runaway production in present experiments, and extrapolation to be made to ITER. The runaway current typically becomes more peaked on the magnetic axis than the pre-disruption current. In fact, the central current density can rise although the total current falls, which may have implications for post-disruption plasma stability. Furthermore, it is found that the runaway current easily spreads radially in a filament way due to the high sensitivity of the runaway generation efficiency to plasma parameters. (authors)
Trade studies of plasma elongation for next-step tokamaks
Energy Technology Data Exchange (ETDEWEB)
Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.
1988-09-01
The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs.
Plasma production in a tokamak with force-balanced coils
International Nuclear Information System (INIS)
We manufactured a small pulsed tokamak with force balanced coils (FBCs), which balance the net centering force and the net radial hoop force due to the poloidal and toroidal current components, respectively. The centering force was demonstrated to be reduced by an order of magnitude compared with the computed one of the TF coils of the same dimension. The plasma current up to 10 kA was achieved by two-step FBC magnetization since the force-balanced winding provides not only toroidal magnetic fields but also the poloidal magnetic flux to induce the plasma current. The plasma column was well centered in the vacuum vessel within the time constant of shell effects of the vessel. (author)
Forbidden line emission from highly ionized atoms in tokamak plasmas
Feldman, U.; Doschek, G. A.; Bhatia, A. K.
1982-01-01
Considerable interest in the observation of forbidden spectral lines from highly ionized atoms in tokamak plasmas is related to the significance of such observations for plasma diagnostic applications. Atomic data for the elements Ti Cr, Mn, Fe, Ni, and Kr have been published by Feldman et al. (1980) and Bhatia et al. (1980). The present investigation is concerned with collisional excitation rate coefficients and radiative decay rates, which are interpolated for ions of elements between calcium, and krypton and for levels of the 2s2 2pk, 2s 2p(k+1), and 2p(k+2) configurations, and for the O I, N I, C I, B I, and Be I isoelectronic sequences. The provided interpolated atomic data can be employed to calculate level populations and relative line intensities for ions of the considered sequences, taking into account levels of the stated configurations. Important plasma diagnostic information provided by the forbidden lines includes the ion temperature
Line-integrated emissivity in tokamak plasma radiation diagnostics
International Nuclear Information System (INIS)
In tokamak plasma radiation diagnostics, an emissivity profile derived from line-integrated radiation measurement data with Abel inversion or tomographic analysis is averaged over a poloidal cross-section of the plasma chord. The width of this cross-section restricts the spatial resolution of the given diagnostic system. In the cases where the condition for infinitesimal-detector approximation is not satisfied, this width appears to be markedly greater than that defined by the angle viewed from the detector center to the poloidal dimension of an aperture, and the contributions to the detected signal from different points on the given cross-section of the plasma chord are not the same, so the average of the emissivity over the cross-section is a weighted one. In fact, many radiation diagnostic systems on tokamaks do not satisfy the infinitesimal-detector approximation condition. On the basis of the derived explicit formula for the line-integrated emissivity, the relevant questions are discussed in detail
Kinetic integrated modeling of plasma heating in tokamaks
International Nuclear Information System (INIS)
Plasma heating and current drive by waves, neutral beam, and fusion reaction cause deformation of momentum distribution function of heated species. The deformation affects the heating, the loss mechanism, and the radial transport of energetic ions. The fusion reaction rate and the propagation of RF electric field is also influenced by the deformation. Therefore self-consistent heating analysis including the deformation of distribution function is required for quantitative analysis. In the present analysis, the deformation of the distribution function and the effect of radial diffusion during heating in tokamak plasmas are studied using integrated tokamak modeling code TASK and its components TASK/FP and TASK/WM. The newly extended bounce averaged Fokker-Planck component TASK/FP implements relativistic and energy conservative collision model, and it enables us to solve the Fokker-Planck equation for each species simultaneously. Using the integrated code TASK, multi-species plasma heating is analyzed and the effect of radial diffusion models are examined. (author)
Neoclassical Physics for Current Drive in Tokamak Plasmas
International Nuclear Information System (INIS)
The Lie transform formalism is applied to charged particle dynamics in tokamak magnetic topologies, in order to build a Fokker-Planck type operator for Coulomb collisions usable for current drive. This approach makes it possible to reduce the problem to three dimensions (two in velocity space, one in real space) while keeping the wealth of phase-space cross-term coupling effects resulting from conservation of the toroidal canonical momentum (axisymmetry). This kinetic approach makes it possible to describe physical phenomena related to the presence of strong pressure gradients in plasmas of an unspecified form, like the bootstrap current which role will be paramount for the future ITER machine. The choice of coordinates and the method used are particularly adapted to the numerical resolution of the drift kinetic equation making it possible to calculate the particle distributions, which may present a strong variation with respect to the Maxwellian under the effect of an electric field (static or produced by a radio-frequency wave). This work, mainly dedicated to plasma physics of tokamaks, was extended to those of space plasmas with a magnetic dipole configuration. (author)
Multimode observations and 3D magnetic control of the boundary of a tokamak plasma
Levesque, J. P.; Rath, N.; Shiraki, D.; Angelini, S.; Bialek, J.; Byrne, P. J.; DeBono, B. A.; Hughes, P. E.; Mauel, M. E.; Navratil, G. A.; Peng, Q.; Rhodes, D. J.; Stoafer, C. C.
2013-07-01
We present high-resolution detection and control of the 3D magnetic boundary in the High Beta Tokamak-Extended Pulse (HBT-EP) device. Measurements of non-axisymmetric radial and poloidal fields are made using 216 magnetic sensors positioned near the plasma surface. Control of 3D fields is accomplished using 40 independent saddle coils attached to the passive stabilizing wall. The control coils are energized with high-power solid-state amplifiers, and massively parallel, high-throughput feedback control experiments are performed using low-latency connections between PCI Express analogue input and output modules and a graphics processing unit. The time evolution of unstable and saturated wall-stabilized external kink modes are studied with and without applying magnetic perturbations using the control coils. The 3D dynamic structure of the magnetic field surrounding the plasma is determined through biorthogonal decomposition using the full set of magnetic sensors without the need to fit either a Fourier or a model-based basis. Naturally occurring external kinks are composed of multiple independent helical modes. Smooth transitions between dominant poloidal mode numbers are observed for simultaneous n = 1 and n = 2 modes as the edge safety factor changes. Relative amplitudes of coexistent m/n = 3/1 and 6/2 modes depend on the plasma's major radius and edge safety factor. When stationary 3/1 magnetic perturbations are applied, the resonant response can be linear, saturated, or disruptive, depending upon the perturbation amplitude and the edge safety factor; increased plasma-wall interactions from the perturbed plasma are proposed as a saturation mechanism. Initial feedback experiments have used 40 sensors and 40 control coils, producing mode amplification or suppression, and acceleration or deceleration depending on the feedback phase angle.
Gyrokinetic simulation of isotope scaling in tokamak plasmas
International Nuclear Information System (INIS)
A three-dimensional global gyrokinetic particle code in toroidal geometry has been used for investigating the transport properties of ion temperature gradient (ITG) drift instabilities in tokamak plasmas. Using the isotopes of hydrogen (H+), deuterium (D+) and tritium (T+), we have found that, under otherwise identical conditions, there exists a favorable isotope scaling for the ion thermal diffusivity, i.e., Xi decreases with mass. Such a scaling, which exists both at the saturation of the instability and also at the nonlinear steady state, can be understood from the resulting wavenumber and frequency spectra
Real-Time Software for the Compass Tokamak Plasma Control
Energy Technology Data Exchange (ETDEWEB)
Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)
2009-07-01
This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)
Solitary radial electric field structure in tokamak plasmas
International Nuclear Information System (INIS)
The solitary structure solution of the radial electric field Er in the tokamak plasmas is obtained. It is shown to be stable under the external power supply like a biased electrode. The radial gradient is governed by the ion viscosity and the nonlinearity of the perpendicular conductivity. The radial structure of Er and reduction of turbulent transport are self-consistently determined. A bifurcation from a radially-uniform one to solitary one occurs at a certain applied voltage, and a hysteresis is associated. (author)
Plasma rotation from momentum transport by neutrals in tokamaks
Omotani, John; Fülöp, Tünde
2016-01-01
Neutral atoms can strongly influence the intrinsic rotation and radial electric field at the tokamak edge. Here, we present a framework to investigate these effects when the neutrals dominate the momentum transport. We explore the parameter space numerically, using highly flexible model geometries and a state of the art kinetic solver. We find that the most important parameters controlling the toroidal rotation and electric field are the major radius where the neutrals are localized and the plasma collisionality. This offers a means to influence the rotation and electric field by, for example, varying the radial position of the X-point to change the major radius of the neutral peak.
Erosion of plasma-facing materials during a tokamak disruption
International Nuclear Information System (INIS)
The behaviour of divertor materials during a major disruption in a tokamak reactor is very important to successful and reliable operation of the device. Erosion of material surfaces due to a thermal energy dump can severely limit the lifetimes of plasma-facing components and thus diminish the reactor's economic feasibility. A comprehensive numerical model has been developed and used in this analysis, which includes all major physical processes taking place during plasma/material interactions. Models to account for material thermal evolution, plasma/vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud above the exposed surface are implemented in a self-consistent manner to realistically assess disruption damage. The extent of self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. Models to study detailed effects of the strong magnetic field on the behaviour of the vapor cloud and on the net erosion rate have also been developed and analyzed. Candidate materials such as beryllium and carbon are considered in this analysis. The dependence of divertor disruption lifetime on disruption physics and reactor conditions is analyzed and discussed. In addition, material erosion from melting of plasma-facing components during a tokamak disruption is also a serious problem that limits reactor operation and economical reactor lifetime. In particular, metallic low-Z components such as Be will be subjected to severe melting during disruptions and edge localized modes (ELMs). Loss of the developed melt layer will critically shorten the lifetime of these components, severely contaminate the plasma, and seriously inhibit successful and reliable operation of the reactor. In this study mechanisms responsible for melt-layer loss during a disruption are also modeled and evaluated. Implications of melt-layer loss on the performance of
Study of edge turbulence in tokamak plasmas
International Nuclear Information System (INIS)
The aim of this work is to propose a new frame to study turbulent transport in plasmas. In order to avoid the restraint of scale separability the forcing by flux is used. A critical one-dimension self-organized cellular model is developed. In keeping with experience the average transport can be described by means of diffusion and convection terms whereas the local transport could not. The instability due to interchanging process is thoroughly studied and some simplified equations are derived. The proposed model agrees with the following experimental results: the relative fluctuations of density are maximized on the edge, the profile shows an exponential behaviour and the amplitude of density fluctuations depends on ionization source strongly. (A.C.)
Equilibrium Reconstruction in EAST Tokamak
Institute of Scientific and Technical Information of China (English)
QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang
2009-01-01
Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier ex-pansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.
Shear Alfven waves in tokamaks
International Nuclear Information System (INIS)
Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma
THz time-domain spectroscopy for tokamak plasma diagnostics
Energy Technology Data Exchange (ETDEWEB)
Causa, F.; Zerbini, M.; Buratti, P.; Gabellieri, L.; Pacella, D.; Romano, A.; Tuccillo, A. A.; Tudisco, O. [ASSOCIAZIONE EURATOM ENEA sulla Fusione, C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Johnston, M. [Clarendon Laboratory, Department of Physics, University of Oxford, Parks Road, Oxford OX1 3PU (United Kingdom); Doria, A.; Gallerano, G. P.; Giovenale, E. [ENEA C.R. Frascati UTAPRAD, via E. Fermi 45, 00044 Frascati (Roma) (Italy)
2014-08-21
The technology is now becoming mature for diagnostics using large portions of the electromagnetic spectrum simultaneously, in the form of THz pulses. THz radiation-based techniques have become feasible for a variety of applications, e.g., spectroscopy, imaging for security, medicine and pharmaceutical industry. In particular, time-domain spectroscopy (TDS) is now being used also for plasma diagnostics in various fields of application. This technique is promising also for plasmas for fusion applications, where plasma characteristics are non-uniform and/or evolve during the discharge This is because THz pulses produced with femtosecond mode-locked lasers conveniently span the spectrum above and below the plasma frequency and, thus, can be used as very sensitive and versatile probes of widely varying plasma parameters. The short pulse duration permits time resolving plasma characteristics while the large frequency span permits a large dynamic range. The focus of this work is to present preliminary experimental and simulation results demonstrating that THz TDS can be realistically adapted as a versatile tokamak plasma diagnostic technique.
Particle simulation of plasma turbulence and neoclassical Er at tokamak plasma edge
International Nuclear Information System (INIS)
Particle simulations of turbulence, L-H transition and neoclassical electric field for various tokamaks both near the edge and inside the plasma are presented. Five dimensional Monte Carlo guiding centre orbit following code ASCOT, which simulates neoclassical physics, and its gyrokinetic upgrade ELMFIRE, which takes into account also electrostatic turbulence, are used. (author)
Current Sharing between Plasma and Walls in Tokamak Disruptions
International Nuclear Information System (INIS)
Full text: Plasma disruptions in tokamaks represent a significant obstacle in enhancing performance of the plasma regime, especially in the next step machines, such as ITER. Although, for the global forces due to disruptions on the vacuum vessel there is sufficient certainty because of explicit scalings, e.g., from JET to ITER, many important aspects of plasma interaction with the plasma facing components (localization of forces, their impulse, rotation, etc) require additional consideration. Here, the new aspects of electric current sharing between plasma and the wall during vertical disruption events (VDE) will be presented. Recently it was understood that theory predicted currents play the major role in VDEs. Called the 'Hiro' currents, they are excited in the wall by the plasma motion into the wall. Regarding them, the instability, which acts as a 'current' generator, provides large currents independent of resistivity of the plasma-wall contact. The Hiro currents can flow along the tiles surface while the plasma itself shorts out the electric circuit between tiles. The effect of the Hiro currents might be significant for the ITER plasma facing beryllium tiles. As a result, significant forces (both vertical and sideways) can be applied to the tiles themselves. Also, the edges of the tiles can be potentially damaged by significant Hiro currents flowing between tiles. Realistic numerical simulations of this effect with a presently being developed Disruption Simulation Code (DSC) will be presented. Also, the role of the counterpart of the Hiro currents (edge currents flowing in the same direction as the plasma current) during VDEs will be clarified by simulating VDE. The ESC code is appropriately modified for this purposes. These currents may suggest an alternative interpretation of the tile current measurements during VDE in contrast to the presently adopted 'halo' current concept. (author)
International Nuclear Information System (INIS)
A review of tracking charged particle motion in an axisymmetric toroidal plasma and of Monte Carlo modelling of particle-background interactions is given. Computational methods for efficient modelling of electron and ion guiding center orbits in tokamaks are described and the Monte Carlo orbit-following code ASCOT is reviewed. The efficiency of the code is based on the use of a coordinate system specifically designed for a toroidal system, on preventing numerical error accumulation, and on accelerating interaction time scales. Solutions for enhancing the computational efficiency of the Monte Carlo operators without deterioration of accuracy are described. Applications of the ASCOT code to studies of reverse runaway electrons, lower hybrid (LH) and ion cyclotron (IC) heating and current drive are presented. Relativistic effects are found to increase the reverse runaway probability of fast electrons during current ramp-up. Collisions, acting to diminish the total energy of the electrons towards thermal energy, have a lesser effect on the velocity of the test electron at relativistic energies. Combined to the effect of pitch collisions which bring the electrons towards the trapping cone, this relativistic effect enables the electrons to reach the trapping cone at a large total velocity, where the trapping cone is wide and the region traversed during trapped orbit motion is larger. This brings forth a notable increase in the reverse runaway probability. In a realistic tokamak configuration with smooth wave diffusion and fusion reactivity profiles, fusion-born alpha particles are found to interact with lower hybrid waves by absorbing energy from the wave. Special absorbing boundary conditions must be applied at the perpendicular energy boundary of the wave region in order to reverse the direction of energy transfer. A parameter study of ion cyclotron heating and current drive indicates that the power efficiency of minority ion current generation by IC waves is optimized
Continuum Kinetic Modeling of the Tokamak Plasma Edge
Dorf, Mikhail
2015-11-01
The problem of edge plasma transport provides substantial challenges for analytical or numerical analysis due to (a) complex magnetic geometry including both open and closed magnetic field lines B, (b) steep radial gradients comparable to ion drift-orbit excursions, and (c) a variation in the collision mean-free path along B from long to short compared to the magnetic connection length. Here, the first 4D continuum drift-kinetic transport simulations that span the magnetic separatrix of a tokamak are presented, motivated in part by the success of continuum kinetic codes for core physics and in part by the potential for high accuracy. The calculations include fully-nonlinear Fokker-Plank collisions and electrostatic potential variations. The problem of intrinsic toroidal rotation driven by ion orbit loss is addressed in detail. The code, COGENT, developed by the Edge Simulation Laboratory collaboration, is distinguished by a fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex magnetic X-point divertor geometry with high accuracy. Previously, successful performance of high-order algorithms has been demonstrated in a simpler closed magnetic-flux-surface geometry for the problems of neoclassical transport and collisionless relaxation of geodesic acoustic modes in a tokamak pedestal, including the effects of a strong radial electric field under H-mode conditions. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344.
Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks
International Nuclear Information System (INIS)
The contents of this report present the progress achieved to date on the Heavy Neutral Beam Probe project. This effort is an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadien de Fusion Magnetique (CCFM). The overall objective of the effort is to develop and apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes (TdeV) facility in Montreal, Canada. To achieve this goal, a research and development project was established to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present the project is in the middle of its second budget period with the instrumentation on-site at TdeV. The first half of this budget period was used to complete total system tests at InterScience, Inc., dismantle and ship the hardware to TdeV, re-assemble and install the HNBP on the tokamak. Integration of the diagnostic into the TdeV facility has progressed to the point of first beam production and measurement on the plasma. At this time, the HNBP system is undergoing final de-bugging prior to re-start of machine operation in early Fall of this year
Anomalous energy transport in hot plasmas: solar corona and Tokamak
International Nuclear Information System (INIS)
Anomalous energy transport is studied in two hot plasmas and appears to be associated with a heating of the solar corona and with a plasma deconfining process in tokamaks. The magnetic structure is shown to play a fundamental role in this phenomenon through small scale instabilities which are modelized by means of a nonlinear dynamical system: the Beasts' Model. Four behavior classes are found for this system, which are automatically classified in the parameter space thanks to a neural network. We use a compilation of experimental results relative to the solar corona to discuss current-based heating processes. We find that a simple Joule effect cannot provide the required heating rates, and therefore propose a dimensional model involving a resistive reconnective instability which leads to an efficient and discontinuous heating mechanism. Results are in good agreement with the observations. We give an analytical expression for a diffusion coefficient in tokamaks when magnetic turbulence is perturbing the topology, which we validate thanks to the standard mapping. A realistic version of the Beasts' Model allows to test a candidate to anomalous transport: the thermal filamentation instability
Study of neutral particle transport in Aditya Tokamak plasma using DEGAS2 Code
International Nuclear Information System (INIS)
Aditya tokamak is a medium sized air-core tokamak having a limiter configuration. The circular poloidal ring limiter is placed at one particular toroidal location. The spatial profile of neutral particles are experimentally observed in this tokamak and the observation suggests important roles of charge exchange processes into the penetration of neutral particle in plasma core. Therefore, to understand the neutral dynamics in Aditya tokamak, the neutral particle transport studies have been carried out using the DEGAS2 code. This code is based on Monte Carlo algorithms and extensively used for investigating the dynamics of neutrals in various tokamaks having divertors as the plasma facing component. The required modification has been carried out in the machine geometries and plasma parameter files through the user developed programs for ADITYA tokamak plasma. Modifications are successfully implemented in this code and the radial profile of Hα emissivity has been obtained. The simulated results are then compared with the experimental observations. In this paper, details on the implementation of the code on Aditya tokamak plasmas are presented and the simulation results are compared with the experiments to understand the neutral particle behaviour in Aditya tokamak plasma. (author)
Langmuir probe evaluation of the plasma potential in tokamak edge plasma for non-Maxwellian EEDF
Energy Technology Data Exchange (ETDEWEB)
Popov, Ts.K. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Ivanova, P. [Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Hasan, E. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Horacek, J.; Dejarnac, R.; Stoeckel, J.; Weinzettl, V. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)
2014-04-15
The First derivative probe technique for a correct evaluation of the plasma potential in the case of non-Maxwellian EEDF is presented and used to process experimental data from COMPASS tokamak. Results obtained from classical and first derivative techniques are compared and discussed. The first derivative probe technique provides values for the plasma potential in the scrape-off layer of tokamak plasmas with an accuracy of about ±10%. Classical probe technique can provide values of the plasma potential only, if the electron and ion temperatures are known as well as the coefficient of secondary electron emission. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
A Model Reference Adaptive Control for Radial Plasma Position on HL-2A Tokamak
Institute of Scientific and Technical Information of China (English)
MAOSuying; YUANBaoshan; LIQiang
2003-01-01
The radial plasma position control is one of the basic plasma controls of tokamak. In order to maintain a plasma column in the geometrical center of its vacuum vessel, the externally applied vertical field (VF) must be adjusted to the changes in the plasma parameters such as the plasma current, poloidal beta and the internal inductance.
Electron temperature gradient driven instability in the tokamak boundary plasma
Energy Technology Data Exchange (ETDEWEB)
Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.
1992-12-15
A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t{sup {minus}1/2}e{sup {gamma}mt}.
Low temperature plasma near a tokamak reactor limiter
Energy Technology Data Exchange (ETDEWEB)
Braams, B.J.; Singer, C.E.
1985-01-01
Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10/sup 13/cm/sup -3/ or less and the thermal diffusivity in the edge region is 2 x 10/sup 4/cm/sup 2//s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility.
High-current runaway electron beam in a tokamak plasma
International Nuclear Information System (INIS)
An equilibrium of toroidal plasma with a large electron-beam current has been realized using the runaway effect in a tokamak. Reproducible runaway-mode discharges are obtained with pure hydrogen gas by the help of intense titanium flashing which results in a low electron density. The beam current is estimated to be more than a half of the total toroidal current. The equilibrium of this discharge is maintained by a strong vertical field because the beam pressure gives rise to an additional increase in Shafranov Λ. The beam pressure is estimated to be more than 80% of the total pressure. The kinetic energy and the spatial distributions of beam electrons are studied by seeing X-ray emission from a tungsten wire inserted into the plasma. The increase of Shafranov Λ due to beam pressure is enhanced by puffing gas into the discharge. (author)
Effects of suprathermal fusion particles in tokamak plasmas
International Nuclear Information System (INIS)
Several crucial properties of suprathermal (> 500 keV) fusion-products are explored, both in their initial phase and during their slowing-down period. A guiding center drift theory, which predicts the effect of energy loss on the motion of these suprathermals, is derived for a low-β, symmetric (non-ripple) tokamak. Velocity-space scattering is ignored. Among the important implications of this theory are: (1) the net inward drift of fusion particles during their slow-down phase and (2) the importance of the plasma density and temperature in determining this drift. The effect the inward drifting has on the spatial profile for the suprathermals approaching thermal energies, on the energy distribution, and on the plasma heating profile is demonstrated for five reactor cases, ranging from near-term low-current devices to conceptual power reactors
Study of the plasma edge turbulence in tokamaks
International Nuclear Information System (INIS)
The plasma edge in tokamaks is known to be very turbulent. We investigate here the non linear stability of a test mode in presence of an helical potential perturbation, i.e. a pump mode, which simulates the plasma turbulence. The particle trajectories in this perturbed equilibrium are derived using an hamiltonian formalism. The electrons appear to have trapped trajectories in the potential well of the pump mode, while the ions experience a large convective motion. These two effects have a large influence on the test mode stability. First, non linearly trapped electrons supply an energy source for the test mode. Second, the ion convective motion introduces a radial scale of the test mode larger than the ion Larmor radius, in agreement with experimental data. These two phenomena allow a bifurcation in the turbulence level and provide therefore an explanation for the L-H transition
Effects of suprathermal fusion particles in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Petrie, T.W.
1978-01-01
Several crucial properties of suprathermal (> 500 keV) fusion-products are explored, both in their initial phase and during their slowing-down period. A guiding center drift theory, which predicts the effect of energy loss on the motion of these suprathermals, is derived for a low-..beta.., symmetric (non-ripple) tokamak. Velocity-space scattering is ignored. Among the important implications of this theory are: (1) the net inward drift of fusion particles during their slow-down phase and (2) the importance of the plasma density and temperature in determining this drift. The effect the inward drifting has on the spatial profile for the suprathermals approaching thermal energies, on the energy distribution, and on the plasma heating profile is demonstrated for five reactor cases, ranging from near-term low-current devices to conceptual power reactors.
Czech Academy of Sciences Publication Activity Database
Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan
96-97, October (2015), s. 974-979. ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044
International Nuclear Information System (INIS)
The motion of charged particles in a magnetized plasma column, such as that of a magnetic mirror trap or a tokamak, is determined in the framework of the canonical perturbation theory through a method of variation of constants which preserves the energy conservation and the symmetry invariance. The choice of a frame of coordinates close to that of the magnetic coordinates allows a relatively precise determination of the guiding-center motion with a low-ordered approximation in the adiabatic parameter. A Hamiltonian formulation of the motion equations is obtained
Insight of breaking of powerful axisymmetrically-polarized laser pulses in under-dense plasma
Nakanii, Nobuhiko; Pathak, Naveen C; Masuda, Shinichi; Zhidkov, Alexei G; Nakahara, Hiroki; Iwasa, Kenta; Mizuta, Yoshio; Takeguchi, Naoki; Otsuka, Takamitsu P; Sueda, Keiichi; Nakamura, Hirotaka; Mori, Michiaki; Kando, Masaki; Kodama, Ryosuke
2015-01-01
Interaction of axisymmetrically-polarized (radially or azimuthally-polarized), relativistically intense laser pulses (ALP) with under-dense plasma is shown experimentally to be different from the interaction of conventional Gaussian pulses. The difference is clearly observed in distinct spectra of scattered laser light as well as in appearance of a strong side emission of second harmonic in the vicinity of focus spot. According 3D particle-in-cell simulations, this is a result of instability in the propagation of ALP in under-dense plasma. Laser wakefield acceleration of electrons by ALP, therefore, is less efficient than that by Gaussian laser pulses but ALP may be interesting for efficient electron self-injection.
International Nuclear Information System (INIS)
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method
Energy Technology Data Exchange (ETDEWEB)
Kim, Dong-Hwan [Department of Nanoscale Semiconductor Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Hong, Suk-Ho [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); National Fusion Research Institute (NFRI), Daejeon 305-333 (Korea, Republic of); Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook, E-mail: joykang@hanyang.ac.kr [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of)
2015-12-15
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.
Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook
2015-12-01
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method. PMID:26724028
High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks
International Nuclear Information System (INIS)
High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)
High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks
Goodall, D. H. J.
1982-12-01
High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs.
High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Goodall, D.H.J. (Euratom/UKAEA Fusion Association, Abingdon (UK). Culham Lab.)
High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs.
Protection of tokamak plasma facing components by a capillary porous system with lithium
Energy Technology Data Exchange (ETDEWEB)
Lyublinski, I. [JSC «Red Star» Moscow (Russian Federation); National Research Nuclear University MEPhI, Moscow (Russian Federation); Vertkov, A. [JSC «Red Star» Moscow (Russian Federation); Mirnov, S.; Lazarev, V. [SSC TRINITI, Troitsk, Moscow (Russian Federation)
2015-08-15
Highlights: • CPS filled with liquid lithium applied as structural material of PFC in tokamak. • Behavior of lithium CPS at normal and disruption condition has been studied. • Concept of closed loop of lithium circulation in tokamak chamber has been confirmed experimentally. - Abstract: Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.
Protection of tokamak plasma facing components by a capillary porous system with lithium
International Nuclear Information System (INIS)
Highlights: • CPS filled with liquid lithium applied as structural material of PFC in tokamak. • Behavior of lithium CPS at normal and disruption condition has been studied. • Concept of closed loop of lithium circulation in tokamak chamber has been confirmed experimentally. - Abstract: Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements
Analysis of equilibrium and topology of tokamak plasmas
International Nuclear Information System (INIS)
In a tokamak, the plasma is confined by means of a magnetic field. There exists an equilibrium between outward forces due to the pressure gradient in plasma and inward forces due to the interaction between currents flowing inside the plasma and the magnetic field. The equilibrium magnetic field is characterized by helical field lines that lie on nested toroidal surfaces of constant flux. The equilibrium yields values for global and local plasma parameters (e.g. plasma position, total current, local pressure). Thus, precise knowledge of the equilibrium is essential for plasma control, for the understanding of many phenomena occurring in the plasma (in particular departures from the ideal equilibrium involving current filamentation on the flux surfaces that lead to the formation of islands, i.e. nested helical flux surfaces), and for the interpretation of many different types of measurements (e.g. the translation of line integrated electron density measurements made by laser beams probing the plasma into a local electron density on a flux surface). The problem of determining the equilibrium magnetic field from external magnetic field measurements has been studied extensively in literature. The problem is 'ill-posed', which means that the solution is unstable to small changes in the measurement data, and the solution has to be constrained in order to stabilize it. Various techniques for handling this problem have been suggested in literature. Usually ad-hoc restrictions are imposed on the equilibrium solution in order to stabilize it. More equilibrium solvers are not able to handle very dissimilar measurement data which means information on the equilibrium is lost. The generally do not allow a straightforward error estimate of the obtained results to be made, and they require large amounts of computing time. This problems are addressed in this thesis. (author). 104 refs.; 42 figs.; 6 tabs
International Nuclear Information System (INIS)
Flux-force relation, a fundamental relation that relates transport fluxes to forces, for non-axisymmetric tori in general magnetic flux coordinates that are not Hamada coordinates, is derived. The derivation is based on kinetic theory instead of fluid theory. It is shown that pressure force also contributes to the relation in non-Hamada coordinates in general to make the relation compatible with kinetic theory and to make it coordinates invariant. The results are applied to the theory for the neoclassical toroidal viscosity in tokamaks that have error fields or resistive magnetohydrodynamic (MHD) modes.
Helical temperature perturbations associated with tearing modes in tokamak plasmas
International Nuclear Information System (INIS)
An investigation is made into the electron temperature perturbations associated with tearing modes in tokamak plasmas, with a view to determining the mode structure using Electron Cyclotron Emission (ECE) data. It is found that there is a critical magnetic island width below which the conventional picture where the temperature is flattened inside the separatrix is invalid. This effect comes about because of the stagnation of magnetic field lines in the vicinity of the rational surface and the finite parallel thermal conductivity of the plasma. For islands whose widths lie below the critical value there is no flattening of the electron temperature inside the separatrix. Such islands have quite different ECE signatures to conventional magnetic islands. In fact the two island types could, in principle, be differentiated experimentally. It should also be possible to map out the outer ideal magnetohydrodynamical eigenfunctions using ECE data. Islands whose widths are much less than the critical value are not destabilized by the perturbed bootstrap current, unlike conventional magnetic islands. This effect is found to have a number of very interesting consequences and may, indeed, provide an explanation for some puzzling experimental results regarding error field induced magnetic reconnection. All islands whose widths are much greater than the critical width possess a boundary layer on the separatrix which enables heat to be transported from one side of the island to the other via the X-point region. The structure of this boundary layer is described in some detail. Finally, the critical island width is found to be fairly substantial in conventional tokamak plasmas, provided that the long mean free path nature of parallel heat transport and the anomalous nature of perpendicular heat transport are taken into account in the calculation
Mechanisms of plasma disruption and runaway electron losses in tokamaks
Abdullaev, S S; Wongrach, K; Tokar, M; Koslowski, H R; Willi, O; Zeng, L
2015-01-01
Based on the analysis of data from the numerous dedicated experiments on plasma disruptions in the TEXTOR tokamak mechanisms of the formation of runaway electron beams and their losses are proposed. The plasma disruption is caused by strong stochastic magnetic field formed due to nonlinearly excited low-mode number MHD modes. It is hypothesized that the runaway electron beam is formed in the central plasma region confined inside the intact magnetic surface located between $q=1$ and the closest low--order rational [$q=4/3$ or $q=3/2$] magnetic surfaces. The thermal quench time caused by the fast electron transport in a stochastic magnetic field is calculated using the collisional transport model. The current decay stage is due to the ambipolar particle transport in a stochastic magnetic field. The runaway electron beam in the confined plasma region is formed due to their acceleration the inductive toroidal electric field. The runaway electron beam current is modeled as a sum of toroidally symmetric part and a ...
Modelisation of synchrotron radiation losses in realistic tokamak plasmas
International Nuclear Information System (INIS)
Synchrotron radiation losses become significant in the power balance of high-temperature plasmas envisaged for next step tokamaks. Due to the complexity of the exact calculation, these losses are usually roughly estimated with expressions derived from a plasma description using simplifying assumptions on the geometry, radiation absorption, and density and temperature profiles. In the present article, the complete formulation of the transport of synchrotron radiation is performed for realistic conditions of toroidal plasma geometry with elongated cross-section, using an exact method for the calculation of the absorption coefficient, and for arbitrary shapes of density and temperature profiles. The effects of toroidicity and temperature profile on synchrotron radiation losses are analyzed in detail. In particular, when the electron temperature profile is almost flat in the plasma center, as for example in ITB confinement regimes, synchrotron losses are found to be much stronger than in the case where the profile is represented by its best generalized parabolic approximation, though both cases give approximately the same thermal energy contents. Such an effect is not included in present approximate expressions. Finally, we propose a seven-variable fit for the fast calculation of synchrotron radiation losses. This fit is derived from a large database, which has been generated using a code implementing the complete formulation and optimized for massively parallel computing. (author)
NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK
International Nuclear Information System (INIS)
OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability
Determination of the plasma position for its real-time control in the COMPASS tokamak
Energy Technology Data Exchange (ETDEWEB)
Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)
2011-10-15
An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.
Profile relaxation and transport in a Tokamak plasma
International Nuclear Information System (INIS)
Following the idea that physical processes occur in such a way as to produce relaxed states where an extremal of some relevant quantity is obtained, the tendency of temperature profiles to adopt gaussian-like shapes in tokamak plasmas is studied. The case in which these profiles are the result of a relaxation process that minimized the entropy production rate in steady state is firstly investigated. This procedure gives certain bounds for the temperature dependence of cross-field transport coefficients (heat conductivity and electrical resistivity). A second possibility explored focuses on the effect of profile variation on the total energy, for MHD equilibria obtained from the Grad-Shafranov equation. It is found that the energy tends to minimize values when the corresponding equilibrium temperature profiles approach the observed 'natural' profiles. (Author)
Impurity behavior in JIPP T-II tokamak plasma
International Nuclear Information System (INIS)
Resonance lines from oxygen and iron ions in JIPP T-II tokamak plasma are measured before and during the rise of the electron density. Spatial profiles of volume emissivities of OVII, FeXII and FeXVI are obtained. The comparison of experimental and computational results is made by using a one-dimensional impurity transport code. The code assumes oxygen influx by the desorption due to charge-exchanged hydrogen neutrals hitting the wall and iron influx by self-recycling. The increment of oxygen influx results in the increments of oxygen line intensities during the rise of electron density due to gas puffing, and the enhanced ionization process due to the increased electron density results in the outward movement of the peak position of volume emissivity of OVII. It is surmised that the diffusion of oxygen ions is dominated by anomalous diffusion while diffusion of iron ions is affected by both the neoclassical and anomalous diffusion. (author)
Theory of self-organized critical transport in tokamak plasmas
International Nuclear Information System (INIS)
A theoretical and computational study of the ion temperature gradient and ηi instabilities in tokamak plasmas has been carried out. In toroidal geometry the modes have a radially extended structure and their eigenfrequencies are constant over many rational surfaces that are coupled through toroidicity. These nonlocal properties of the ITG modes impose strong constraint on the drift mode fluctuations and the amciated transport, showing a self-organized characteristic. As any significant deviation away from marginal stability causes rapid temperature relaxation and intermittent bursts, the modes hover near marginality and exhibit strong kinetic characteristics. As a result, the temperature relaxation is self-semilar and nonlocal, leading to a radially increasing heat diffusivity. The nonlocal transport leads to the Bohm-like diffusion scaling. The heat input regulates the deviation of the temperature gradient away from marginality. The obtained transport scalings and properties are globally consistent with experimental observations of L-mode charges
Effects of turbulence on radiative properties of Tokamak edge plasmas
International Nuclear Information System (INIS)
The effect of turbulent temperature fluctuations on the radiative losses in tokamak edge and divertor plasmas is investigated. A statistical model is developed. The formalism involves both the probability density function of the fluid quantities and the expression of the emitting energy level population in terms of these quantities. We apply the statistical model to calculations of radiative power losses, successively for lithium and hydrogen radiation. In the former case, the energy level populations are obtained with an analytical collisional-radiative model accounting for non-coronal and transport effects. In the hydrogen case, the emitting level population is calculated by a collisional-radiative code. The role of electron temperature fluctuations is discussed in detail. Application to hydrogen line radiation in JET conditions reveals the significant role of turbulence in the repartition of the radiated energy inside the divertor (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Low-Zsub(eff) plasma confinement in TFR 600 tokamak
International Nuclear Information System (INIS)
TFR 600 is a tokamak which has mainly been devoted so far to one goal: the study of stable discharges at high plasma densities average value of n approximately >7x1013cm-3 for Isub(p)=200kA, Bsub(T)=40kG. In this case, Zsub(eff) is close to one, the ions contribute to the energy balance and seem to behave non-neoclassically. By polluting the vacuum chamber, it has been possible to reproduce TRF-400-type discharges, dominated by the electrons and impurity ions (Zsub(eff) approximately 5). The MHD activity investigated is characterized by a high-level m=3 tearing mode and by the evidence of a minor disruption during the overlap of m=3, m=2 modes. (author)
RF wave propagation and scattering in turbulent tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Horton, W., E-mail: wendell.horton@gmail.com; Michoski, C. [Institute for Fusion Studies, The University of Texas at Austin, Austin, TX 78654 (United States); Peysson, Y.; Decker, J. [CEA, IRFM, 13108, Saint-Paul, Durance Cedex (France)
2015-12-10
Drift wave turbulence driven by the steep electron and ion temperature gradients in H-mode divertor tokamaks produce scattering of the RF waves used for heating and current drive. The X-ray emission spectra produced by the fast electrons require the turbulence broaden RF wave spectrum. Both the 5 GHz Lower Hybrid waves and the 170 GHz electron cyclotron [EC] RF waves experience scattering and diffraction by the electron density fluctuations. With strong LHCD there are bifurcations in the coupled turbulent transport dynamics giving improved steady-state confinement states. The stochastic scattering of the RF rays makes the prediction of the distribution of the rays and the associated particle heating a statistical problem. Thus, we introduce a Fokker-Planck equation for the probably density of the RF rays. The general frame work of the coupled system of coupled high frequency current driving rays with the low-frequency turbulent transport determines the profiles of the plasma density and temperatures.
RF wave propagation and scattering in turbulent tokamak plasmas
Horton, W.; Michoski, C.; Peysson, Y.; Decker, J.
2015-12-01
Drift wave turbulence driven by the steep electron and ion temperature gradients in H-mode divertor tokamaks produce scattering of the RF waves used for heating and current drive. The X-ray emission spectra produced by the fast electrons require the turbulence broaden RF wave spectrum. Both the 5 GHz Lower Hybrid waves and the 170 GHz electron cyclotron [EC] RF waves experience scattering and diffraction by the electron density fluctuations. With strong LHCD there are bifurcations in the coupled turbulent transport dynamics giving improved steady-state confinement states. The stochastic scattering of the RF rays makes the prediction of the distribution of the rays and the associated particle heating a statistical problem. Thus, we introduce a Fokker-Planck equation for the probably density of the RF rays. The general frame work of the coupled system of coupled high frequency current driving rays with the low-frequency turbulent transport determines the profiles of the plasma density and temperatures.
Effect of density changes on tokamak plasma confinement
Spineanu, F
2015-01-01
A change of the particle density (by gas puff, pellets or impurity seeding) during the plasma discharge in tokamak produces a radial current and implicitly a torque and rotation that can modify the state of confinement. After ionization the newly born ions will evolve toward the periodic neoclassical orbits (trapped or circulating) but the first part of their excursion, which precedes the periodicity, is an effective radial current. It is short, spatially finite and unique for each new ion, but multiplied by the rate of ionization and it can produce a substantial total radial current. The associated torque induces rotation which modify the transport processes. We derive the magnitude of the radial current induced by ionization by three methods: the analysis of a simple physical picture, a numerical model and the neoclassical drift-kinetic treatment. The results of the three approaches are in agreement and show that the current can indeed be substantial. Many well known experimental observations can be reconsi...
Turbulent transport of alpha particles in tokamak plasmas
Croitoru, A; Vlad, M; Spineanu, F
2016-01-01
We investigate the ExB diffusion of fusion born \\alpha particles in tokamak plasmas. We determine the transport regimes for a realistic model that has the characteristics of the ion temperature gradient (ITG) or of the trapped electron modes (TEM) driven turbulence. It includes a spectrum of potential fluctuations that is modeled using the results of the numerical simulations, the drift of the potential with the effective diamagnetic velocity and the parallel motion. Our semi-analytical statistical approach is based on the decorrelation trajectory method (DTM), which is adapted to the gyrokinetic approximation. We obtain the transport coefficients as a function of the parameters of the turbulence and of the energy of the \\alpha particle. According to our results, signficant turbulent transport of the \\alpha particles can appear only at energies of the order of 100KeV. We determine the corresponding conditions.
An Overview of Plasma Confinement in Toroidal Systems
Dini, Fatemeh; Baghdadi, Reza; Amrollahi, Reza; Khorasani, Sina
2009-01-01
This overview presents a tutorial introduction to the theory of magnetic plasma confinement in toroidal confinement systems with particular emphasis on axisymmetric equilibrium geometries, and tokamaks. The discussion covers three important aspects of plasma physics: Equilibrium, Stability, and Transport. The section on equilibrium will go through an introduction to ideal magnetohydrodynamics, curvilinear system of coordinates, flux coordinates, extensions to axisymmetric equilibrium, Grad-Sh...
Modeling of noble gas injection into tokamak plasmas
International Nuclear Information System (INIS)
Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)
Power supplies for plasma column control in the COMPASS tokamak
Energy Technology Data Exchange (ETDEWEB)
Havlicek, J. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Hauptmann, R. [ČKD Elektrotechnika, Kolbenova 936/5e, 190 00 Praha 9 (Czech Republic); Peroutka, O.; Tadros, M. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Hron, M., E-mail: hron@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Janky, F.; Vondracek, P. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Cahyna, P. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Mikulín, O. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Physical Electronics, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Šesták, D.; Junek, P.; Pánek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)
2013-10-15
Highlights: ► Design of power supplies for fast control of plasma position in COMPASS tokamak. ► Design of power supply for experiments with ELM control by vertical plasma movement. ► Common regulator for power supplies for vertical plasma position and for ELMs control. ► Current status of construction, commissioning, and operation of these power supplies. -- Abstract: The main magnetic fields in COMPASS – i.e. The Toroidal, Magnetising, Equilibrium, and Shaping Fields – are created by a set of four corresponding thyristor power supplies controlled in a 0.5 ms loop. The plasma position has to be controlled both radially and vertically by two additional magnetic fields provided by two fast amplifiers (FAs) based on MOSFET technology, each supplying ±100 V and up to ±5 kA. Currently, an ongoing project aims at ELM triggering by fast changes of the vertical position of the plasma column, also referred to as vertical kicks. For this purpose, a new Vertical Kicks Power Supply (VKPS) capable of quick change of vertical plasma position is being constructed. This power supply should operate at up to 1.2 kV with switching frequency up to 5 kHz. It is designed as a H-bridge but based on IGBT transistors which can be operated at higher voltages than MOSFETs. In this paper, we focus on the FAs and VKPS engineering design and required output parameters. Both the power supplies are based on modern components with highest available ratings in their categories. Unique design of the power supplies takes advantage of the short duration of the COMPASS discharge by overloading the transistors above their maximal steady-state rating. The FA is regularly operating, so that in addition to describing its design, we also describe the achieved performance parameters. Finally, the common controller unit, communication, and error handling is described.
Theory of self-organized critical transport in tokamak plasmas
International Nuclear Information System (INIS)
A theoretical and computational study of the ion temperature gradient (ITG) and ηi instabilities in tokamak plasmas has been carried out. In a toroidal geometry the modes have a radially extended structure and their eigenfrequencies are constant over many rational surfaces that are coupled through toroidicity. These nonlocal properties of the ITG modes impose a strong constraint on the drift mode fluctuations and the associated transport, showing self-organized criticality. As any significant deviation away from marginal stability causes rapid temperature relaxation and intermittent bursts, the modes hover near marginality and exhibit strong kinetic characteristics. As a result of this, the temperature relaxation is self-similar and nonlocal, leading to radially increasing heat diffusivity. The nonlocal transport leads to Bohm-like diffusion scaling. Heat input regulates the deviation of the temperature gradient away from marginality. We present a critical gradient transport model that describes such a self-organized relaxed state. Some of the important aspects in tokamak transport like Bohm diffusion, near marginal stability, radially increasing fluctuation energy and heat diffusivity, intermittency of the wave excitation, and resilient tendency of the plasma profile can be described by this model, and these prominent features are found to belong to one physical category that originates from the radially extended nonlocal drift modes. The obtained transport properties and scalings are globally consistent with experimental observations of low confinement mode (L-mode) discharges. The nonlocal modes can be disintegrated into smaller radial islands by a poloidal shear flow, suggesting that the transport changes from Bohm-like to near gyro-Bohm. copyright 1996 American Institute of Physics
ICRF-driven convective cells in the tokamak edge plasma
International Nuclear Information System (INIS)
Recently, the authors have shown that the release of metal impurities during ICRF heating on JET could be explained by ion acceleration into the Faraday screen (FS) surface caused by rf sheaths, which form when local magnetic field is imperfectly aligned with the FS. The theory explained many of the dependences of the impurity data, including the virtual elimination of impurities with beryllium screens and dipole antenna phasing. The good agreement between the impurity generation model and experimental data can be taken as evidence of the relevance of rf sheaths to the scrape-off-layer (SOL) plasma in tokamaks. A crucial physical point in sheath theory is that the requirement of no time-averaged current into the boundary leads to the rectification of the oscillating rf fields to produce a net time-averaged and spatially-varying potential and, hence, a DC electric field. Here, the authors investigate the possibility that the rectified sheath potential Φo can drive convective cells in the SOL which may explain the experimental observations of ICRF-enhanced edge transport on many tokamaks. Temperature (and sometimes density) profile flattening and induced DC electric fields in the SOL are often observed during ICRF heating on JET, particularly in monopole phasing. The attainment of the H-mode with ICRF heating alone is also sensitive to the phasing of the antenna. These observations suggest an rf-sheath related effect, as the magnitude of Φo is much larger in monopole phasing (∼1 kV near the FS). The authors speculate that enhanced edge cooling by rapid convection may account for the phasing sensitivity of the H-mode transition. In the present work, a modified convective cell equation for the SOL plasma is derived, which explicitly takes into account the finite length of the field lines and the appropriate sheath boundary condition for Jparallel
Neoclassical and anomalous transport in axisymmetric toroidal plasmas with electrostatic turbulence
International Nuclear Information System (INIS)
Neoclassical and anomalous transport fluxes are determined for axisymmetric toroidal plasmas with weak electrostatic fluctuations. The neoclassical and anomalous fluxes are defined based on the ensemble-averaged kinetic equation with the statistically averaged nonlinear term. The anomalous forces derived from that quasilinear term induce the anomalous particle and heat fluxes. The neoclassical banana-plateau particle and heat fluxes and the bootstrap current are also affected by the fluctuations through the parallel anomalous forces and the modified parallel viscosities. The quasilinear term, the anomalous forces, and the anomalous particle and heat fluxes are evaluated from the fluctuating part of the drift kinetic equation. The averaged drift kinetic equation with the quasilinear term is solved for the plateau regime to derive the parallel viscosities modified by the fluctuations. The entropy production rate due to the anomalous transport processes is formulated and used to identify conjugate pairs of the anomalous fluxes and forces, which are connected by the matrix with the Onsager symmetry. (author)
Sawtooth Activity in Ohmically Heated Plasma on HT-7 Tokamak
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
Sawtooth activity on HT-7 tokamak has been investigated experimentally mainly by using soft x-ray diode array and magnetic probes. Their behaviors and occurrences are correlatedclosely to the discharge conditions: the electron density Ne, the electron temperature Te, the safetyfactor qa on plasma boundary and wall condition etc. When central line-averaged electron densityNe(0) is over 2.0×1013cm-3, major sawtooth activity emerges with a period of up to 6.5 ms and afluctuation amplitude of up to 2～30 % of SXR radiation signal. In some cases such as the safetyfactor between 4.2～4.7 and Zeff=3.0～6.0, a monster sawtooth activity often emerges withoutapparent deterioration of plasma confinement and without major disruption. During these events,abundant MHD phenomena are observed including partial sawtooth oscillations. In this paper, theobserved sawtooth behaviors and their dependence on the and their dependence density Ne andwall condition in ohmically heated plasma are introduced, the results are discussed and presented.
Reactor aspects of counterstreaming-ion tokamak plasmas
International Nuclear Information System (INIS)
Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)
Plasma pressure in the discharge column of the Novillo Tokamak
International Nuclear Information System (INIS)
The design and construction of an acquisition system for the measurement of the plasma pressure in the Novillo Tokamak is described in detail. The system includes a high voltage ramp generator, a hardware and a software interface with a personal computer. It is used to determine experimentally the variations of the pressure in the plasma column in the cleaning and main discharges. The measurement of the pressure is made with a Pirani sensor adapted to the acquisition hardware and synchronized with the discharge in the plasma. The software is made in object oriented programming as a graphic interface designed to be used easily. It controls the acquisition, records the data, displays in graphic form the results and save the measurements. The graphic interface is a building block that can be used in different acquisition tasks. The ramp generator can deliver a signal of 200 V peak to peak with a current of 200 m A and offset control. The acquisition time is 2.5 μ s for every measurement, 8192 measurements can be stored in the acquisition board for every discharge. (Author)
Tokamak operation with high-Z plasma facing components
International Nuclear Information System (INIS)
Due to wall lifetime requirements and the problem of tritium co-deposition in hydrocarbon layers, a future burning plasma will most probably have a full high-Z wall. The prime candidate material is tungsten, which exhibits good thermo-mechanical properties and has a high energy threshold for physical sputtering. To investigate the reactor-compatibility of this wall material, ASDEX Upgrade is being converted into a full-tungsten coated tokamak in a step-by-step approach, with presently almost 70 % of W wall coverage. The effect of the reduction of primary carbon coverage on the plasma is so far moderate. Under tokamak conditions, carbon behaves like a recycling impurity, due to the deposition and re-erosion of soft hydrocarbon layers on the tungsten surface. During high density H-mode operation, the central tungsten concentrations remain typically low, i.e. well under 10-5. The situation is more critical in the improved H-mode or hybrid scenario. Here, the combination of hot edge conditions and peaked central density profiles result in high central tungsten concentrations of up to 10-4, which would be critical in a reactor. However, core electron density peaking is reduced by use of central ICR or ECR heating and thus in turn suppresses central tungsten accumulation. For extrapolation to reactor conditions, we need to separate the effects of the tungsten wall source, the penetration over the edge transport barrier (ETB) and the core transport with its strong neoclassical contribution. These issues are addressed by inspecting the tungsten behaviour in various discharge scenarios and parameters in ASDEX Upgrade. These include radiative cooling by medium-Z seed impurities and ELM frequency control by pellet injection to simulate a reactor plasma with small edge and divertor impurity radiation levels and a separatrix power flux close to the H-L threshold. Fast ions produced by NBI and ICR heating at the low field side appear to be an important tungsten erosion mechanism
Experimental device for the X-ray energetic distribution measurement in a tokamak plasma
International Nuclear Information System (INIS)
An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author)
ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM
Energy Technology Data Exchange (ETDEWEB)
HUMPHREYS,DA; FERRON,JR; GAROFALO,AM; HYATT,AW; JERNIGAN,TC; JOHNSON,RD; LAHAYE,RJ; LEUER,JA; OKABAYASHI,M; PENAFLOR,BG; SCOVILLE,JT; STRAIT,EJ; WALKER,ML; WHYTE,DG
2002-10-01
A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.
Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks
Czech Academy of Sciences Publication Activity Database
Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel
2008-01-01
Roč. 15, č. 8 (2008), 082505-1-082505-8. ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146
Dispersion equations for field-aligned cyclotron waves in axisymmetric magnetospheric plasmas
Directory of Open Access Journals (Sweden)
N. I. Grishanov
2006-03-01
Full Text Available In this paper, we derive the dispersion equations for field-aligned cyclotron waves in two-dimensional (2-D magnetospheric plasmas with anisotropic temperature. Two magnetic field configurations are considered with dipole and circular magnetic field lines. The main contribution of the trapped particles to the transverse dielectric permittivity is estimated by solving the linearized Vlasov equation for their perturbed distribution functions, accounting for the cyclotron and bounce resonances, neglecting the drift effects, and assuming the weak connection of the left-hand and right-hand polarized waves. Both the bi-Maxwellian and bi-Lorentzian distribution functions are considered to model the ring current ions and electrons in the dipole magnetosphere. A numerical code has been developed to analyze the dispersion characteristics of electromagnetic ion-cyclotron waves in an electron-proton magnetospheric plasma with circular magnetic field lines, assuming that the steady-state distribution function of the energetic protons is bi-Maxwellian. As in the uniform magnetic field case, the growth rate of the proton-cyclotron instability (PCI in the 2-D magnetospheric plasmas is defined by the contribution of the energetic ions/protons to the imaginary part of the transverse permittivity elements. We demonstrate that the PCI growth rate in the 2-D axisymmetric plasmasphere can be significantly smaller than that for the straight magnetic field case with the same macroscopic bulk parameters.
Real-Time Control of Tokamak Plasmas: from Control of Physics to Physics-Based Control
Felici, Federico
2011-01-01
Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solut...
International Nuclear Information System (INIS)
Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission
High density plasma heating in the Tokamak à configuration variable
International Nuclear Information System (INIS)
The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2nd harmonic EC power in ordinary (O2) or extraordinary (X2) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3rd harmonic EC power (X3) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X2 and X3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation of the AMR
3D simulation studies of tokamak plasmas using MHD and extended-MHD models
International Nuclear Information System (INIS)
The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω*i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data
Theory of "clumps" in drift-wave turbulence in tokamak plasma
Wang, Xiaogang; Qiu, Xiaoming; X, M. Qhiu
1986-08-01
Basing on the new method of trajectory stochastic treatment advanced by one of the authors of this paper, the theory of "clumps" in driftwave turbulence in tokamak plasmas has been developed. It is shown that, as a longer time behaviour, plasmas in tokamaks will have the same "clumps" effects as those in uniform magnetic fields, although the diffusion crossing magnetic field lines in tokamaks will be enhanced. The influence of the non-uniformity of the magnetic field, such as curvature, shear, etc., on the transverse diffusion and the "clump" life-time is discussed.
Reedy, Todd Mitchell
An experimental investigation evaluating the effects of flow control on the near-wake downstream of a blunt-based axisymmetric body in supersonic flow has been conducted. To better understand and control the physical phenomena that govern these massively separated high-speed flows, this research examined both passive and active flow-control methodologies designed to alter the stability characteristics and structure of the near-wake. The passive control investigation consisted of inserting splitter plates into the recirculation region. The active control technique utilized energy deposition from multiple electric-arc plasma discharges placed around the base. The flow-control authority of both methodologies was evaluated with experimental diagnostics including particle image velocimetry, schlieren photography, surface flow visualization, pressure-sensitive paint, and discrete surface pressure measurements. Using a blowdown-type wind tunnel reconstructed specifically for these studies, baseline axisymmetric experiments without control were conducted for a nominal approach Mach number of 2.5. In addition to traditional base pressure measurements, mean velocity and turbulence quantities were acquired using two-component, planar particle image velocimetry. As a result, substantial insight was gained regarding the time-averaged and instantaneous near-wake flow fields. This dataset will supplement the previous benchmark point-wise laser Doppler velocimetry data of Herrin and Dutton (1994) for comparison with new computational predictive techniques. Next, experiments were conducted to study the effects of passive triangular splitter plates placed in the recirculation region behind a blunt-based axisymmetric body. By dividing the near-wake into 1/2, 1/3, and 1/4 cylindrical regions, the time-averaged base pressure distribution, time-series pressure fluctuations, and presumably the stability characteristics were altered. While the spatial base pressure distribution was
Anomalous electron streaming due to electrostatic modes in tokamak plasmas
International Nuclear Information System (INIS)
The motion of circulating electrons in a tokamak interacting with electrostatic waves (such as lower-hybrid waves) is given by a guiding center Hamiltonian and studied by numerical integration. The unperturbed motion of electron guiding centers is first shown to be integrable, and, in a manner similar to that used in previous works, a set of action-angle coordinates for the orbits are derived which take into account finite aspect ratio and noncircular plasma cross section. Electrostatic modes in the low-frequency, long-wavelength limit are treated as a perturbation to the guiding center Hamiltonian. The waves are generated with low integral values of the toroidal and poloidal mode numbers n and m and satisfy the approximate lower-hybrid dispersion relation kperpendicular/kparallel ∼ ωpe/ω ∼ 101.5. If the number of modes is greater than three, the electron motion parallel to the magnetic field is observed to be stochastic in the phase-space region where vparallel is near the wave parallel phase velocity. On surfaces with rational values of the safety factor q, superposition of modes with degenerate values of the parallel mode number n + (m/q) is shown to result in electron streaming perpendicular to the magnetic field. The speed and direction of this radial motion are observed to have sinusoidal dependence on the poloidal angle. For models including finite magnetic-field shear, the authors find a limit to the extent of the radial streaming of the electrons. Results for the speed of the electron radial motion for typical tokamak parameters are presented
Kinetic modelling of runaway electron avalanches in tokamak plasmas
Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R. S.; Saint-Laurent, F.; Vlainic, M.
2015-09-01
Runaway electrons can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force owing to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate runaway electrons mainly through knock-on collisions (Hender et al 2007 Nucl. Fusion 47 S128-202), where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of runaway electrons. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. For this purpose, a bounce-averaged knock-on source term is derived. The generation of runaway electrons from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a solver of the 3D linearized bounce-averaged relativistic electron Fokker-Planck equation (Decker and Peysson 2004 DKE: a fast numerical solver for the 3D drift kinetic equation Report EUR-CEA-FC-1736, Euratom-CEA), through the calculation of the response of the electron distribution function to a constant parallel electric field. The model, which has been successfully benchmarked against the standard Dreicer runaway theory now describes the runaway generation by knock-on collisions as proposed by Rosenbluth (Rosenbluth and Putvinski 1997 Nucl. Fusion 37 1355-62). This paper shows that the avalanche effect can be important even in non-disruptive scenarios. Runaway formation through knock-on collisions is found to be strongly reduced when taking place off the magnetic axis, since trapped electrons can not contribute to the runaway electron population. Finally, the
High-Q plasmas in the TFTR tokamak
International Nuclear Information System (INIS)
In the Tokamak Fusion Test Reactor, the highest neutron source strength Sn and D-D fusion power gain QDD are realized in the neutral-beam fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as Pb1.8. The highest-Q shots are characterized by high Te, Ti, and stored energy highly peaked density profiles, broad Te profiles, and lower Zeff. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles, and improved alignment with the plasma, have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, QDD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness during the beam pulse. To date the best fusion results are Sn = 5 x 1016 n/s, QDD = 1.85 x 10-3, and neutron yield = 4.0 x 1016 n/pulse, obtained at Ip = 1.6 to 1.9 MA and beam energy Eb = 95 to 103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50--60% of Sn arises from beam-target reactions, with the remainder divided between beam-beam and thermonuclear reactions, the thermonuclear fraction increasing with Pb. The simulations predict that QDT = 0.3 to 0.4 would be obtained for the best present plasma conditions, if half the deuterium neutral beams were to be replaced by tritium beams. Somewhat higher values are calculated if D beams are injected into a predominantly tritium target plasma. The projected central beta of fusion alphas is 0.4--0.6%, a level sufficient for the study of alpha-induced collective effects. 16 refs., 8 figs., 3 tabs
Modeling of the equilibrium of a tokamak plasma
International Nuclear Information System (INIS)
The simulation and the control of a plasma discharge in a tokamak require an efficient and accurate solving of the equilibrium because this equilibrium needs to be calculated again every microsecond to simulate discharges that can last up to 1000 seconds. The purpose of this thesis is to propose numerical methods in order to calculate these equilibrium with acceptable computer time and memory size. Chapter 1 deals with hydrodynamics equation and sets up the problem. Chapter 2 gives a method to take into account the boundary conditions. Chapter 3 is dedicated to the optimization of the inversion of the system matrix. This matrix being quasi-symmetric, the Woodbury method combined with Cholesky method has been used. This direct method has been compared with 2 iterative methods: GMRES (generalized minimal residual) and BCG (bi-conjugate gradient). The 2 last chapters study the control of the plasma equilibrium, this work is presented in the formalism of the optimized control of distributed systems and leads to non-linear equations of state and quadratic functionals that are solved numerically by a quadratic sequential method. This method is based on the replacement of the initial problem with a series of control problems involving linear equations of state. (A.C.)
Plasma control issues for an advanced steady state tokamak reactor
International Nuclear Information System (INIS)
This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)
Experimental observations of driven and intrinsic rotation in tokamak plasmas
Rice, J. E.
2016-08-01
Experimental observations of driven and intrinsic rotation in tokamak plasmas are reviewed. For momentum sources, there is direct drive from neutral beam injection, lower hybrid and ion cyclotron range of frequencies waves (including mode conversion flow drive), as well as indirect \\mathbf{j}× \\mathbf{B} forces from fast ion and electron orbit shifts, and toroidal magnetic field ripple loss. Counteracting rotation drive are sinks, such as from neutral drag and toroidal viscosity. Many of these observations are in agreement with the predictions of neo-classical theory while others are not, and some cases of intrinsic rotation remain puzzling. In contrast to particle and heat fluxes which depend on the relevant diffusivity and convection, there is an additional term in the momentum flux, the residual stress, which can act as the momentum source for intrinsic rotation. This term is independent of the velocity or its gradient, and its divergence constitutes an intrinsic torque. The residual stress, which ultimately responds to the underlying turbulence, depends on the confinement regime and is a complicated function of collisionality, plasma shape, and profiles of density, temperature, pressure and current density. This leads to the rich intrinsic rotation phenomenology. Future areas of study include integration of these many effects, advancement of quantitative explanations for intrinsic rotation and development of strategies for velocity profile control.
Kinetic modelling of runaway electron avalanches in tokamak plasmas
Nilsson, E; Peysson, Y; Granetz, R S; Saint-Laurent, F; Vlainic, M
2015-01-01
Runaway electrons (REs) can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force due to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate REs mainly through knock-on collisions, where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of REs. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3-D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. A bounce-averaged knock-on source term is derived. The generation of REs from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a s...
Czech Academy of Sciences Publication Activity Database
Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.
2006-01-01
Roč. 12, č. 6 (2006), s. 14-19. ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http://vant.kipt.kharkov.ua/TABFRAME.html
International Nuclear Information System (INIS)
In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity χ to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.)
Motions of dust particles in a complex plasma with an axisymmetric nonuniform magnetic field
Saitou, Yoshifumi
2016-01-01
We investigate the motions of dust particles in a complex plasma by applying an axisymmetric nonuniform magnetic field, B , introduced with a permanent magnet. The magnetic field changes its direction from upward to downward within the experimental area. The distribution of dust particles is conical in the meridional plane, and its central area is a void. The dust particles are generally stagnant in the vertical direction and distributed in multiple layers. The horizontal plane is separated into two regions where the vertical component of B can and cannot be regarded as zero. The distribution of the dust particles in the horizontal plane is concentric. The dust particles along the inner and outer edges rotate in opposite directions due to the direction of the vertical component of B and generate shear flow at a certain height. The rotation velocities of the particles at the edges are compared with the theory of Kaw et al. [Phys. Plasmas 9, 387 (2002)]. The vortex-like structure is not easy to observe even in the presence of a shear flow because of the influence of the other dust particles as well as the small Reynolds number of the dust fluid.
The direct criterion of Newcomb for the ideal MHD stability of an axisymmetric toroidal plasma
Glasser, A. H.
2016-07-01
A method is presented for determining the ideal magnetohydrodynamic stability of an axisymmetric toroidal plasma, based on a toroidal generalization of the method developed by Newcomb for fixed-boundary modes in a cylindrical plasma. For toroidal mode number n ≠ 0 , the stability problem is reduced to the numerical integration of a high-order complex system of ordinary differential equations, the Euler-Lagrange equation for extremizing the potential energy, for the coupled amplitudes of poloidal harmonics m as a function of the radial coordinate ψ in a straight-fieldline flux coordinate system. Unlike the cylindrical case, different poloidal harmonics couple to each other, which introduces coupling between adjacent singular intervals. A boundary condition is used at each singular surface, where m = nq and q ( ψ ) is the safety factor, to cross the singular surface and continue the solutions beyond it. Fixed-boundary instability is indicated by the vanishing of a real determinant of a Hermitian complex matrix constructed from the fundamental matrix of solutions, the generalization of Newcomb's crossing criterion. In the absence of fixed-boundary instabilities, an M × M plasma response matrix W P , with M the number of poloidal harmonics used, is constructed from the Euler-Lagrange solutions at the plasma-vacuum boundary. This is added to a vacuum response matrix W V to form a total response matrix W T . The existence of negative eigenvalues of W T indicates the presence of free-boundary instabilities. The method is implemented in the fast and accurate DCON code.
Improvement of plasma performance with wall conditioning in the HL-1M tokamak
International Nuclear Information System (INIS)
Studies and selection of plasma-facing materials continue to be a concern for future fusion devices, and ongoing effort are being made in the HL-1M Tokamak. Significant improvements in Tokamak plasma performance have been obtained by using boron, silicon or lithium-containing substance as a material for wall coatings. The new type of wall conditioning is plasma chemical vapor deposition. The author gives an overview of experimental methods and results in HL-1M. The modification of the wall and improvement of plasma performance are summarized
Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach
International Nuclear Information System (INIS)
Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- βplasma (where the parameter β identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing β, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite β are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the AITG mode is analysed
Electromagnetic microinstabilities in tokamak plasmas using a global spectral approach
Energy Technology Data Exchange (ETDEWEB)
Falchetto, G. L
2002-03-01
Electromagnetic microinstabilities in tokamak plasmas are studied by means of a linear global eigenvalue numerical code. The code is the electromagnetic extension of an existing electrostatic global gyrokinetic spectral toroidal code, called GLOGYSTO. Ion dynamics is described by the gyrokinetic equation, so that ion finite Larmor radius effects are taken into account to all orders. Non adiabatic electrons are included in the model, with passing particles described by the drift-kinetic equation and trapped particles through the bounce averaged drift-kinetic equation. A low frequency electromagnetic perturbation is applied to a low -but finite- {beta}plasma (where the parameter {beta} identifies the ratio of plasma pressure to magnetic pressure); thus, the parallel perturbations of the magnetic field are neglected. The system is closed by the quasi-neutrality equation and the parallel component of Ampere's law. The formulation is applied to a large aspect ratio toroidal configuration, with circular shifted surfaces. Such a simple configuration enables one to derive analytically the gyrocenter trajectories. The system is solved in Fourier space, taking advantage of a decomposition adapted to the toroidal geometry. The major contributions of this thesis are as follows. The electromagnetic effects on toroidal Ion Temperature Gradient driven (ITG) modes are studied. The stabilization of these modes with increasing {beta}, as predicted in previous work, is confirmed. The inclusion of trapped electron dynamics enables the study of its coupling to the ITG modes and of Trapped Electron Modes (TEM) .The effects of finite {beta} are considered together with those of different magnetic shear profiles and of the Shafranov shift. The threshold for the destabilization of an electromagnetic mode is identified. Moreover, the global formulation yields for the first time the radial structure of this so-called Alfvenic Ion Temperature Gradient (AITG) mode. The stability of the
High-Q plasmas in the TFTR tokamak
International Nuclear Information System (INIS)
In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength Sn and D--D fusion power gain QDD are realized in the neutral-beam-fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as P1.8b. The highest-Q shots are characterized by high Te (up to 12 keV), Ti (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad Te profiles, and lower Zeff. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, QDD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [ne(0)/left-angle ne right-angle] during the beam pulse. To date, the best fusion results are Sn=5x1016 n/sec, QDD=1.85x10-3, and neutron yield=4.0x1016 n/pulse, obtained at Ip=1.6--1.9 MA and beam energy Eb=95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of Sn arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with Pb
Optimization of tokamak plasma equilibrium shape using parallel genetic algorithms
International Nuclear Information System (INIS)
In the device of non-circular cross sectional tokamaks, the plasma equilibrium shape has a strong influence on the confinement and MHD stability. The plasma equilibrium shape is determined by the configuration of the poloidal field (PF) system. Usually there are many PF systems that could support the specified plasma equilibrium, the differences are the number of coils used, their positions, sizes and currents. It is necessary to find the optimal choice that meets the engineering constrains, which is often done by a constrained optimization. The Genetic Algorithms (GAs) based method has been used to solve the problem of the optimization, but the time complexity limits the algorithms to become widely used. Due to the large search space that the optimization has, it takes several hours to get a nice result. The inherent parallelism in GAs can be exploited to enhance their search efficiency. In this paper, we introduce a parallel genetic algorithms (PGAs) based approach which can reduce the computational time. The algorithm has a master-slave structure, the slave explore the search space separately and return the results to the master. A program is also developed, and it can be running on any computers which support massage passing interface. Both the algorithm and the program are detailed discussed in the paper. We also include an application that uses the program to determine the positions and currents of PF coils in EAST. The program reach the target value within half an hour and yield a speedup rate of 5.21 on 8 CPUs. (author)
Current profile control and improved confinement in JT-60 tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Fujita, Takaaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
2002-08-01
In a tokamak, one of toroidal shaped magnetic confinement systems, a large toroidal current induced in a high temperature plasma produces helical magnetic fields to confine the plasma itself. Therefore, the spatial structure of confining magnetic fields can be determined by measuring the profile of the plasma current. This can lead to the understanding of the transport and stability properties of the plasma. In this review, a novel technique using the motional Stark effect to measure the current profile in the JT-60 tokamak is shown. With the aid of this technique, a new operational mode of a tokamak, with which the plasma confinement can be substantially improved, has been discovered. Recent experimental results of confinement improvement with the current profile control are presented. (author)
Current profile control and improved confinement in JT-60 tokamak plasmas
International Nuclear Information System (INIS)
In a tokamak, one of toroidal shaped magnetic confinement systems, a large toroidal current induced in a high temperature plasma produces helical magnetic fields to confine the plasma itself. Therefore, the spatial structure of confining magnetic fields can be determined by measuring the profile of the plasma current. This can lead to the understanding of the transport and stability properties of the plasma. In this review, a novel technique using the motional Stark effect to measure the current profile in the JT-60 tokamak is shown. With the aid of this technique, a new operational mode of a tokamak, with which the plasma confinement can be substantially improved, has been discovered. Recent experimental results of confinement improvement with the current profile control are presented. (author)
Two-fluid and parallel compressibility effects in tokamak plasmas
International Nuclear Information System (INIS)
The MHD, or single fluid, model for a plasma has long been known to provide a surprisingly good description of much of the observed nonlinear dynamics of confined plasmas, considering its simple nature compared to the complexity of the real system. On the other hand, some of the supposed agreement arises from the lack of the detailed measurements that are needed to distinguish MHD from more sophisticated models that incorporate slower time scale processes. At present, a number of factors combine to make models beyond MHD of practical interest. Computational considerations still favor fluid rather than particle models for description of the full plasma, and suggest an approach that starts from a set of fluid-like equations that extends MHD to slower time scales and more accurate parallel dynamics. This paper summarizes a set of two-fluid equations for toroidal (tokamak) geometry that has been developed and tested as the MH3D-T code [1] and some results from the model. The electrons and ions are described as separate fluids. The code and its original MHD version, MH3D [2], are the first numerical, initial value models in toroidal geometry that include the full 3D (fluid) compressibility and electromagnetic effects. Previous nonlinear MHD codes for toroidal geometry have, in practice, neglected the plasma density evolution, on the grounds that MHD plasmas are only weakly compressible and that the background density variation is weaker than the temperature variation. Analytically, the common use of toroidal plasma models based on aspect ratio expansion, such as reduced MHD, has reinforced this impression, since this ordering reduces plasma compressibility effects. For two-fluid plasmas, the density evolution cannot be neglected in principle, since it provides the basic driving energy for the diamagnetic drifts of the electrons and ions perpendicular to the magnetic field. It also strongly influences the parallel dynamics, in combination with the parallel thermal
Plasma driving system requirements for commercial tokamak fusion reactors
International Nuclear Information System (INIS)
The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration
Plasma driving system requirements for commercial tokamak fusion reactors
International Nuclear Information System (INIS)
The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration
Damping of kinetic Alfven eigenmodes in tokamak plasmas
International Nuclear Information System (INIS)
The ability to predict the stability of fast-particle-driven Alfven eigenmodes in burning fusion plasmas requires a detailed understanding of the dissipative mechanisms that damp these modes. In order to address this question, the linear gyro-kinetic, electromagnetic code LIGKA is employed to investigate their behaviour in realistic tokamak geometry. LIGKA is based on an eigenvalue formulation and self-consistently calculates the coupling of large-scale MHD modes to gyro-radius scale length kinetic Alfven waves. It uses the drift-kinetic HAGIS code to accurately describe the unperturbed particle orbits in general geometry. In addition, a newly developed antenna-like version of LIGKA allows for a frequency scan, analogous to an external antenna. With these tools the properties of the kinetically modified TAE in or near the gap (KTAE, radiative damping or 'tunnelling') and its coupling to the continuum close to the edge are numerically analysed. The results are compared with previous calculations based on fluid and other gyro-kinetic models. Also first linear calculations on cascade modes are presented. (author)
Control of safety factor profile in infinite dimension tokamak plasmas
International Nuclear Information System (INIS)
The increasing energy needs of the world population require the development, the control and the supply of new forms of energy. In this context, nuclear fusion is a track of extremely promising research. World project ITER is intended to prove the scientific and technical feasibility of nuclear fusion. One of the many key-goal is the control of the current profile spatial distribution in plasmas of tokamak, which is one of the main parameter for the stability and the performance of the experiments. The spatio-temporal evolution of this current is described by a set of nonlinear partial differential equations. In this document stabilization is proposed considering robust control of current profile spatial distribution in infinite dimension. Two approaches are proposed: the first one is based on sliding mode approach and the second one (of type proportional and proportional integral) is based on the Lyapunov functions in infinite dimension. The design of the control law is based on the 1D equation resistive diffusion of the magnetic flux. The control laws are calculated in infinite dimension without space discretization. (author)
Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST
Energy Technology Data Exchange (ETDEWEB)
Xu, X Q
2007-11-09
We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D ({psi}, {theta}, {epsilon}, {mu}) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices.
Control of plasma poloidal shape and position in the DIII-D tokamak
Energy Technology Data Exchange (ETDEWEB)
Walker, M.L.; Humphreys, D.A.; Ferron, J.R.
1997-11-01
Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks.
Dynamics and Feedback Control of Plasma Equilibrium Position in a Tokamak.
Burenko, Oleg
A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems. The major parameters governing the plasma equilibrium position stability of a tokamak are shown to be (1) external magnetic field decay index, (2) transformer iron core effect, (3) plasma current, (4) radial rate-of-change inductance parameter, (5) vertical rate-of-change inductance parameter, and (6) vacuum vessel eddy-current time constant. An important and unique result is derived, showing that for a vacuum vessel eddy-current time constant exceeding a certain value the vertical plasma equilibrium position is stable, in spite of an intentional vertical instability design represented by a negative decay index. It is shown that a tokamak design having a theoretical set of positive decay index, negative radical rate-of-change inductance parameter, and positive vertical rate-of-change inductance parameter is expected to have a better plasma equilibrium position stability tolerance than a tokamak design having the same set with the signs reversed. The results of an actual hardware ISX-A tokamak plasma displacement feed-back control system design are presented. It is shown that a theoretical design computer
A novel flexible field-aligned coordinate system for tokamak edge plasma simulation
Leddy, Jarrod; Romanelli, Michele; Shanahan, Brendan; Walkden, Nick
2016-01-01
Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (ie. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines begin to intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry...
On the Dirichlet Problem of Mixed Type for Lower Hybrid Waves in Axisymmetric Cold Plasmas
Lupo, Daniela; Monticelli, Dario D.; Payne, Kevin R.
2015-07-01
For a class of linear second order partial differential equations of mixed elliptic-hyperbolic type, which includes a well known model for analyzing possible heating in axisymmetric cold plasmas, we give results on the weak well-posedness of the Dirichlet problem and show that such solutions are characterized by a variational principle. The weak solutions are shown to be saddle points of natural functionals suggested by the divergence form of the PDEs. Moreover, the natural domains of the functionals are the weighted Sobolev spaces to which the solutions belong. In addition, all critical levels will be characterized in terms of global extrema of the functionals restricted to suitable infinite dimensional linear subspaces. These subspaces are defined in terms of a robust spectral theory with weights which is associated to the linear operator and is developed herein. Similar characterizations for the weighted eigenvalue problem and nonlinear variants will also be given. Finally, topological methods are employed to obtain existence results for nonlinear problems including perturbations in the gradient which are then applied to the well-posedness of the linear problem with lower order terms.
Thermal Event Recognition Applied to Protection of Tokamak Plasma-Facing Components
Martin, Vincent; Travere, Jean-Marcel; Bremond, François; Moncada, Victor; Dunand, Gwenaël
2010-01-01
Magnetic confinement fusion tokamaks are complex devices where a large amount of power is required to make the fusion reactions happen. In such experimental conditions, plasma facing components (PFCs) are subjected to high heat fluxes that can damage them. Machine protection functions must then be developed to operate current and future devices like ITER inthe safest way. In current tokamaks like Tore Supra, IR thermographic diagnostics based on image analysis and feedback control are used to...
Next-generation plasma control in the DIII-D tokamak
International Nuclear Information System (INIS)
The advanced tokamak (AT) operating mode, which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. This paper describes progress towards the DIII-D AT mission goal through both improvements in real-time computational hardware and control algorithm capability. A number of device constraints, some unique to DIII-D, and their impact on operational shape and position control are discussed. Some partial solutions are described
Transport simulation of a large-bootstrap-current Tokamak plasma driven by the ohmic seed current
International Nuclear Information System (INIS)
A large-bootstrap-current tokamak plasma driven by the ohmic seed current has been studied for inductively-operated ultra-long-pulse tokamak fusion reactors. The safety factor profile with the negative magnetic shear has been demonstrated only by the combination of the ohmic and bootstrap currents, by adjusting the density profile, and the bootstrap current fraction more than 80% is realized with the help of the increased q(0) value. 7 refs., 4 figs., 1 tab
Plasma-Facing Components in Tokamaks : Material Modification and Fuel Retention
Ivanova, Darya
2012-01-01
Fuel inventory and generation of carbon and metal dust in a tokamak are perceived to be serious safety and economy issues for the steady-state operation of a fusion reactor, e.g. ITER. These topics have been explored in this thesis in order to contribute to a better understanding and the development of methods for controlling and curtailing fuel accumulation and dust formation in controlled fusion devices. The work was carried out with material facing fusion plasmas in three tokamaks: TEXTOR ...
Plasma boundary determination in Damavand tokamak by using current filament method
International Nuclear Information System (INIS)
The shape and position of the plasma and consequently the plasma boundary are determined by using the Current Filament (Cf) method from the experimental data of the magnetic measurements in Damavand tokamak. The method can calculate the magnetic flux without solving the equilibrium equation directly by coupling with the Current Moment (CM) method. The plasma and current-carrying coils in the tokamak will be modeled by using this method as some virtual filaments that will enable us to calculate the flux and consequently the plasma boundary. To calculate the flux of these virtual filaments, one needs to determine the Green Function and the inverse by means of the Singular Value Decomposition (SVD) method. Finally, the model was evaluated by employing 12 independent pickup coils with mean error of less than 2%. The aim of this paper is to give a brief exposition of CF method applied in Damavand tokamak. (author)
Study of electron density and its fluctuations in tokamaks plasmas by fast infrared interferometry
International Nuclear Information System (INIS)
The electron density knowledge in tokamak plasma is fundamental for controlled fusion research. Its study can be made by interferometric measurement of plasma refraction index. Density and density fluctuation measurements are given for present and future tokamak, the wavelength used must be in the far infrared. The interferometer used type employs two identical lasers. Waveguide type submillimetric lasers, optically pumped by a CO2 laser, have been developed and optimized. Detectors used are Schottky diodes. The interferometer allows a radial study of the plasma and presents a great stability during the measurement
On the validity of drift-reduced fluid models for tokamak plasma simulation
Leddy, Jarrod; Romanelli, Michele
2015-01-01
Drift-reduced plasma fluid models are commonly used in plasma physics for analytics and simulations; however, the validity of such models must be veri?ed for the regions of parameter space in which tokamak plasmas exist. By looking at the linear behaviour of drift-reduced and full-velocity models one can determine that the physics lost through the simplification that the drift-reduction provides is important in the core region of the tokamak. It is more acceptable for the edge-region but one must determine speci?cally for a given simulation if such a model is appropriate.
International Nuclear Information System (INIS)
The accessibility of electron frequency range electromagnetic waves in magnetized inhomogeneous tokamak plasmas is studied using the cold plasma model. Simple equations for cutoff and resonance surfaces are obtained for parabolic density profile plasma. Using cartesian coordinates in the poloidal cross section of the tokamak, these equations are arranged in a convenient standard form, in order to facilitate the plotting of the various surfaces simultaneously. Graphical outputs for several cases are presented, together with the principal cutoff and resonance frequency profiles along the minor radius. (Author)
Edge Plasma Performance of Lower Hybrid Wave Injection on the HL-1M Tokamak
Institute of Scientific and Technical Information of China (English)
HONGWenyu; WANGEnyao; CAOJianyong; LIQiang
2001-01-01
Recently, the L-mode to H-mode (L-H) transition in tokamak plasma confinement was found to be related to the presence of the poloidal flow shear near the plasma edge. An important mechanism is the ion orbit loss caused by interaction with the limiter. A complementary explanation is the generation of poloidal flows by plasma fluctuations via the Reynolds stress and the poloidal spin-up of plasmas from poloidal asymmetryof particle and momentum sources.
Electron density and temperature determination in a Tokamak plasma using light scattering
International Nuclear Information System (INIS)
A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs
Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering
Stejner, M.; Korsholm, S. B.; Nielsen, S.K.; Salewski, M.; Bindslev, H.; Brezinsek, S.; Furtula, V.; Leipold, F.; Michelsen, P. K.; Meo, F.; Moseev, D.; Burger, A.; Kantor, M.; M.R. de Baar,
2012-01-01
We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with wave
Electron density and temperature determination in a Tokamak plasma using light scattering
International Nuclear Information System (INIS)
A theoretical foundation review for light scattering by plasmas is presented. Furthemore, a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, is included in a Tokamak plasma using spectral analysis of the scattered radiation. (author)
Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering
DEFF Research Database (Denmark)
Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh; Salewski, Mirko; Bindslev, Henrik; Brezinsek, S.; Furtula, Vedran; Leipold, Frank; Michelsen, Poul; Meo, Fernando; Moseev, Dmitry; Bürger, A.; Kantor, M.; de Baar, M.
2012-01-01
We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with...
Wall conditioning of the TBR-1 Tokamak by plasma generated by microwaves
International Nuclear Information System (INIS)
A new system of vaccum chamber wall conditioning in the TBR-1 Tokamak, using electron cyclotron resonance plasma of hydrogen for the discharge cleaning process is presented. The construction and performance of equipments are described, and the cleaning process to otimize the conditioning efficiency by chase of plasma parameters. (author)
Energy Technology Data Exchange (ETDEWEB)
Colunga S, S
1990-07-15
In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)
Non-axisymmetric ideal equilibrium and stability of ITER plasmas with rotating RMPs
Ham, C. J.; Cramp, R. G. J.; Gibson, S.; Lazerson, S. A.; Chapman, I. T.; Kirk, A.
2016-08-01
The magnetic perturbations produced by the resonant magnetic perturbation (RMP) coils will be rotated in ITER so that the spiral patterns due to strike point splitting which are locked to the RMP also rotate. This is to ensure even power deposition on the divertor plates. VMEC equilibria are calculated for different phases of the RMP rotation. It is demonstrated that the off harmonics rotate in the opposite direction to the main harmonic. This is an important topic for future research to control and optimize ITER appropriately. High confinement mode (H-mode) is favourable for the economics of a potential fusion power plant and its use is planned in ITER. However, the high pressure gradient at the edge of the plasma can trigger periodic eruptions called edge localized modes (ELMs). ELMs have the potential to shorten the life of the divertor in ITER (Loarte et al 2003 Plasma Phys. Control. Fusion 45 1549) and so methods for mitigating or suppressing ELMs in ITER will be important. Non-axisymmetric RMP coils will be installed in ITER for ELM control. Sampling theory is used to show that there will be significant a {{n}\\text{coils}}-{{n}\\text{rmp}} harmonic sideband. There are nine coils toroidally in ITER so {{n}\\text{coils}}=9 . This results in a significant n = 6 component to the {{n}\\text{rmp}}=3 applied field and a significant n = 5 component to the {{n}\\text{rmp}}=4 applied field. Although the vacuum field has similar amplitudes of these harmonics the plasma response to the various harmonics dictates the final equilibrium. Magnetic perturbations with toroidal mode number n = 3 and n = 4 are applied to a 15 MA, {{q}95}≈ 3 burning ITER plasma. We use a three-dimensional ideal magnetohydrodynamic model (VMEC) to calculate ITER equilibria with applied RMPs and to determine growth rates of infinite n ballooning modes (COBRA). The {{n}\\text{rmp}}=4 case shows little change in ballooning mode growth rate as the RMP is
DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section
International Nuclear Information System (INIS)
The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)
Designing a tokamak fusion reactor—How does plasma physics fit in?
Freidberg, J. P.; Mangiarotti, F. J.; Minervini, J.
2015-07-01
This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a , R 0 , B 0 , T i , T e , p , n , τ E , I . The one exception is the value of the toroidal current I , which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.
International Nuclear Information System (INIS)
Different bootstrap current formulations are implemented in a self-consistent equilibrium calculation obtained from a direct variational technique in fixed boundary tokamak plasmas. The total plasma current profile is supposed to have contributions of the diamagnetic, Pfirsch-Schlueter, and the neoclassical Ohmic and bootstrap currents. The Ohmic component is calculated in terms of the neoclassical conductivity, compared here among different expressions, and the loop voltage determined consistently in order to give the prescribed value of the total plasma current. A comparison among several bootstrap current models for different viscosity coefficient calculations and distinct forms for the Coulomb collision operator is performed for a variety of plasma parameters of the small aspect ratio tokamak ETE (Experimento Tokamak Esferico) at the Associated Plasma Laboratory of INPE, in Brazil. We have performed this comparison for the ETE tokamak so that the differences among all the models reported here, mainly regarding plasma collisionality, can be better illustrated. The dependence of the bootstrap current ratio upon some plasma parameters in the frame of the self-consistent calculation is also analysed. We emphasize in this paper what we call the Hirshman-Sigmar/Shaing model, valid for all collisionality regimes and aspect ratios, and a fitted formulation proposed by Sauter, which has the same range of validity but is faster to compute than the previous one. The advantages or possible limitations of all these different formulations for the bootstrap current estimate are analysed throughout this work. (author)
Energy Technology Data Exchange (ETDEWEB)
Andrade, Maria Celia Ramos; Ludwig, Gerson Otto [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: mcr@plasma.inpe.br
2004-07-01
Different bootstrap current formulations are implemented in a self-consistent equilibrium calculation obtained from a direct variational technique in fixed boundary tokamak plasmas. The total plasma current profile is supposed to have contributions of the diamagnetic, Pfirsch-Schlueter, and the neoclassical Ohmic and bootstrap currents. The Ohmic component is calculated in terms of the neoclassical conductivity, compared here among different expressions, and the loop voltage determined consistently in order to give the prescribed value of the total plasma current. A comparison among several bootstrap current models for different viscosity coefficient calculations and distinct forms for the Coulomb collision operator is performed for a variety of plasma parameters of the small aspect ratio tokamak ETE (Experimento Tokamak Esferico) at the Associated Plasma Laboratory of INPE, in Brazil. We have performed this comparison for the ETE tokamak so that the differences among all the models reported here, mainly regarding plasma collisionality, can be better illustrated. The dependence of the bootstrap current ratio upon some plasma parameters in the frame of the self-consistent calculation is also analysed. We emphasize in this paper what we call the Hirshman-Sigmar/Shaing model, valid for all collisionality regimes and aspect ratios, and a fitted formulation proposed by Sauter, which has the same range of validity but is faster to compute than the previous one. The advantages or possible limitations of all these different formulations for the bootstrap current estimate are analysed throughout this work. (author)
International Nuclear Information System (INIS)
In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)
Impedance of an intense plasma-cathode electron source for tokamak startup
Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.
2016-05-01
An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.
Divertor coil power supply in Aditya Tokamak for improved plasma operation
International Nuclear Information System (INIS)
The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a Tokamak with divertor configuration. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)
Positive plasma biasing in front of the lower hybrid grill of Castor tokamak
International Nuclear Information System (INIS)
Parasitic generation of suprathermal particles in front of lower hybrid (LH) antennas in tokamaks represents a serious danger for the parts of tokamak first wall connected with this region directly by magnetic field lines. Presence of electrons in radially very narrow wave-plasma interaction region, accelerated up to the energy 200 eV, has been proved recently on Castor tokamak by Langmuir probes (a substantial drop of floating potential is observed). Using emissive Langmuir probes, first experimental evidence of increase of plasma potential in the interaction region is given in this paper. This result confirms predictions of theory about the charge separation due to the escape of accelerated electrons with successive acceleration of plasma ions. (authors)
Kinetic shear Alfvén instability in the presence of impurity ions in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Lu, Gaimin; Shen, Y.; Xie, T.; He, Zhixiong; He, Hongda [Southwestern Institute of Physics, P. O. Box 432, Chengdu 610041 (China); Qi, Longyu [Institute for Fusion Theory and Simulation, Zhejiang University, Hangzhou 310027 (China); Cui, Shaoyan [School of Mathematics and Information, Ludong University, Yantai 264025 (China)
2013-10-15
The effects of impurity ions on the kinetic shear Alfvén (KSA) instability in tokamak plasmas are investigated by numerically solving the integral equations for the KSA eigenmode in the toroidal geometry. The kinetic effects of hydrogen and impurity ions, including transit motion, finite ion Larmor radius, and finite-orbit-width, are taken into account. Toroidicity induced linear mode coupling is included through the ballooning-mode representation. Here, the effects of carbon, oxygen, and tungsten ions on the KSA instability in toroidal plasmas are investigated. It is found that, depending on the concentration and density profile of the impurity ions, the latter can be either stabilizing or destabilizing for the KSA modes. The results here confirm the importance of impurity ions in tokamak experiments and should be useful for analyzing experimental data as well as for understanding anomalous transport and control of tokamak plasmas.
Effect of magnetic perturbations on the 3D MHD self-organization of shaped tokamak plasmas
Bonfiglio, D; Veranda, M; Chacón, L; Escande, D F
2016-01-01
The effect of magnetic perturbations (MPs) on the helical self-organization of shaped tokamak plasmas is discussed in the framework of the nonlinear 3D MHD model. Numerical simulations performed in toroidal geometry with the \\textsc{pixie3d} code [L. Chac\\'on, Phys. Plasmas {\\bf 15}, 056103 (2008)] show that $n=1$ MPs significantly affect the spontaneous quasi-periodic sawtoothing activity of such plasmas. In particular, the mitigation of sawtooth oscillations is induced by $m/n=1/1$ and $2/1$ MPs. These numerical findings provide a confirmation of previous circular tokamak simulations, and are in agreement with tokamak experiments in the RFX-mod and DIII-D devices. Sawtooth mitigation via MPs has also been observed in reversed-field pinch simulations and experiments. The effect of MPs on the stochastization of the edge magnetic field is also discussed.
Tokamak Plasmas : Measurement of temperature ﬂuctuations and anomalous transport in the SINP tokamak
Indian Academy of Sciences (India)
R Kumar; S K Saha
2000-11-01
Temperature ﬂuctuations have been measured in the edge region of the SINP tokamak. We ﬁnd that these ﬂuctuations have a comparatively high level (30–40%) and a broad spectrum. The temperature ﬂuctuations show a quite high coherence with density and potential ﬂuctuations and contribute considerably to the anomalous particle ﬂux.
Main Physical Factors Limiting the Accuracy of Polarimetric Measurements in Tokamak Plasma
Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco
The paper reviews and discusses the main factors, limiting the accuracy of polarimetric measurements in tokamak plasma. Theoretical methods, describing evolution of polarimetry state in tokamak plasma, are demonstrated not to contribute noticeably to inaccuracy at sufficiently short beam wavelengths. Based on the literature data as well as on our preliminary estimates it is possible to conclude that the following factors dominate: i) calibration procedure; ii) refraction in the inhomogeneous plasma; iii) influence of weak relativistic effects on plasma dielectric permittivity. The contribution of these factors to is within the range of several per cent. Other causes of measurement inaccuracies (absorption in plasma, diffraction of sounding beam, ray torsion, nonstationary processes in plasma) seem to be less significant.
Howard, N. T.; Holland, C.; White, A. E.; Greenwald, M.; Candy, J.
2016-01-01
The transport of heat in laboratory and astrophysical plasmas is dominated by the complex nonlinear dynamics of plasma turbulence. In magnetically confined plasmas used for fusion energy research, turbulence is responsible for cross-field transport that limits the performance of tokamak reactors. We report a set of novel gyrokinetic simulations that capture ion and electron-scale turbulence simultaneously, revealing the dynamics of cross-scale energy transfer and zonal flow modification that give rise to heat losses. Multi-scale simulations are required to match experimental ion and electron heat fluxes and electron profile stiffness, establishing the applicability of the newly discovered physics to experiment. Importantly, these results provide a likely explanation for the loss of electron heat from tokamak plasmas, the ‘great unsolved problem’ (Bachelor et al (2007 Plasma Sci. Technol. 9 312-87)) in plasma turbulence and the projected dominant loss channel in ITER.
Gridded ionization chamber for detection of x-ray wave activity in tokamak plasmas
International Nuclear Information System (INIS)
In order to carry out x-ray observations of magnetohydrodynamic wave activity of the plasma during DD and DT (deuterium-beam-heated deuterium and tritium plasmas, respectively) operation of the Tokamak Fusion Test Reactor (TFTR), we will need detectors not susceptible to nuclear radiation damage. We have investigated the use of gridded ionization chambers as fast nondamageable x-ray detectors. A prototype chamber is described which was tested on the PDX tokamak. These tests and laboratory tests with a pulsed x-ray source suggest that the detector has sufficient sensitivity and speed for the required measurements
Gridded ionization chamber for detecion of x-ray wave activity in tokamak plasmas
International Nuclear Information System (INIS)
In order to carry out X-ray observations of magnetohydrodynamic wave activity of the plasma during DD and DT (deuterium-beam-heated deuterium and tritium plasmas, respectively) operation of the Tokamak Fusion Test Reactor (TFTR), we will need detectors not susceptible to nuclear radiation damage. We have investigated the use of gridded ionization chambers as fast nondamageable X-ray detectors. A prototype chamber is described, which was tested on the PDX tokamak. These tests and laboratory tests with a pulsed X-ray source suggest that the detector has sufficient sensitivity and speed for the required measurements
Validity of Self-Organized Criticality model for the CASTOR tokamak edge plasmas
Czech Academy of Sciences Publication Activity Database
Ďuran, Ivan; Stöckel, Jan; Horáček, Jan; Jakubka, Karel; Kryška, Ladislav; Hron, Martin
volume 24B. Mulhouse : European Physical Society, 2000 - (Szegö, K.; Todd, T.; Zoletnik, S.), s. 1693-1696 - (Europhysics Conference Abstracts.. 24B). [European Physical Society Conference on Controlled Fusion and Plasma Physics/27th./. Budapest (HU), 12.06.2000-16.06.2000] Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, plasma Subject RIV: BL - Plasma and Gas Discharge Physics
International Nuclear Information System (INIS)
Influence of shear flows of the dense plasma created under conditions of the electron cyclotron resonance (ECR) gas breakdown on the plasma confinement in the axisymmetric mirror trap (''vortex'' confinement) was studied experimentally and theoretically. A limiter with bias potential was set inside the mirror trap for plasma rotation. The limiter construction and the optimal value of the potential were chosen according to the results of the preliminary theoretical analysis. This method of ''vortex'' confinement realization in an axisymmetric mirror trap for non-equilibrium heavy-ion plasmas seems to be promising for creation of ECR multicharged ion sources with high magnetic fields, more than 1 T.
García Carrasco, Álvaro
2014-01-01
Understanding of material migration and its impact on the formation of co-deposited mixed material layers on plasma-facing components is essential for the development of fusion reactors. This thesis focuses on this topic. It is based on experiments performed at JET and TEXTOR tokamaks. The major objectives were to determine: (i) fuel and impurity removal from plasma-facing components by ICWC in different gas mixtures, (ii) fuel and impurity transport connected to ICWC operation, (iii) plasma ...
International Nuclear Information System (INIS)
A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.
About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils
International Nuclear Information System (INIS)
In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate
Flow shear induced fluctuation suppression in finite aspect ratio shaped tokamak plasma
International Nuclear Information System (INIS)
The suppression of turbulence by the E x B flow shear and parallel flow shear is studied in an arbitrary shape finite aspect ratio tokamak plasma using the two point nonlinear analysis previously utilized in a high aspect rat ampersand tokamak plasma. The result shows that only the E x B flow shear is responsible for the suppression of flute-like fluctuations. This suppression occurs regardless of the plasma rotation direction and is therefore, relevant for the VH mode plasma core as well as for the H mode plasma edge. Experimentally observed in-out asymmetry of fluctuation reduction behavior can be addressed in the context of flux expansion and magnetic field pitch variation on a given flux surface. The adverse effect of neutral particles on confinement improvement is also discussed in the context of the charge exchange induced parallel momentum damping
A flexible software design to determine the plasma boundary in Damavand tokamak
International Nuclear Information System (INIS)
A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)
A flexible software design to determine the plasma boundary in Damavand tokamak
Ghadiri, Rasoul; Sadeghi, Yahya; Esteki, Mohammad Hossein
2014-06-01
A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code "The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)" was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C# language.
Kinetic Thermal Ions Effects on Alfvenic Fluctuations in Tokamak Plasmas
International Nuclear Information System (INIS)
Full text: The early observation of beta induced Alfven eigenmodes (BAE) and a variety of recent experimental observations have attracted attention on studying the low-frequency Alfvenic fluctuations in tokamaks. The generalized fishbone-like dispersion relation theoretical framework has been adopted for extending the hybrid model by taking into account both thermal ion compressibility and diamagnetic effects in addition to energetic particles (EP) kinetic behaviours. The extended model has been used for implementing an eXtended version of HMGC (XHMGC). In general, the new version of HMGC can have two species of kinetic particles. On one hand, one can use XHMGC for investigating thermal ion kinetic effects on Alfvenic modes driven by EP. In this case, EP dynamics contribute in the ideal MHD region; while wave-particle resonances with core-plasma ions are important only in a narrow inertial layer centred about the mode rational surface, where the dynamics of EP can be neglected due to their large perpendicular orbits (compared to the layer width). On the other hand, it may be interesting to use XHMGC as a tool to simulate two coexisting EP species, generated e.g. by both ICRH and NBI heating, in order to study linear excitation of Alfvenic fluctuations and Energetic Particle Modes (EPM), as well as the interplay between the respective nonlinear physics. Results of initial-value simulations show that the observed frequency is always slightly higher than the BAE accumulation point and is the same at different radial positions; consistent with the characteristics of a discrete BAE-SAW eigenmode (termed as kinetic BAE or KBAE); however, no discrete eigenmode is found within the gap when MHD is ideally stable. Meanwhile, preliminary simulations of KBAE/EPM driven by purely circulating EP have also been done. So far, the results show that the mode frequency is higher than either theoretical BAE accumulation point frequency or EP transit frequency, and increases with
3D simulation studies of tokamak plasmas using MHD and extended-MHD models
International Nuclear Information System (INIS)
The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω*i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code, studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data
Experience and technical issues of liquid lithium application as plasma facing material in tokamaks
International Nuclear Information System (INIS)
The following critical issues of liquid lithium used in tokamak conditions are considered: major physical properties of lithium, physico-chemical aspects of lithium interaction and compatibility with structural materials of fusion reactors. Lithium capillary-porous system (CPS) is considered as advanced plasma facing material for power fusion reactor and its main properties are presented. Review of plasma facing element (PFE) structures based on lithium CPS and tests results in T-11M, T-10 and FTU tokamaks are included. Brief review of projects of lithium limiter of FTU with active system for thermal stabilization and module of lithium divertor for KTM tokamak with liquid metal (Na-K) cooling system based on the lithium CPS use are presented.
3D simulation studies of tokamak plasmas using MHD and extended-MHD models
International Nuclear Information System (INIS)
The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented: high-β disruption studies in reversed shear plasmas using the MHD level MH3D code; ω*i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code; studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code; and unstructured mesh MH3D++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data. (author). 18 refs, 5 figs
Equilibrium Plasma Position Control for a Large Tokamak Using Modern Control Theory
Fukunishi, Kohyu; Saito, Seiji; Ogata, Atsushi; Ninomiya, Hiromasa
1980-09-01
Optimal control techniques are applied to maintain the plasma in its equilibrium position in a large tokamak. The application of the state space equation to plasma position control is also discussed. Optimal controls with states, which are plasma current, OH coil current and vertical field current, and integrated plasma displacement feedbacks are formulated as linear, time invariant expressions with quadratic performance indices. Effective plasma position control was obtained with integral state feedback in computer simulations for the JT-60. These control techniques will be applied to the JT-60.
Equilibrium plasma position control for a large tokamak using modern control theory
International Nuclear Information System (INIS)
Optimal control techniques are applied to maintain the plasma in its equilibrium position in a large tokamak. The application of the state space equation to plasma position control is also discussed. Optimal controls with states, which are plasma current, OH coil current and vertical field current, and integrated plasma displacement feedbacks are formulated as linear, time-invariant expressions with quadratic performance indices. Effective plasma position control was obtained with integral state feedback in computer simulations for the JT-60. These control techniques will be applied to the JT-60. (author)
Large potential change induced by pellet injection in JIPP T-IIU tokamak plasmas
International Nuclear Information System (INIS)
A large, rapid change in the local plasma potential is found to be induced by off-axis hydrogen ice-pellet injection into a tokamak plasma. The polarity of the rapid change is reversed when the pellet is injected into the upper and lower halves of the poloidal plasma cross-section. This change can be interpreted as being due to the gradient-B drift of particles in the high-density plasmas of the pellet cloud, before the increase of the plasma density due to the ablation becomes uniform on the magnetic surface. (author)
Advanced control of the Tokamak plasma shape and position by the quick response power supply
International Nuclear Information System (INIS)
The research on large tokamaks to get high parameter plasmas has been greatly extended. However, a number of engineering problems such as plasma vertical instability and unexpected pulse termination are still serious. For this reason the control of poloidal field power supplies employed to maintain the plasma in stable equilibrium with complex X points around plasma have been more and more important. Hybrid matrix control of shape and position for changing plasmas and development of the quick response power amplifier are reported. (author). 2 refs.; 6 figs.; 1 tab
Intrinsic plasma rotation determined by neoclassical toroidal plasma viscosity in tokamaks
International Nuclear Information System (INIS)
Intrinsic toroidal plasma rotation due to the neoclassical toroidal plasma viscosity (NTV) effect induced by a three-dimensional helical magnetic field ripple in tokamaks is investigated in this paper. The intrinsic rotation is determined self-consistently by searching for the roots of the ambipolarity constraint, after evaluation of the particle fluxes from the numerical modelling. In the low-collisionality case, there are three roots, in which two are stable roots. One corresponds to the ‘ion root’ in the counter-current direction, and the other stable one corresponds to the ‘electron root’ in the co-current direction, near which the electron flux is dominant. Both of the two stable roots scale like the diamagnetic frequency. In the high-collisionality case, there is only one ‘ion’ root. The application of this modelling for International Thermonuclear Experimental Reactor (ITER) cases is discussed. In a large range of plasma radii, there are three roots. The NTV torque drives plasma rotation in ITER towards one of the stable roots, depending on the initial condition. The amplitudes of the electron roots near the pedestal in both baseline and steady-state scenarios are much larger than that of the ion roots. The amplitudes of the NTV torque density and the electron roots near the pedestal increase with increasing height of the temperature pedestal in the ITER baseline scenario. (paper)
TPX diagnostics for tokamak operation, plasma control and machine protection
International Nuclear Information System (INIS)
The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process
Cherenkov-type diagnostics of fast electrons within tokamak plasmas
Jakubowski, Lech; Sadowski, Marek J.; Zebrowski, Jaroslaw; Malinowski, Karol; Rabinski, Marek; Jakubowski, Marcin J.; Mirowski, Robert
2014-05-01
This paper presents a summary of the most important results of fast electron measurements performed so far within different tokamaks by means of Cherenkov-type detectors. In the ISTTOK tokamak (IPFN, IST, Lisboa, Portugal), two measuring heads were applied, each equipped with four radiators made of different types of alumina-nitrate poly-crystals. A two-channel measuring head equipped with diamond radiators was also used. Within the COMPASS tokamak (IPP AS CR, Prague, Czech Republic) some preliminary measurements have recently been performed by means of a new single-channel Cherenkov-type detector. The experimental data from the TORE SUPRA tokamak (CEA, IFRM, Cadarache, France), which were collected by means of a DENEPR-2 probe during two recent experimental campaigns, have been briefly analyzed. A new Cherenkov probe (the so-called DENEPR-3) has been mounted within the TORE SUPRA machine, but the electron measurements could not be performed because of the failure of this facility. Some conclusions concerning the fast electron emission are presented.
Diagnostic applications of transient synchrotron radiation in tokamak plasmas
International Nuclear Information System (INIS)
Transient radiation, resulting from a brief, deliberate perturbation of the velocity distribution of superthermal tokamak electrons, can be more informative than the steady background radiation that is present in the absence of the perturbation. It is possible to define a number of interesting inverse problems, which exploit the two-dimensional frequency-time data of the transient radiation signal. 17 refs
The Averaged Fokker - Planck equation in tokamak plasma
International Nuclear Information System (INIS)
In this paper the numerical code which has been developed to solve averaged Fokker-Planck equation, and its applications for studying the time evolution of the electron distribution function in tokamak device of medium size and performances are discussed. The electron collisions and DC electric field effects are analysed in details
2-D Imaging of Electron Temperature in Tokamak Plasmas
Energy Technology Data Exchange (ETDEWEB)
T. Munsat; E. Mazzucato; H. Park; C.W. Domier; M. Johnson; N.C. Luhmann Jr.; J. Wang; Z. Xia; I.G.J. Classen; A.J.H. Donne; M.J. van de Pol
2004-07-08
By taking advantage of recent developments in millimeter wave imaging technology, an Electron Cyclotron Emission Imaging (ECEI) instrument, capable of simultaneously measuring 128 channels of localized electron temperature over a 2-D map in the poloidal plane, has been developed for the TEXTOR tokamak. Data from the new instrument, detailing the MHD activity associated with a sawtooth crash, is presented.
2-D Imaging of Electron Temperature in Tokamak Plasmas
International Nuclear Information System (INIS)
By taking advantage of recent developments in millimeter wave imaging technology, an Electron Cyclotron Emission Imaging (ECEI) instrument, capable of simultaneously measuring 128 channels of localized electron temperature over a 2-D map in the poloidal plane, has been developed for the TEXTOR tokamak. Data from the new instrument, detailing the MHD activity associated with a sawtooth crash, is presented
Energy Technology Data Exchange (ETDEWEB)
Perez-Navarro Gomerz, A.; Zurro Hernandez, B.
1976-07-01
A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs.
Spatially resolved soft x-ray (1 - 33 nm) spectroscopy of tokamak plasmas
International Nuclear Information System (INIS)
We describe the space-resolved soft x-ray (1 - 33 nm) instrumentation developed for the Tore Supra Tokamak. By using a programmable hydraulic jack to move the spectrometer, several spatial profiles (up to ten) of many impurity lines are obtained during a single plasma discharge, with a time resolution which can be as short as 600 ms
Transport of energetic ions in MHD-active high-beta plasmas of spherical tokamaks
International Nuclear Information System (INIS)
It is shown that high β (β is the ratio of plasma pressure to the magnetic field pressure) may deteriorate the confinement of trapped energetic ions in spherical tokamaks (ST) during MHD events, such as sawtooth oscillations and internal reconnection events (IRE). This result indicates that moderate rather than very high β may be preferable in STs. (author)
Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas
DEFF Research Database (Denmark)
Militello, F; Fundamenski, W; Naulin, Volker; Nielsen, Anders Henry
2012-01-01
The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machines...
Plasma vortexes induced by an external rotating helical magnetic perturbation in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Pankratov, I.M. [Institute of Plasma Physics, National Science Center ' Kharkov Institute of Physics and Technology' , Akademicheskaya str., 1, 61108 Kharkov (Ukraine)]. E-mail: pankratov@kipt.kharkov.ua; Omelchenko, A.Ya. [Institute of Plasma Physics, National Science Center ' Kharkov Institute of Physics and Technology' , Akademicheskaya str., 1, 61108 Kharkov (Ukraine); Olshansky, V.V. [Institute of Plasma Physics, National Science Center ' Kharkov Institute of Physics and Technology' , Akademicheskaya str., 1, 61108 Kharkov (Ukraine)
2005-08-01
The occurrence of two or four vortexes per one poloidal perturbation period has been found near the resonant surface as a plasma motion response on the penetration of an external low frequency helical magnetic perturbation in tokamaks. The investigation is carried out on the basis of the two-fluid MHD equations in the linear approximation for the cylindrical model.
Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code
Directory of Open Access Journals (Sweden)
Yutthapong Pinanroj
2014-04-01
Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.
Edge plasma characteristics in the helicity injected torus (HIT-II) spherical tokamak
International Nuclear Information System (INIS)
The helicity injected torus (HIT-II) device is a spherical tokamak capable of both inductive (Ohmic) and co-axial helicity injection (CHI) current drive. The HIT-II plasma edge, in both Ohmic and CHI discharges, has been characterized using a triple Langmuir probe. An Ohmic discharge develops in two phases, a slide-away phase followed by a normal Ohmic discharge. During the normal Ohmic discharge, the floating potential is negative, just as in a conventional large-aspect-ratio tokamak. The plasma density increases sharply from the plasma edge into the centre. The auto-power spectrum of Ohmic plasma edge fluctuations shows a nearly constant auto-power at low frequencies, with auto-power decreasing at higher frequencies, similar to observations in conventional large-aspect-ratio tokamaks. In HIT-II CHI discharges, the magnetic field lines at the plasma edge are clearly connected to the injector electrodes, as expected. However, the time-evolution of the floating potential in the core plasma is significantly different from that of the edge, which may indicate a decoupling of the core plasma from the CHI electrodes. Finally, the fluctuations at the edge of high-performance CHI discharges exhibit a coherent oscillation at a frequency similar to that of the observed n = 1 mode
Plasma diagnostics at Aditya Tokamak by two views visible light tomography
Energy Technology Data Exchange (ETDEWEB)
Goswami, Mayank, E-mail: mggm1982@gmail.com [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur (India); Munshi, Prabhat [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur (India); Department of Mechanical Engineering, Indian Institute of Technology, Kanpur (India); Saxena, Anupam [Department of Mechanical Engineering, Indian Institute of Technology, Kanpur (India); Kumar, Manoj; Kumar, Ajai [Institute for Plasma Research (India)
2014-11-15
Graphical abstract: - Highlights: • Improved algorithm works equally well for central as well as for peripherical plasma regions. • Entropy optimized smoothening parameters eliminate user dependencies. • Real time fusion grade plasma diagnostics images. - Abstract: This visible light computerized tomography exercise is a part of a project to establish an auxiliary imaging method to assist other imaging facilities at the Institute of Plasma Research (IPR), India. Space constraints around Aditya Tokamak allow only two orthogonal ports. Each port has one detector array (64 sensors) sensitive to the visual spectrum emitted by H{sub α} emission. The objective here is to report the developments on limited view tomography for hot plasma imaging. Spatially filtered entropy maximization algorithm with non-uniform discretization grids is employed. Estimation of unique kernel smoothening parameters (mask size and exponent factor) depends on entropy function and projection data. It removes requirement of any arbitrary/user-based decision for choosing a regularization factor thus minimizes the chance for biasedness or errors. Synthetic projection data is used to analyse the performance of this modification. The error band in the process of recovery remains under acceptable level (less than 15%) irrespective of the origin of the emissions from the core. Reconstructed hot plasma images/profiles from Aditya Tokamak are shown. These profiles may improve the current understanding about (a) plasma–wall interaction or edge plasma turbulence, (b) control and generation of plasma and (c) correlations between theoretical and engineering advancements in Tokamak reactors.
Plasma diagnostics at Aditya Tokamak by two views visible light tomography
International Nuclear Information System (INIS)
Graphical abstract: - Highlights: • Improved algorithm works equally well for central as well as for peripherical plasma regions. • Entropy optimized smoothening parameters eliminate user dependencies. • Real time fusion grade plasma diagnostics images. - Abstract: This visible light computerized tomography exercise is a part of a project to establish an auxiliary imaging method to assist other imaging facilities at the Institute of Plasma Research (IPR), India. Space constraints around Aditya Tokamak allow only two orthogonal ports. Each port has one detector array (64 sensors) sensitive to the visual spectrum emitted by Hα emission. The objective here is to report the developments on limited view tomography for hot plasma imaging. Spatially filtered entropy maximization algorithm with non-uniform discretization grids is employed. Estimation of unique kernel smoothening parameters (mask size and exponent factor) depends on entropy function and projection data. It removes requirement of any arbitrary/user-based decision for choosing a regularization factor thus minimizes the chance for biasedness or errors. Synthetic projection data is used to analyse the performance of this modification. The error band in the process of recovery remains under acceptable level (less than 15%) irrespective of the origin of the emissions from the core. Reconstructed hot plasma images/profiles from Aditya Tokamak are shown. These profiles may improve the current understanding about (a) plasma–wall interaction or edge plasma turbulence, (b) control and generation of plasma and (c) correlations between theoretical and engineering advancements in Tokamak reactors
International Nuclear Information System (INIS)
Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length
Determination of plasma column transverse section in the TBR-1 tokamak
International Nuclear Information System (INIS)
The temporal evolution of plasma column transverse section in the TBR-1 tokamak is determined. The experimental melhod is based on the simulation of toroidal current distribution in plasma by a set of toroidal filaments. The currents in these filaments are determined by minimization of square error between the magnetic field produced by filaments and the field measured into the tokamak vacuum vessel. For the measurement of magnetic field, twenty small magnetic coils were constructed and installated in the region protected by current limiters. The plasma column transverse cross section is determined by poloidal field produced by the currents in filaments. The multipole moments of plasma current distribution and the Λ Shafranov parameter were obtained. (M.C.K.)
Upgrade of plasma density feedback control system in HT-7 tokamak
Institute of Scientific and Technical Information of China (English)
ZHAO Da-Zheng; LUO Jia-Rong; LI Gang; JI Zhen-Shan; WANG Feng
2004-01-01
The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.
Influence of collisions on parametric instabilities induced by lower hybrid waves in tokamak plasmas
Castaldo, C.; Di Siena, A.; Fedele, R.; Napoli, F.; Amicucci, L.; Cesario, R.; Schettini, G.
2016-01-01
Parametric instabilities induced at the plasma edge by lower hybrid wave power externally coupled to tokamak plasmas have, via broadening of the antenna spectrum, strong influence on the power deposition and current drive in the core. For modeling the parametric instabilities at the tokamak plasma edge in lower hybrid current drive experiments, the effect of the collisions has been neglected so far. In the present work, a specific collisional parametric dispersion relation, useful to analyze these nonlinear phenomena near the lower hybrid antenna mouth, is derived for the first time, based on a kinetic model. Numerical solutions show that in such cold plasma regions the collisions prevent the onset of the parametric instabilities. This result is important for present lower hybrid current drive experiments, as well as in fusion reactor scenarios.
Observation of ICRF [ion cyclotron range of frequencies] wave-packet propagation in a tokamak plasma
International Nuclear Information System (INIS)
Experimental observation of ICRF wave-packet propagation in a tokamak plasma is reported. Studies were carried out in the Caltech Research Tokamak in a pure hydrogen plasma and in a regime where fast-wave damping was sufficiently small to permit multiple toroidal transits of the wave-packet. Waves were launched by exciting a small loop antenna with a short burst of rf current and were detected with shielded magnetic probes. Probe scans revealed a large increase in wave-packet amplitude at smaller minor radii, and the packet velocity was found to be independent of radial position. Measurement of the packet transit time yielded direct information about the wave group velocity. Packet velocity was investigated as a function of the fundamental excitation frequency, plasma density, and toroidal magnetic field. Results are compared with the predictions of a cold plasma model which includes a vacuum layer at the edge. 24 refs., 8 figs
Startup of Plasma Current in J-TEXT Tokamak Prompted by the Hα Line Emission Criterion
Institute of Scientific and Technical Information of China (English)
GAO Li; ZHUANG Ge; HU Xiwei; ZHANG Ming
2009-01-01
An Hα line-emission detection system was developed on the joint texas experimental tokamak (J-TEXT), which is used to determine the Hα emission level during the gas breakdown and hereafter to control the startup of the plasma current. The detector consists of an Hα in-terference filter, a focusing lens, a photodiode and a preamplifier. In the J-TEXT operation, the Hα emission is taken as a monitor signal which is highly sensitive to the generation of a plasma.Furthermore, the power supply control system using the above signal as an input is capable of de-termining whether and when to fire the Ohmic heating capacitor banks, which are applied to drive the plasma current ramp-up. The experimental results confirm that the Hα emission criterion is acceptable for controlling the plasma current promotion in the J-TEXT tokamak.
High βp plasma formation using off-axis ECCD in Ohmic heated plasma in the spherical tokamak QUEST
Directory of Open Access Journals (Sweden)
Mishra Kishore
2015-01-01
Full Text Available High poloidal beta (ɛβp ~ 1 operation in steady state condition in tokamaks is of great interest and has previously been demonstrated using NBI, LHCD and low current (Ip plasma for a short time (<0.5 s. A very few experiments however, have been performed towards the investigation of highest obtainable βp in tokamak plasma. In this work we report the first result of high βp production and its sustainment though an off axis ECCD at two different frequencies (fundamental and second harmonic in Ohmic (OH target plasma. With application of ECCD, plasma βp increased to encounter an equilibrium limit and the standard limiter configuration is transformed to an Inboard Poloidal field Null (IPN configuration. Both off-axis and on-axis ECCD is studied and found to have some distinctive features, which are discussed in this paper.
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe
Czech Academy of Sciences Publication Activity Database
Walkden, N.R.; Adámek, Jiří; Allan, S.; Dudson, B.D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, Michael
2015-01-01
Roč. 86, č. 2 (2015), 023510-023510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ball pen probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://dx.doi.org/10.1063/1.4908572
On-line Model Structure Selection for Estimation of Plasma Boundary in a Tokamak
Czech Academy of Sciences Publication Activity Database
Škvára, Vít; Šmídl, Václav; Urban, Jakub
Bristol: Institute of Physics Publishing Ltd, 2015, 012010. ISSN 1742-6588. [12th European Workshop on Advanced Control and Diagnosis (ACD 2015). Pilsen (CZ), 19.11.2015] Institutional support: RVO:67985556 ; RVO:61389021 Keywords : model structure estimation * tokamak * plasma boundary Subject RIV: BC - Control Systems Theory; BL - Plasma and Gas Discharge Physics (UFP-V) http://library.utia.cas.cz/separaty/2015/AS/skvara-0450942.pdf
Density peaking in the JFT-2M tokamak plasma with counter neutral beam injection
International Nuclear Information System (INIS)
A significant particle pinch and reduction of the effective thermal diffusivity are observed after switching the neutral beam direction from co- to counter- injection in the JFT-2M tokamak. A time delay in the occurrence of density peaking to that of plasma rotation is found. This shows that the particle pinch is related to the profile of the electric field as determined by the plasma rotation profile. The measured particle flux shows qualitative agreement with the theoretically-predicted inward pinch. (author)
Theory-based scaling of the SOL width in circular limited tokamak plasmas
Czech Academy of Sciences Publication Activity Database
Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Gunn, J. P.; Horáček, Jan; Kočan, M.; LaBombard, B.; Silva, C.
2013-01-01
Roč. 53, č. 12 (2013), s. 122001-122001. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : SOL * tokamak * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/12/122001/pdf/0029-5515_53_12_122001.pdf
Energy Confinement of High-Density Pellet-Fueled Plasmas in the Alcator C Tokamak
Greenwald, M.; Gwinn, D.; Milora, S.; Parker, J.; Parker, R.; Wolfe, S.; Besen, M.; Camacho, F.; Fairfax, S.; Fiore, C.; Foord, M.; Gandy, R.; Gomez, C.; Granetz, R.; Labombard, B.; Lipschultz, B.; Lloyd, B.; Marmar, E.; McCool, S.; Pappas, D.; Petrasso, R.; Pribyl, P.; Rice, J.; Schuresko, D.; Takase, Y.; Terry, J.; Watterson, R.
1984-07-01
A series of pellet-fueling experiments has been carried out on the Alcator C tokamak. High-speed hydrogen pellets penetrate to within a few centimeters of the magnetic axis, raise the plasma density, and produce peaked density profiles. Energy confinement is observed to increase over similar discharges fueled only by gas puffing. In this manner record values of electron density, plasma pressure, and Lawson number (n τ) have been achieved.
Ideal Magnetohydrodynamic stabilities in JT-60 and DIII-D tokamak plasmas
International Nuclear Information System (INIS)
Magnetohydrodynamic (MHD) instabilities of plasmas with a good (enhanced) energy confinement, obtained in the H-mode or after pellet injection in JT-60 and DIII-D tokamaks, are investigated. In enhanced-confinement plasmas, many kinds of MHD instabilities occurred and they inhibit further improvement. In the present thesis, therefore, the mechanism of MHD stability is investigated by computational analysis, and ways to further improve the confinement by suppression of MHD instabilities are discussed. (J.P.N.)
Measurement of Sheared Flows in the Edge Plasma of the CASTOR Tokamak
Czech Academy of Sciences Publication Activity Database
Brotánková, Jana; Stöckel, Jan; Horáček, Jan; Seidl, Jakub; Ďuran, Ivan; Hron, Martin; Van Oost, G.
2009-01-01
Roč. 35, č. 11 (2009), s. 980-986. ISSN 1063-780X. [IAEA Technical Meeting on Research Using Small Fusion Devices/18th./. Alushta (Krym), 25.09.2008-27.09.2008] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * shear ed flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.584, year: 2009 http://www.springerlink.com/content/u571504gmq118314/
Parameterization of Balmer-alpha asymmetric line shape in tokamak SOL plasmas
International Nuclear Information System (INIS)
A parameterization of the Balmer-alpha spectral line shape asymmetry in the tokamak scrape-off layer (SOL) plasmas is suggested, which describes the contribution of nonMaxwellian components of the neutral atom velocity distribution function. Parameterization is needed for a fast-routine interpretation of high-resolution spectroscopy data and should be incorporated into the algorithms for the recovery of hydrogen neutral atom parameters in the SOL. We illustrate the efficiency of the parameterization on the example of spectral data calculated using the predictive modeling of the International Thermonuclear Experimental Reactor (ITER) tokamak operation
Directory of Open Access Journals (Sweden)
Minashin P.V.
2015-01-01
Full Text Available A method of spectroscopic diagnostics of the average perpendicular-to-magnetic-field momentum of the superthermal component of the electron velocity distribution (EVD, based on the high-number-harmonic electron cyclotron (EC radiation, is suggested for nuclear fusion-reactor plasmas under condition of a strong auxiliary heating (e.g. in tokamak DEMO, a next step after tokamak ITER. The method is based on solving an inverse problem for reconstruction of the EVD in parallel and perpendicular-to-magnetic-field components of electron momentum at high and moderate energies responsible for the emission of the high-number-harmonic EC radiation.
International Nuclear Information System (INIS)
In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs
Velocity-space stability of counterstreaming-ion tokamak plasmas
International Nuclear Information System (INIS)
The steady-state CBT/CIT ion velocity distributions are stable to all electrostatic and electromagnetic infinite-medium modes. This stability is due to the large thermal spread of the distributions (T/sub hot/ approximately greater than 1.5 T/sub e/, v/sub ti//u greater than 1/3), and to the fact that tokamak equilibrium allows the mean streaming velocity, u, to be only a small fraction of the Alfven velocity
International Nuclear Information System (INIS)
Plasma disruption in a tokamak fusion reactor is supposed to be the most critical phenomenon concerning the structural integrity of the plasma-neighbouring components. This paper shows the effects of an electrically conducting first wall on the mechanical loading of the blankets. The spatial distribution, time history and maximum of the induced forces and mechanical stresses in the structure are very sensitive to the electrical design of the first wall. The paper also introduces a method for dynamic modelling of the plasma current which is inductively coupled to the current in the structure, i.e. the time evolution of the plasma current is not prescribed but part of the results. (orig.)
Reconstruction of plasma current profile of tokamaks using combinatorial optimization techniques
International Nuclear Information System (INIS)
New methods to reconstruct plasma shape and plasma current distribution from magnetic measurements are proposed. The reconstruction of plasma current profile from magnetic measurements is regarded as an optimum allocation problem of currents into cross section of the vacuum vessel of the tokamak. For solving this optimization problem, the authors use two types of solutions: a genetic algorithm and a combined method of a Hopfield neural network and a genetic algorithm. The effectiveness of these methods is shown by the application of these techniques to JT-60U plasmas
Poloidal rotation induced by injecting lower hybrid waves in tokamak plasma edge
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
The poloidal rotation of the magnetized edge plasma in tokamak driven by theponderomotive force which is generated by injecting lower hybrid wave(LHW) electric field hasbeen studied. The LHW is launched from a waveguide in the plasma edge, and by Brambilla’sgrill theory, analytic expressions for the wave electric field in the slab model of an inhomogeneouscold plasma have been derived. It is shown that a strong wave electric field will be generated inthe plasma edge by injecting LH wave of the power in MW magnitude, and this electric field willinduce a poloidal rotation with a sheared poloidal velocity.PACS: 52.55.Fa
Role of Pressure Gradient on Intrinsic Toroidal Rotation in Tokamak Plasmas
International Nuclear Information System (INIS)
The toroidal plasma rotation generated by the external momentum input and by the plasma itself (intrinsic rotation) has been separated through a novel momentum transport analysis in the JT-60U tokamak device. The toroidal rotation, which is not determined by the momentum transport coefficients and the external momentum input, has been observed. It is found that this intrinsic rotation is locally determined by the local pressure gradient and increases with increasing pressure gradient. This trend is almost the same for various plasmas: low and high confinement mode, co and counterrotating plasmas
International Nuclear Information System (INIS)
Radial electric field shear and poloidal plasma rotation are important factors affecting transport and confinement in tokamaks. Alteration of the electric field and plasma rotation in the vicinity of magnetic islands is also an important factor in tokamak plasma confinement. In the STOR-M tokamak, fast (∼1 ms) simultaneous alterations of the radial electric field, plasma rotation (Mparallel = 0-0.4 in the plasma current direction), floating potential fluctuations in the periphery and MHD activity generated by rotating islands have been observed experimentally during normal ohmic discharges. The observed time and magnitude of the changes depend on the average electron density and poloidal beta at the beginning of the discharge. In discharges with high initial poloidal beta these changes are accompanied by a reduction in Hα emission and an increase in the line averaged density. Drastic decreases in Hα and increases in line averaged electron density and estimation of poloidal beta suggest that STOR-M confinement is significantly affected in ohmic discharges without an external additional energy input or biasing. MHD activity in STOR-M is damped when a negative electric field is observed at the limiter region of the plasma edge. MHD frequency is observed to decrease with the negative electric field
International Nuclear Information System (INIS)
High performance operation with Internal Transport Barrier (ITB) is effective to improve the core plasma confinement in the future fusion reactor. Numerous plasma experiments with ITB were confirmed in the reversed magnetic shear. It is considered that ITB formation could be controlled by external fueling. In this study, firstly, the feasibility of pellet injection condition is simulated in tokamak reactor. Secondly, the effect of the pellet injection on the core plasma profile and ITB formation is analyzed at tokamak and helical reactors. Simulations are carried out using the toroidal transport linkage code TOTAL. In case of the operation with pellet injection from high magnetic-field side (HFS), the feasibility of pellet injection condition for ITB formation is demonstrated in the ITER-like tokamak reactor, TR-1. In both tokamak and helical reactors, it is shown that pellet injection depth is not related to the position of ITB formation, but it has significant effect to the radial profile. In helical case, wide-ranged ITB is formed when the pellet is injected centrally. (author)
Analysis and optimization of the edge plasma of joule tokamak discharges
International Nuclear Information System (INIS)
The following topics are discussed: Introduction (the principle of a Tokamak); Diagnostics of the edge plasma (Thomson scattering, Langmuir probes, laser-induced fluorescence); Computer Simulations of the Plasma Edge (atom physical data, transport models, neutral gas models); Characterization of the Edge Plasma (electron temperature and density, particle fluxes and recycling, transport within the plasma edge, edge limits, isotope effect); emission and transport of impurities (physical sputtering, plasma induced discharges, evaporation, chemical spray, sublimation, transport of impurities, profiles of impurities' concentration); Joule Discharges of high density (phenomenology, time dependence, particle transport in the edge plasma, correlations of the edge parameters, comparison with the detached plasma); Injection of Hydrogen Pellets (pellet ablation, plasma response, relaxation after pellet injection); Synopsis. (AH)
Energy Technology Data Exchange (ETDEWEB)
Meslin, B
1998-04-30
Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density
Aspect ratio scaling of toroidal plasma equilibria and the tokamak bootstrap effect
International Nuclear Information System (INIS)
The aspect ratio scaling of toroidal plasma equilibria is examined using a parametrization of an exact Solov close-quote ev solution to the Grad endash Shafranov equation in Boozer coordinates. The equilibrium analysis suggests that simultaneous enhancements in magnetohydrodynamic (MHD) stability and the bootstrap effect are possible in tight aspect ratio (A→1) tokamaks. The fundamental physical mechanism causing the enhancements is shown to be the natural increase of the MHD safety factor q in tight aspect ratio toroidal geometries. The results of the scaling model suggest that the lowest bootstrap current fractions are obtained in tokamaks with aspect ratios A∼3. It is also shown that a tight aspect ratio bootstrapped tokamak can be a weakly paramagnetic device. copyright 1997 American Institute of Physics
International Nuclear Information System (INIS)
The role of impurity dynamics in resistivity gradient driven turbulence is investigated in the context of modeling tokamak edge plasma phenomena. The effects of impurity concentration fluctuations and gradients on the linear behavior of rippling instabilities and on the nonlinear evolution and saturation of resistivity gradient driven turbulence are studied both analytically and computationally. At saturation, fluctuation levels and particle and thermal diffusivities are calculated. In particular, the mean-square turbulent radial velocity is given by 2> = (E0L/sub s/B/sub z/)2 (L/sub/eta/-1 + L/sub z-1)2. Thus, edged peaked impurity concentrations tend to enhance the turbulence, while axially peaked concentrations tend to quench it. The theoretical predictions are in semi-quantitative agreement with experimental results from the TEXT, Caltech, and Tosca tokamaks. Finally, a theory of the density clamp observed during CO-NBI on the ISX-B tokamak is proposed
International Nuclear Information System (INIS)
The research we have accomplished during the past year has focussed on ICRF coupling, heating and breakeven studies for tokamaks and ECRF fundamental second harmonic heating in tandem mirrors. The studies have included ICRF Fokker-Planck heating and breakeven studies for large tokamaks such as JET, fundamental work on a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas, a formulation and code development for ICRF waveguide coupling in tokamak edge regions. ECRF ray tracing studies have been carried out for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA. The two-dimensional velocity space, time dependent Fokker-Planck heating studies have concentrated on D-T breakeven scenarios for fundamental minority deuterium and second harmonic tritium regimes
Real-time control of Tokamak plasmas: from control of physics to physics-based control
International Nuclear Information System (INIS)
Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have
International Nuclear Information System (INIS)
The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support
Dynamic Optimization of Trajectory for Ramp-up Current Profile in Tokamak Plasmas
Ren, Zhigang; Ou, Yongsheng
2016-01-01
In this paper, we consider an open-loop, finite-time, optimal control problem of attaining a specific desired current profile during the ramp-up phase by finding the best open-loop actuator input trajectories. Average density, total power, and plasma current are used as control actuators to manipulate the profile shape in tokamak plasmas. Based on the control parameterization method, we propose a numerical solution procedure directly to solve the original PDE-constrained optimization problem using gradient-based optimization techniques such as sequential quadratic programming (SQP). This paper is aimed at proposing an effective framework for the solution of PDE-constrained optimization problem in tokamak plasmas. A more user-friendly and efficient graphical user interface (GUI) is designed in MATLAB and the numerical simulation results are verified to demonstrate its applicability. In addition, the proposed framework of combining existing PDE and numerical optimization solvers to solve PDE-constrained optimiz...
Peeling-off of the external kink modes at tokamak plasma edge
Zheng, L J
2014-01-01
It is pointed that there is a current jump between the edge plasma inside the last closed magnetic surface and the scrape-off layer and the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the diverters. In particular, the peeling or peeling-ballooning modes can become the "peeling-off" modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.
Peeling-off of the external kink modes at tokamak plasma edge
International Nuclear Information System (INIS)
It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture
Peeling-off of the external kink modes at tokamak plasma edge
Energy Technology Data Exchange (ETDEWEB)
Zheng, L. J. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Furukawa, M. [Graduate School of Engineering, Tottori University, Tottori 680-8552 (Japan)
2014-08-15
It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.
Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas
Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.
1980-01-01
The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.
Institute of Scientific and Technical Information of China (English)
Shi Bing-Ren; Qu Wen-Xiao
2006-01-01
A ballooning mode equation for tokamak plasma, with the toroidicity and the Shafranov shift effects included, is derived for a shift circular flux tokamak configuration. Using this equation, the stability of the plasma configuration with an internal transport barrier (IT2 against the high n (the toroidal mode number) ideal magnetohydrodynamic (MHD) ballooning mode is analysed. It is shown that both the toroidicity and the Shafranov shift effects are stabilizing.In the ITB region, these effects give rise to a low shear stable channel between the first and the second stability regions.Out of the ITB region towards the plasma edge, the stabilizing effect of the Shafranov shift causes the unstable zone to be significantly narrowed.
Plasma formation and sustainment by a multijunction grill on the CASTOR tokamak
International Nuclear Information System (INIS)
Radiofrequency power up to 40 kW, injected into the vacuum chamber of the CASTOR tokamak by a multijunction grill, was used for plasma production during the ramp-up phase of a toroidal magnetic field. When electron cyclotron resonance (ECR) appears inside the tokamak chamber for the given pumping frequency (f=1.25 GHz) plasma with a density greater than 2x1018 m-3 and a temperature of Te=10 to 40 eV is produced. The plasma is sustained at some lower value of density during the whole RF pulse. Simultaneously, a toroidal current of up to ≅ 0.2 kA is generated. The energy confinement time is estimated to be about 30 μs during the ECR breakdown. (author)
DIII-D integrated plasma control tools applied to next generation tokamaks
International Nuclear Information System (INIS)
A complete software suite for integrated tokamak plasma control has been developed within the DIII-D program. The suite consists of software for real-time control of all aspects of the plasma, modeling, simulation and design tools for analysis and development of controllers, a flexible and modular architecture for implementation and testing of algorithms and many fully validated models. Many elements of the system have been applied to and implemented on NSTX and MAST. The DIII-D realtime plasma control system together with the integrated modeling and simulation suite have been selected for operational use by both the KSTAR and EAST tokamaks, and are also being used at General Atomics to investigate control issues for ITER
Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers
International Nuclear Information System (INIS)
The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensional (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles
The trapping of a gun-injected plasma by a tokamak
International Nuclear Information System (INIS)
A Marshall gun was used to refuel a tokamak discharge on the Tokapole II device. Gun injection was able to increase the line-averaged density of the discharge by 50%. The density profile became more peaked due to gun injection. A model is discussed which describes the trapping of a gun injected plasma in a pure octupole field, due to a depolarization current. This model is expanded to include arbitrary toroidal fields added to the poloidal field. A slowing time, tau/sub s/, is derived for the trapping of an injected plasma of density n/sub b/, and temperature, T/sub e/, into poloidal field, B/sub p/ and toroidal field, B/sub t/. The experiment is extended to the tokamak discharge by the addition of plasma density and current to the vacuum fields, B/sub p/ and B/sub t/. Plasma density is seen to not significantly affect trapping. The increase in trapping with plasma current is explained in terms of additional poloidal field added to the central current channel. An extrapolation is made of the refueling system to the reactor-size tokamak TFTR. An effective system seems easily obtainable
Analysis of measurement errors for Thomson diagnostics of non-Maxwellian plasmas in tokamak reactors
Sdvizhenskii, P. A.; Kukushkin, A. B.; Kurskiev, G. S.; Mukhin, E. E.; Bassan, M.
2016-01-01
The study is stimulated by the expected noticeable deviation of the electron velocity distribution function (eVDF) from a Maxwellian under condition of a strong auxiliary heating of electron plasmas in tokamak-reactors. The key principles of accuracy estimation of the Thomson scattering diagnostic of non-Maxwellian plasmas in tokamak-reactors are presented. The algorithm extends the conventional approach to the assessment of non-Maxwellian plasmas measurements errors for a broad class of deviations of the eVDF from a Maxwellian. The algorithm is based on solving the inverse problem many times to determine main parameters of the eVDF with allowance for all possible sources of error and statistical variation of the input parameters of the problem. The method is applied to a preliminary analysis of the advantages of the formerly suggested use of various wavelengths of probing laser radiation in the Thomson diagnostics of non-Maxwellian plasma on the example of the core plasma Thomson scattering diagnostic system which is under design for ITER tokamak. The results obtained confirm the relevance of the diversification of the probing laser radiation wavelength.
FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak
Energy Technology Data Exchange (ETDEWEB)
Suratia, Pooja, E-mail: poojasuratia@yahoo.com [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Patel, Jigneshkumar, E-mail: jjp@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Rajpal, Rachana, E-mail: rachana@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Kotia, Sorum, E-mail: smkotia-eed@msubaroda.ac.in [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Govindarajan, J., E-mail: govindarajan@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India)
2012-11-15
Highlights: Black-Right-Pointing-Pointer Evaluation and comparison of the working performance of FLC is done with that of PID Controller. Black-Right-Pointing-Pointer FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. Black-Right-Pointing-Pointer FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. Black-Right-Pointing-Pointer Developed FLC controller is able to maintain the plasma column within required range of {+-}0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional-Integral-Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).
FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak
International Nuclear Information System (INIS)
Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).
Momentum balance and radial electric fields in axisymmetric and nonaxisymmetric toroidal plasmas
International Nuclear Information System (INIS)
It is investigated how symmetry properties of toroidal magnetic configurations influence mechanisms of determining the radial electric field such as the momentum balance and the ambipolar particle transport. Both neoclassical and anomalous transport of particles, heat, and momentum in axisymmetric and nonaxisymmetric toroidal systems are taken into account. Generally, in nonaxisymmetric systems, the radial electric field is determined by the neoclassical ambipolarity condition. For axisymmetric systems with up-down symmetry and quasisymmetric systems with stellarator symmetry, it is shown by using a novel parity transformation that the particle fluxes are automatically ambipolar up to O(δ2) and the determination of the radial electric field Es requires solving the O(δ3) momentum balance equations, where δ denotes the ratio of the thermal gyroradius to the characteristic equilibrium scale length. In axisymmetric systems with large E x B flows on the order of the ion thermal velocity vTi, the radial fluxes of particles, heat, and toroidal momentum are dependent on Es and its radial derivative while the time evolution of the Es profile is governed by the O(δ2) toroidal momentum balance equation. In nonaxisymmetric systems, E x B flows of O(vTi) are not generally allowed even in the presence of quasisymmetry because the nonzero radial current is produced by the large flow term in the equilibrium force balance for which the Boozer and Hamada coordinates cannot be constructed. (author)
Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment
Lucia, Matthew James
The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance
Identification of Plasma Boundary and Position for HL-2A Tokamak
Institute of Scientific and Technical Information of China (English)
Wang Zhongtian; Mao Guoping; Yang Qingwei; Zhang Jinghua; Gao Zhe; He Yexi
2005-01-01
Using the virtual-case principle, the plasma boundary, the plasma current center,and the x-point are identified for the HL-2A tokamak. The plasma current is represented by the current center and the virtual multipole moments which produce a magnetic flux in a form of polynomial. Adaptive parameters in the polynomial are determined by the least-square fit of the poloidal magnetic fields. The measurement of the magnetic field is performed using pick-up coils. The virtual-case principle is applied outside the plasma boundary. The virtual-case currents decide the position of the current center and produce a negative confinement magnetic field inside the plasma and the magnetic field generated by the plasma current outside the plasma boundary. The convergence is fast enough to get a picture between the sequent shots. The configuration reconstructed is in good agreement with the TV image taken by camera with a tangential view.
International Nuclear Information System (INIS)
Surface and structural damage to plasma-facing components (PFCs) due to the frequent loss of plasma confinement remains a serious problem for the tokamak reactor concept. The deposited plasma energy causes significant surface erosion, possible structural failure, and frequent plasma contamination. Surface damage consists of vaporization, spallation, and liquid splatter of metallic materials. Structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. To evaluate the lifetimes of plasma-facing materials and nearby components and to predict the various forms of damage that they experience, comprehensive models (contained in the HEIGHTS computer simulation package) are developed, integrated self-consistently, and enhanced. Splashing mechanisms such as bubble boiling and various liquid magnetohydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials are being examined. The design requirements and implications of plasma-facing and nearby components are discussed, along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components
International Nuclear Information System (INIS)
A multi-functional fusion test reactor concept named FDS-MFX (multi- functional engineering experimental reactor) proposed as a scenario option of China Fusion Engineering Test Reactor (CFETR) has been presented by FDS Team. FDS- MFX has been proposed for checking and validating the fusion DEMO reactor relevant technologies based on viable technologies. The preferred fusion core of FDS-MFX is regular Tokamak, with alternative choices such as spherical Tokamak and magnetic mirror, etc. In this paper, the core plasma parameters of FDS-MFX based on regular Tokamak were designed with the independently developed fusion system optimization and economic analysis code SYSCODE and analyzed based on the 'ITER Physics Basis'. We simulated the plasma equilibrium configuration and plasma discharge using the Tokamak Simulation Code (TSC); the result showed the core plasma parameters of FDS-MFX were preliminarily feasible. (authors)
International Nuclear Information System (INIS)
Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (ne ∼ 1 - 5 x 1014 (cm-3), Te ∼ 4 - 10 (eV), Bt ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of CII(4267A, emitted by the C+ ions) and CIII (4647A, emitted by the C++ ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ions is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is ''classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the Hα emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time τp is comparable with the plasma energy confinement time τE; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy
Characteristics of disruptive plasma current decay in the HT-2 tokamak
Energy Technology Data Exchange (ETDEWEB)
Abe, Mitsushi; Takeuchi, Kazuhiro; Otsuka, Michio (Hitachi Ltd., Ibaraki (Japan). Energy Research Lab.)
1993-04-01
Motions of plasma current channel and time evolutions of eddy current distribution on the vacuum vessel during disruptive plasma current decay were studied experimentally in the Hitachi tokamak HT-2. The plasmas are vertically elongated and circularly shaped plasmas. A disruptive plasma current decay has three phases. During the first phase, a large displacement of the plasma position without plasma current decay is observed. Rapid plasma current decay is observed during the second phase and the decay rate is roughly constant with time. The eddy current distribution is like that due to the shell effect which creates a poloidal field to reduce the plasma displacement. During the third phase, the plasma current decays exponentially. The second phase is observed in slightly elongated and high plasma current (> 20 kA) circularly shaped plasmas. The plasma current decay rates in the second phase depend on the plasma cross sectional shape, but they do not in the third phase. The magnetic axis moves from the plasma area to the vacuum vessel wall between the second and third phases. (author).
Axisymmetric Control in Alcator C-Mod
Tinios, Gerasimos
1995-01-01
This thesis investigates the degree to which linear axisymmetric modeling of the response of a tokamak plasma can reproduce observed experimental behavior. The emphasis is on the vertical instability. The motivation for this work lies in the fact that, once dependable models have been developed, modern control theory methods can be used to design feedback laws for more effective and efficient tokamak control. The models are tested against experimental data from the Alcator C-Mod tokamak. A linear model for each subsystem of the closed-loop system constituting an Alcator C-Mod discharge under feedback control has been constructed. A non-rigid, approximately flux-conserving, perturbed equilibrium plasma response model is used in the comparison to experiment. A detailed toroidally symmetric model of the vacuum vessel and the supporting superstructure is used. Modeling of the power supplies feeding the active coils has been included. Experiments have been conducted with vertically unstable plasmas where the feedback was turned off and the plasma response was observed in an open -loop configuration. The closed-loop behavior has been examined by injecting step perturbations into the desired vertical position of the plasma. The agreement between theory and experiment in the open-loop configuration was very satisfactory, proving that the perturbed equilibrium plasma response model and a toroidally symmetric electromagnetic model of the vacuum vessel and the structure can be trusted for the purpose of calculations for control law design. When the power supplies and the feedback computer hardware are added to the system, however, as they are in the closed-loop configuration, they introduce nonlinearities that make it difficult to explain observed behavior with linear theory. Nonlinear simulation of the time evolution of the closed-loop experiments was able to account for the discrepancies between linear theory and experiment. (Copies available exclusively from MIT Libraries
Measurements and modelling of plasma response field to RMP on the COMPASS tokamak
Markovic, T.; Liu, Y. Q.; Cahyna, P.; Pánek, R.; Peterka, M.; Aftanas, M.; Bílková, P.; Bohm, P.; Imríšek, M.; Háček, P.; Havlicek, J.; Havránek, A.; Komm, M.; Urban, J.; Weinzettl, V.; the COMPASS Team
2016-09-01
It has been shown on several tokamaks that application of a resonant magnetic perturbation (RMP) field to the plasma can lead to suppression or mitigation of edge-localized mode (ELM) instabilities. Due to the rotation of the plasma in the RMP field reference system, currents are induced on resonant surfaces within the plasma, consequently screening the original perturbation. In this work, the extensive set of 104 saddle loops installed on the COMPASS tokamak is utilized to measure the plasma response field for two n = 2 RMP configurations of different poloidal mode m spectra. It is shown that spatially the response field is in opposite phase to the original perturbation, and that the poloidal profile of the measured response field does not depend on the poloidal profile of the applied RMP. Simulations of the plasma response by the linear MHD code MARS-F (Liu et al 2000 Phys. Plasmas 7 3681) reveal that both of the studied RMP configurations are well screened by the plasma. Comparison of measured plasma response field with the simulated one shows a good agreement across the majority of poloidal angles, with the exception of the midplane low-field side area, where discrepancy is seen.
Shaing, K. C.; Sabbagh, S. A.
2016-07-01
Theory for neoclassical toroidal plasma viscosity has been developed to model transport phenomena, especially, toroidal plasma rotation for tokamaks with broken symmetry. Theoretical predictions are in agreement with the results of the numerical codes in the large aspect ratio limit. The theory has since been extended to include effects of finite aspect ratio and finite plasma β. Here, β is the ratio of the plasma thermal pressure to the magnetic field pressure. However, there are cases where the radial wavelength of the self-consistent perturbed magnetic field strength B on the perturbed magnetic surface is comparable to the width of the trapped particles, i.e., bananas. To accommodate those cases, the theory for neoclassical toroidal plasma viscosity is further extended here to include the effects of the finite banana width. The extended theory is developed using the orbit averaged drift kinetic equation in the low collisionality regimes. The results of the theory can now be used to model plasma transport, including toroidal plasma rotation, in real finite aspect ratio, and finite plasma β tokamaks with the radial wavelength of the perturbed symmetry breaking magnetic field strength comparable to or longer than the banana width.
Numerical Investigations of Plasma Parameters in COMPASS Tokamak
Czech Academy of Sciences Publication Activity Database
Havlíčková, E.; Zagórski, R.; Pánek, Radomír
Praha: Institute of Plasma Physics AS CR,v.v.i, 2007 - (Brotánková, J.). s. 19-19 [Workshop on Electric Fields, Structures, and Relaxation in Plasmas/10th./. 8.7.2007-9.7.2007, Varšava] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma modeling * COMPASS * detachement Subject RIV: BL - Plasma and Gas Discharge Physics
Fast potential change in sawteeth in JIPP T-IIU tokamak plasmas
International Nuclear Information System (INIS)
Fast changes of electric potential with different polarities are observed during sawtooth oscillation in a core region of a tokamak plasma using a heavy ion beam probe. The potential change inside the inversion radius is found to be positive. The change is negative outside the inversion radius and shows clearly a propagation nature. The observed potential can be interpreted by the mixture of the potentials of two origins. One of them drives the fast MHD plasma motion through E/B drift and the other is a barrier potential induced by mixing of hot and cold plasmas at sawtooth crash. (author)
Growth rate of peeling mode in the near separatrix region of diverted tokamak plasma
International Nuclear Information System (INIS)
An analytical expression of the peeling mode in the near separatrix region of diverted tokamak plasma is derived. It is shown that in diverted plasmas both with single and double X points, though the perturbed potential energy of the unstable peeling mode tends to be large, its growth rate becomes very small due to the even larger kinetic energy. Compared to some recent studies that give qualitatively correct results about this growth rate, our result is directly related with the diverted equilibrium quantities suitable for application to realistic experiments. (physics of gases, plasmas, and electric discharges)
Second-harmonic ion cyclotron resonance heating scenarios of Aditya tokamak plasma
Indian Academy of Sciences (India)
Asim Kumar Chattopadhyay; S V Kulkarni; R Srinivasan; Aditya Team
2015-10-01
Plasma heating with the fast magnetosonic waves in the ion cyclotron range of frequencies (ICRF) is one of the auxiliary heating schemes of Aditya tokamak. Numerical simulation of second-harmonic resonance heating scenarios in low-temperature, low-density Aditya plasma has been carried out for fast magnetosonic wave absorption in ICRF range, using full-wave ion cyclotron heating code TORIC combined with Fokker–Planck quasilinear solver SSFPQL and the results are explained. In such low-temperature, low-density plasma, ion absorption for second-harmonic resonance heating is less but significant amount of direct electron heating is observed.
Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Berta, Miklós; Anda, G.; Bencze, A.; Dunai, D.; Háček, Pavel; Hron, Martin; Kovácsik, A.; Krbec, Jaroslav; Pánek, Radomír; Réfy, D.; Véres, G.; Weinzettl, Vladimír; Zoletnik, S.
96-97, October (2015), s. 795-798. ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : BES * plasma diagnostics * COMPASS tokamak * density fluctuations * plasma density profile Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://www.sciencedirect.com/science/article/pii/S0920379615300442
Multi-machine scaling of main SOL parallel heat flux width in tokamak limiter plasmas
Czech Academy of Sciences Publication Activity Database
Horáček, Jan; Pitts, R.A.; Gunn, J.; Silva, C.; Rudakov, D.; Arnoux, G.; Marsen, S.; Vondráček, Petr; Maddaluno, G.; Pericoli, V.; Viola, B.; Xu, G.S.; Wang, H.; Nie, L.; LaBombard, B.; Brezinsek, S.; Xu, Y.; Shimada, M.; Adámek, Jiří; Popov, Tsv.; Dimitrova, Miglena; Seidl, Jakub; Janky, Filip; Pánek, Radomír; Goldston, R.J.; Stangeby, P.C.
Vol. 39E. Mulhouse : European Physical Society, 2015, O2.105-O2.105. ISBN 2-914771-98-3. - (Europhysics Conference Abstracts. ECA. 39E). [EPS 2015 Conference on Plasma Physics/42./. Lisabon (PT), 22.06.2015-26.06.2015] R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 Institutional support: RVO:61389021 Keywords : tokamak * plasma * SOL * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://ocs.ciemat.es/EPS2015PAP/pdf/O2.105.pdf
Effects of Plasma Control on Runaway Electrons in the COMPASS Tokamak
Czech Academy of Sciences Publication Activity Database
Mlynář, Jan; Ficker, Ondřej; Vlainić, M.; Weinzettl, Vladimír; Imríšek, Martin; Papřok, Richard; Rabinski, M.; Jakubowski, M.; Tomeš, Matěj; Peterka, Matěj; Pánek, Radomír
Vol. 39E. Mulhouse: European Physical Society, 2015, P4.102-P4.102. (Europhysics Conference Abstracts. ECA. 39E). ISBN 2-914771-98-3. [EPS 2015 Conference on Plasma Physics/42./. Lisabon (PT), 22.06.2015-26.06.2015] R&D Projects: GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 Institutional support: RVO:61389021 Keywords : plasma * tokamak * control runaway electrons Subject RIV: BL - Plasma and Gas Discharge Physics http://ocs.ciemat.es/EPS2015PAP/pdf/P4.102.pdf
Real-time software for the COMPASS tokamak plasma control
International Nuclear Information System (INIS)
The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 μs. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.
Real-time software for the COMPASS tokamak plasma control
Energy Technology Data Exchange (ETDEWEB)
Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)
2010-07-15
The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.
42GHz ECRH assisted Plasma Breakdown in tokamak SST-1
Directory of Open Access Journals (Sweden)
Shukla B. K.
2015-01-01
Full Text Available In SST-1, 42GHz ECRH system has been commissioned to carry out breakdown and heating experiments at 0.75T and 1.5T operating toroidal magnetic fields. The 42GHz ECRH system consists of high power microwave source Gyrotron capable to deliver 500kW microwave power for 500ms duration, approximately 20 meter long transmission line and a mirror based launcher. The ECRH power in fundamental O-mode & second harmonic X-mode is launched from low field side (radial port of the tokamak. At 0.75T operation, approximately 300 kW ECH power is launched in second harmonic X-mode and successful ECRH assisted breakdown is achieved at low loop_voltage ~ 3V. The ECRH power is launched around 45ms prior to loop voltage. The hydrogen pressure in tokamak is maintained ~ 1×10-5mbar and the pre-ionized density is ~ 4×1012/cc. At 1.5T operating toroidal magnetic field, the ECH power is launched in fundamental O-mode. The ECH power at fundamental harmonic is varied from 100 kW to 250 kW and successful breakdown is achieved in all ECRH shots. In fundamental harmonic there is no delay in breakdown while at second harmonic ~ 40ms delay is observed, which is normal in case of second harmonic ECRH assisted breakdown.
International Nuclear Information System (INIS)
In the non-circular tokamak TNT-A, we have tried to elongate plasma column by the combined operation of rapid rise of plasma current and rapid decrease in the decay index n (=-(R/Bsub(z)).(deltaBsub(z)/deltaR)) with gas puffing after the current establishment. The increase in elongation ratio is obtained without deleterious effects on MHD instabilities. Dependence of elongation ratio on plasma parameters is calculated and compared with measurements. The growth rate of positional instabilities in the vertical direction is measured as a function of the decay index, and has good agreement with a calculation. (author)
International Nuclear Information System (INIS)
In view of realising the full potential of fusion as an abundant energy source, some challenges must still be solved. They are identified and will be addressed by the implementation of the EU fusion roadmap. The TCV tokamak, with its high plasma shaping capability and the flexibility of its heating and current drive systems, is strongly contributing to this effort, as one of a small number of devices selected by the EU community for the 2014-2020 period. One of the primary challenges lies in the heat exhaust from tokamak plasmas. Indeed, the currently foreseen operational regime of ITER implies heat flows impinging onto the facing materials that are not compatible with a fully operating fusion reactor. TCV has developed alternative plasma configurations, termed 'snowflakes', that strongly reduce the heat flow towards the vessel walls, via an increase in the number of deposition surface areas, as shown in Fig. 1. Measurement of particle fluxes, together with IR camera imaging, show a clear reduction of the heat flow onto the walls. The TCV tokamak is going through major upgrades of its heating systems to expand its operational domain towards burning plasma regimes. The installation of a 1MW neutral beam injector will allow the achievement of high temperature plasmas with equal ion and electron temperatures. An additional 2MW of electron cyclotron resonance heating power will be installed to increase the plasma pressure near the range in which ITER will operate. This will also improve access to and control of high confinement regimes. Varying the power ratio between the two heating systems will furthermore lead to improved understanding of the different plasma turbulence regimes that develop in plasmas with different electron to ion temperature ratios. Acknowledgement: This work was partly supported by the Swiss National Science Foundation. (author)
Real-time control of current and pressure profiles in tokamak plasmas
International Nuclear Information System (INIS)
Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)
Soft x-ray imaging system for measurement of noncircular tokamak plasmas
International Nuclear Information System (INIS)
A soft x-ray camera and image processing system has been constructed to provide measurements of the internal shape of high temperature tokamak plasmas. The camera consists of a metallic-foil-filtered pinhole aperture and a microchannel plate image intensifier/convertor which produces a visible image for detection by a CCD TV camera. A wide-angle tangential view of the toroidal plasma allows a single compact camera to view the entire plasma cross section. With Be filters 12 to 50 μm thick, the signal from the microchannel plate is produced mostly by nickel L-line emissions which orignate in the hot plasma core. The measured toroidal image is numerically inverted to produce a cross-sectional soft x-ray image of the plasma. Since the internal magnetic flux surfaces are usually isothermal and the nickel emissivity depends strongly on the local electron temperature, the x-ray emission contours reflect the shape of the magnetic surfaces in the plasma interior. Initial results from the PBX tokamak experiment show clear differences in internal plasma shapes for circular and bean-shaped discharges
Emission Lines of Boron, Carbon, Oxygen and Iron in Tokamak Plasma
Institute of Scientific and Technical Information of China (English)
DI Long; WAN Bao-Nian; ZHAO Gang; ZHANG Jie; SHI Jian-Rong; WANG Shou-Jun; DONG Quan-Li; ZHAO Jing; LI Yu-Tong; FU Jia; WANG Fu-Di; SHI Yue-Jiang
2011-01-01
The emission lines of B,C,O and Fe in tokamak plasma are reported. The spectra are compared with those calculated by the CHIANTI code,which is based on the collisional-radiative models with a large amount of accurate atomic data.General agreement is obtained between the results of experiment and computation.Most of the lines in the spectra are identified,and the relative number density ratios orB,C,O and Fe are determined.It is found that the processes of line formation in our experiment are similar to those in the stellar coronae.The line-averaged electron density of the tokamak plasma is measured by the HCN laser,indicating a good agreement with the theoretical prediction by the density-dependent line ratio of Fe XXI.
Profile control of advanced tokamak plasmas in view of continuous operation
Energy Technology Data Exchange (ETDEWEB)
Mazon, D., E-mail: Didier.Mazon@cea.fr
2015-07-15
The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named ‘advanced scenarios’ are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated ‘bootstrap’ current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.
Profile control of advanced tokamak plasmas in view of continuous operation
Mazon, D.
2015-07-01
The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named 'advanced scenarios' are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated 'bootstrap' current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.
GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography
Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.
2015-09-01
An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.
Experimental investigation of turbulent transport at the edge of a tokamak plasma
International Nuclear Information System (INIS)
This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)
Profile control of advanced tokamak plasmas in view of continuous operation
International Nuclear Information System (INIS)
The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named ‘advanced scenarios’ are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated ‘bootstrap’ current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described
Czech Academy of Sciences Publication Activity Database
Cavalier, Jordan; Weinzettl, Vladimír; Varju, Jozef; Pánek, Radomír
Prague: Czech Technical University in Prague, Faculty of Electrical Engineering, 2014. s. 21. [SPPT 2014 - 26th Symposium on Plasma Physics and Technology/26./. 16.06.2014-19.06.2014, Prague] Keywords : Tokamak edge plasma * gas-puff imaging * diagnostic Subject RIV: BL - Plasma and Gas Discharge Physics
Resistive MHD studies of high-β-tokamak plasmas
International Nuclear Information System (INIS)
Numerical calculations have been performed to study the MHD activity in high-β tokamaks such as ISX-B. These initial value calculations built on earlier low β techniques, but the β effects create several new numerical issues. These issues are discussed and resolved. In addition to time-stepping modules, our system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes at low β to predominantly pressure driven modes at high β is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment
Tokamak plasma self-organization and possibility to have the peaked density profile in ITER
International Nuclear Information System (INIS)
The extrapolation from to-day tokamaks to ITER dimension hitherto remain to be very far. We can't do such extrapolation with the help of scaling built without deep understanding of the physical processes. The main process in the turbulent plasma is its self-organization. Transport coefficients may be changed in plasma in a wide range by changing of the instabilities level. This permit plasma to build the most stable under given external impacts energy distribution profile, i.e., plasma pressure self-consistent profile. This process cannot be realized only in the regions, where the turbulent transport value is not enough for the p(r) regulation. The enhanced grad p appears here. We call these regions 'transport barriers'. The plasma pressure self-consistent profile and transport barriers are two main factors, which determine the plasma pressure profile. At T-10 tokamak the phenomenon of plasma pressure self-consistent profile was investigated and it was shown that normalized plasma pressure does not depend on any plasma parameters if we normalize the minor radius to the radius rc = (IR/kB)1/2, ρ r/rc, k is plasma elongation coefficient, I plasma current, R major radius and B magnetic field strength. In this case p(ρ)/p(0) practically the same for T-10 and JET if in given shot there was not pronounced ITB. The process of p(ρ) restoration very fast and links to plasma minor radius equilibrium. Now we can predict the self-consistent pressure profile for ITER, which can be changed by ITB formation only. As we expect in ITER the peaked power deposition profile, and so the peaked temperature profile, we can conclude that density profile will be flat even in the low-collisionality case. (author)
Effect of plasma inertia on vertical displacement instability in tokamaks
International Nuclear Information System (INIS)
The effect of plasma inertia on vertical displacement instability (VDI) is investigated in both linear and nonlinear regimes. In a linear case, the solution of the dispersion shows the existence of a critical elongation for a certain distance of the conducting wall. In the nonlinear case a difference between the cases with and without inclusion of plasma inertia is found. The inclusion of the plasma inertia leads to a 'softening' of the critical behavior of the displacement against the distance of the resistive wall. Without inclusion of plasma inertia there exists a critical distance between the wall and the plasma below which VDI can not be stabilized no matter how the parameters in feedback controlling scheme are set. An analysis of the dispersion of linear VDI shows that the plasma inertia plays an important role in the properties of the solutions of dispersion and hence in the behavior of VDI. (author)
Conceptual design of plasma position control of SST-1 tokamak using vertical field coil
International Nuclear Information System (INIS)
SST-1 (Steady State Superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. Based on the solution of Grad-Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec. The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span. In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results. (author)
Identification of krypton Kr XVIII to Kr XXIX spectra excited in TFR Tokamak plasmas
International Nuclear Information System (INIS)
The emission spectrum of krypton (injected into TFR tokamak plasmas) has been recorded photographically in the 15-300 A spectral range by means of a 2m grazing incidence spectrograph. Preliminary identification work, based on isoelectronic regularities from known spectra of other ions and ionization equilibrium calculations, has allowed 48 lines (belonging to the O I, F I, Na I, Mg I, Al I, Ar I and K I sequences) to be identified
Self-organized criticality processes in HL-1M tokamak plasma
International Nuclear Information System (INIS)
We study the dynamics of laminar time between successive bursts in anomalous particle flux measured at the HL-1M tokamak plasma edge. The results reveal that the flux fluctuations are self-similar in a narrow range of time scales and that their probability distribution function is not Gaussian. These properties are not consistent with those predicted by self-organized criticality (SOC) models as well as the running sand-pile SOC model developed by Hwa and Kardar
Stabilization of external modes in tokamaks by resistive walls and plasma rotation
International Nuclear Information System (INIS)
It is shown that low-n pressure driven external modes in tokamaks can be fully stabilized by resistive walls in combination with sonic rotation of the plasma. The stabilization depends on the excitation of sound waves by the toroidal coupling to Alfven waves and is affected by ion Landau damping. Two-dimensional stability calculations are presented to show the gains in the beta limit resulting from this wall stabilization. (author) 4 figs., 13 refs
Numerical treatment of the problem of impurity transport in tokamak plasma
International Nuclear Information System (INIS)
The physical and mathematical aspects of the problem of impurity transport in tokamak plasma are examined in view of the computer simulation of this process. Regarding the system of equations, some new conclusions are revealed. Then, the system is treated by three different numerical methods, and correspondingly three codes have been written. Detailed description of the methods and various numerical tests are presented as suggestions for the construction of a code for impurity evolution. (authors)
3-D resistive MHD calculations for tokamak plasmas: beyond the simple reduced set of equations
International Nuclear Information System (INIS)
Numerical studies of the resistive stability of tokamak plasmas in cylindrical geometry have been performed using: (1) the full set of resistive Magnetohydrodynamic (MHD) equations and (2) an extended version of the reduced set of resistive MHD equations including diamagnetic and electron temperature effects. In particular, the nonlinear interaction of tearing modes of many helicities has been investigated. The numerical results confirm many of the features uncovered previously using the simple reduced equations. (author)
ELMy-H mode as limit cycle and chaotic oscillations in tokamak plasmas
International Nuclear Information System (INIS)
A model of Edge Localized Modes (ELMs) in tokamaks is presented. A limit cycle solution is found in time-dependent Ginzburg Landau type model equation of L/H transition, which has a hysteresis curve between the plasma gradient and flux. The oscillation of edge density appears near the L/H transition boundary. Spatial structure of the intermediate state (mesophase) is obtained in the edge region. Chaotic oscillation is predicted due to random neutrals and external oscillations. (author)
3D MHD simulations of pellet injection and disruptions in tokamak plasmas
International Nuclear Information System (INIS)
Nonlinear MHD simulation results of pellet injection show that MHD forces can accelerate large pellets, injected on the high field side of a tokamak, to the plasma center. Magnetic reconnection can produce a reverse shear q profile. Ballooning instability caused by pellets is also reduced by high field side injection. Studies are also reported of the current quench phase of disruptions, which can cause 3D halo currents and runaway electrons. (author)
Energy Technology Data Exchange (ETDEWEB)
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others
2001-01-10
The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] 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 centimeters 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 coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing 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/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling 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 [Research and Development] avenues for their resolution are presented.
The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR
Energy Technology Data Exchange (ETDEWEB)
Tammen, H.F.
1995-01-10
One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).
International Nuclear Information System (INIS)
The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) 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 centimeters 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 coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing 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/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling 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 (Research and Development) avenues for their resolution are presented
The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR
International Nuclear Information System (INIS)
One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)
Development of a FE method for modelling plasma flows in tokamak plasma edges
International Nuclear Information System (INIS)
The purpose of the work is to to develop a two-dimensional Finite-Element-Code. This code should be able to simulate the plasma flow pattern in the burning chamber of fusion devices by an exact and solution-dependent discretisation. Reionisation and other collision processes of recycled neutral particles are described by coupling the fluid model to the kinetic Monte-Carlo neutral-gas-code EIRENE. For comparison and fundamental numerical studies a fast analytical description of recycling is also available. Such rather crude approximations are employed in other codes often as the only option. It is possible to treat the flow of ions and neutral atoms/molecules near complex surface structures of fusion devices consistently. Because of the time step restriction in the special solution algorithm, up to now the electron temperature profile has to be provided from elsewhere. It can, for example be interpolated from experimental data or from results of other independent code calculations. The newly developed code is applied to a typical TOKAMAK-discharge (TEXTOR) and characteristic results are discussed. (orig./HP)
Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma
International Nuclear Information System (INIS)
In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)
First Results of Movable Limiter Experiments and its Effects on the IR-T1 Tokamak Plasma Confinement
International Nuclear Information System (INIS)
Tokamak limiter plays a number of roles in its operation. It serves primarily to protect the wall from the plasma when there are disruptions, runaway electrons, or other instabilities. For this reason it is commonly made of a refractory material, such as carbon, molybdenum or tungsten, capable of withstanding high heat loads. Secondly, the limiter localizes the plasma-surface interaction. The high power and particle density at the limiter surface causes rapid removal of adsorbed gas, oxide layers and other easily desorbed impurities. When only the clean substrate remains it is possible to maintain plasmas with lower impurity levels. Plasma-surface interaction in tokamaks is important in plasma conditions such as impurities, plasma parameters, and plasma confinement. Thirdly, the limiter localizes the particle recycling. A higher neutral density and more radiation are observed in the region near the limiter, than at other positions around the torus . Movable limiter was designed, constructed, and has been used on IR-T1 tokamak to investigate the possibility of modifying the plasma confinement. In this research work we present the first results of a movable limiter experiments and its effects on the tokamak plasma confinement. For this purpose we firstly designed, constructed, and installed a movable localized poloidal limiter, and then measured the effects of limiter position on the time intervals of plasma parameters such as: plasma density, temperature, and energy confinement time. The results compared in different movable limiter positions and discussed. (author)
Plasma Potential Measurements with Emissive Probes in the CASTOR Tokamak
Czech Academy of Sciences Publication Activity Database
Schrittweiser, R.; Adámek, Jiří; Balan, P.; Hron, Martin; Ionita, C.; Jakubka, Karel; Kryška, Ladislav; Martines, E.; Pohoata, V.; Stöckel, Jan; Tichý, M.; Van Oost, G.
volume 25A. Mulhouse: European Physical Society, 2001 - (Silva, C.; Varandas, C.; Campbell, D.), s. 409-412. (Europhysics Conference Abstracts.. 25A). [European Physical Society Conference on Controlled Fusion and Plasma Physics/28th./. Funchal, Madeira (PT), 18.06.2001-22.06.2001] Institutional research plan: CEZ:AV0Z2043910 Subject RIV: BL - Plasma and Gas Discharge Physics
Deuterium--tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor
International Nuclear Information System (INIS)
Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 2, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium - tritium (D--T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high li). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (qa∼4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-li plasmas produced by rapid expansion of the minor cross section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D--T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D--T plasmas with q0>1 and weak magnetic shear in the central region, a toroidal Alfvn eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions. copyright 1997 American Institute of Physics
International Nuclear Information System (INIS)
This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97–104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD
Rapid further heating of tokamak plasma by fast-rising magnetic pulse
International Nuclear Information System (INIS)
The object of the experiment was to study the rapid further heating of a tokamak plasma and its influence on confinement. For this purpose, a high-voltage theta-pinch pulse was applied to a tokamak plasma and production of a high-temperature (keV) plasma was ensured within a microsecond. The magnetic pulse is applied at the plasma current maximum parallel or antiparallel to the study toroidal field. In either case, the pulsed field quickly penetrates the plasma and the plasma resistivity estimated from the penetration time is about 100 times larger than the classical. A burst of energetic neutrals of approximately 1 μs duration was observed and the energy distribution had two components of the order of 1 keV and 0.1 keV in the antiparallel case. Doppler broadening measurement shows heating of ions to a temperature higher than 200 eV; however, the line profile is not always Maxwellian distribution. The X-rays disappear at the moment of applying the magnetic pulse and reappear about 100 μs later with an intensive burst, while both energy levels are the same (approximately 100 keV). (author)
Tokamak Plasmas : Electron temperature $(T_{e})$ measurements by Thomson scattering system
Indian Academy of Sciences (India)
R Rajesh; B Ramesh Kumar; S K Varshney; Manoj Kumar; Chhaya Chavda; Aruna Thakkar; N C Patel; Ajai Kumar; Aditya Team
2000-11-01
Thomson scattering technique based on high power laser has already proved its superoirity in measuring the electron temperature (e) and density (e) in fusion plasma devices like tokamaks. The method is a direct and unambiguous one, widely used for the localised and simultaneous measurements of the above parameters. In Thomson scattering experiment, the light scattered by the plasma electrons is used for the measurements. The plasma electron temperature is measured from the Doppler shifted scattered spectrum and density from the total scattered intensity. A single point Thomson scattering system involving a -switched ruby laser and PMTs as the detector is deployed in ADITYA tokamak to give the plasma electron parameters. The system is capable of providing the parameters e from 30 eV to 1 keV and e from 5 × 1012 cm-3-5× 1013 cm-3. The system is also able to give the parameter proﬁle from the plasma center ( = 0 cm) to a vertical position of = +22 cm to = -14 cm, with a spatial resolution of 1 cm on shot to shot basis. This paper discusses the initial measurements of the plasma temperature from ADITYA.
Study of plasma rotation in Tokamak confinement. Progress report, June 1, 1997--February 28, 1998
International Nuclear Information System (INIS)
This proposal is a collaborative project between Prairie View A ampersand M University and Princeton University. The proposed tasks have been established based on close collaboration between two institutions. We studied the tasks in two aspects: analytical theory of drift current in tokamak plasmas, and computer simulation of non-neutral plasma. Some preliminary results have been presented in the 1997 APS Division of Plasma Physics Meeting, Pittsburgh. Titles of the presentations were open-quotes Magnetic Moment of Bounce Motion in Tokamak Plasmaclose quotes and open-quotes Numerical Simulation of Plasma Confinement in a Non-neutral Plasmaclose quotes. The papers for publication are in preparation. In the coming year, we will further develop the analytic theory and simulation studies. The studies will be focused on understanding of edge electric field in TFTR experiments, and attention will be paid to the effect of the a particles resulting from DT fusion reactions. In addition, in order to establish a stronger fusion plasma research and education base at Prairie View A ampersand M University, we plan to expand our current theoretical project into a coupled theoretical/experimental project. With the help of Oak Ridge National Lab we plan to build a plasma physics laboratory equipped with a small mirror machine. For the second year budget of this proposal, including the funds for the previously proposed theoretical and the newly planned experimental tasks, we request a $245,000 grant. A budget plan and its justification are included in this report
Measurement of the effective plasma ion mass in large tokamaks using Global Alfven Eigenmodes
International Nuclear Information System (INIS)
The ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not depend only on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. In view of the lack of competitive diagnostics, this method has been revisited to assess its likely precision for the JET tokamak. Our results show that the low-n GAE modes are sometimes too close to the continuum edge to be detectable and that the interpretation of the GAE spectrum is therefore rendered less direct than had been hoped. However, information on the effective plasma ion mass is still available in the detectable modes and we present a statistical study on the precision with which this quantity could be estimated from the GAE spectrum on JET, including other directly measured or simply available plasma parameters. (author) 5 figs., 3 tabs., 10 refs
Determination of the plasma column shape in the Tokamak Novillo cross section by magnetic probes
International Nuclear Information System (INIS)
The determination of plasma cross section shape in Tokamaks is an important diagnostic method for equilibrium conditions analysis. In this work, it is obtained a time dependent variation of the plasma column cross section in Novillo Tokamak. The experimental method is based on using one magnetic probe, which is installed inside of the vacuum vessel in a 1 mm. wall thickness stainless steel tube, in the protected region of the limiter shadow. The plasma column cross section is determined measuring the poloidal magnetic field produced by the plasma current. This method, now running for determining the plasma column shape, requires the measurement of magnetic present field outside plasma column. The measurements are carried out from a set of small coils, which are located inside the vacuum chamber in the radial and poloidal direction, so we can measure magnetic field with no current attenuations produced by the penetration time of the stainless steel vacuum chamber. The magnetic probe detect a real time variation of magnetic flux passing through them. In order to obtain the magnetic field values, it is required that the electric signals coming from the magnetic probe be integrated, this operation is carried out by active circuits located between the probe signal and one oscilloscope. The integrated signals can be exhibited photographed on the oscilloscope display. (Author)
Zakharov, Leonid E.; Li, Xujing
2015-06-01
This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97-104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.
Energy Technology Data Exchange (ETDEWEB)
Zakharov, Leonid E. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Li, Xujing [Institute of Computational Mathematics and Scientific/Engineering Computing, Academy of Mathematics and Systems Science, Chinese Academy of Sciences, P.O. Box 2719, Beijing 100190 (China)
2015-06-15
This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97–104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.
Full orbit computations of ripple-induced fusion α-particle losses from burning tokamak plasmas
International Nuclear Information System (INIS)
A full orbit code is used to compute collisionless losses of fusion α particles from three proposed burning plasma tokamaks: the International Tokamak Experimental Reactor (ITER); a spherical tokamak power plant (STPP) [T. C. Hender, A. Bond, J. Edwards, P. J. Karditsas, K. G. McClements, J. Mustoe, D. V. Sherwood, G. M. Voss, and H. R. Wilson, Fusion Eng. Des. 48, 255 (2000)]; and a spherical tokamak components test facility (CTF) [H. R. Wilson, G. M. Voss, R. J. Akers, L. Appel, A. Dnestrovskij, O. Keating, T. C. Hender, M. J. Hole, G. Huysmans, A. Kirk, P. J. Knight, M. Loughlin, K. G. McClements, M. R. O'Brien, and D. Yu. Sychugov, Proceedings of the 20th IAEA Fusion Energy Conference, Invited Paper FT/3-1Ra]. It has been suggested that α particle transport could be enhanced due to cyclotron resonance with the toroidal magnetic field ripple. However, calculations for inductive operation in ITER yield a loss rate that appears to be broadly consistent with the predictions of guiding center theory, falling monotonically as the number of toroidal field coils N is increased (and hence the ripple amplitude is decreased). For STPP and CTF the loss rate does not decrease monotonically with N, but collisionless losses are generally low in absolute terms. As in the case of ITER, there is no evidence that finite Larmor radius effects would seriously degrade fusion α-particle confinement
Di Troia, Claudio
2015-01-01
A class of parametric distribution functions has been proposed in [C.DiTroia, Plasma Physics and Controlled Fusion,54,2012] as equilibrium distribution functions (EDFs) for charged particles in fusion plasmas, representing supra-thermal particles in anisotropic equilibria for Neutral Beam Injection, Ion Cyclotron Heating scenarios. Moreover, the EDFs can also represent nearly isotropic equilibria for Slowing-Down $alpha$ particles and core thermal plasma populations. These EDFs depend on constants of motion (COMs). Assuming an axisymmetric system with no equilibrium electric field, the EDF depends on the toroidal canonical momentum $P_\\phi$, the kinetic energy $w$ and the magnetic moment \\mu. In the present work, the EDFs are obtained from first principles and general hypothesis. The derivation is probabilistic and makes use of the Bayes' Theorem. The bayesian argument allows us to describe how far from the prior probability distribution function (pdf), e.g. Maxwellian, the plasma is, based on the information...
Computation of electromagnetic effects in a tokamak due to a plasma disruption
International Nuclear Information System (INIS)
To model the consequences of a plasma disruption in a tokamak one must combine a code that computes the detailed MHD behavior of the plasma with one that treats the three-dimensional features of the conducting toroidal components around the plasma. The NET (Next European Torus) Team have undertaken a treatment of electromagnetic effects from plasma disruptions using both open loop and closed loop integration of codes. In America, workers at Oak Ridge National Laboratory, Idaho National Engineering Laboratory, and Argonne National Laboratory have looked at plasma disruption effects on the ITER blanket using the codes TSC and EDDYNET. Results show how the forces on a blanket segment depend on the number and size of the segments and on the gap between them. 9 refs., 4 figs., 1 tab
Density fluctuation in JIPP T-IIU tokamak plasmas measured by a heavy ion beam probe
International Nuclear Information System (INIS)
Multiple and small sample-volume measurements of the density turbulence and potential profile measurement in tokamak plasmas were conducted by a heavy ion beam probe. The obtained wavenumber/frequency spectrum S(k,ω) shows that the cross-section of NBI heated plasmas is divided into three regions of different turbulence characteristics. Outside the reversal layer of poloidal propagation direction of density turbulence, a low-frequency and low-wavenumber mode with ion diamagnetic drift direction dominates. The region encircled by the reversal layer is divided into two parts at nearly perpendicular NBI heating, the region where the propagation velocity is near the Er/Bt poloidal rotation velocity and the bad-curvature region of very small wavenumber and high propagation velocity. The region of high propagation velocity, found in NBI plasmas, disappears in ohmic plasmas. In addition, a small component which propagates in the ion diamagnetic drift direction is observed in NBI plasmas. (author)
Dynamical programming based turbulence velocimetry for fast visible imaging of tokamak plasma.
Banerjee, Santanu; Zushi, H; Nishino, N; Mishra, K; Onchi, T; Kuzmin, A; Nagashima, Y; Hanada, K; Nakamura, K; Idei, H; Hasegawa, M; Fujisawa, A
2015-03-01
An orthogonal dynamic programming (ODP) based particle image velocimetry (PIV) technique is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer (SOL) of tokamaks. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights into plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST plasma to estimate the SOL flow in inboard poloidal null-natural divertor configuration. This technique is also applied to investigate the intricate features of the core of the run-away dominated phase following the injection of a large amount of neutrals in the target Ohmic plasma. Development of the ODP-PIV code and its applicability on actual plasma images is reported. PMID:25832227
Unified Ideal Stability Limits for Advanced Tokamak and Spherical Torus Plasmas
International Nuclear Information System (INIS)
Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants