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
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.
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
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
Energy Technology Data Exchange (ETDEWEB)
Long-Poe Ku and Allen H. Boozer
2009-06-05
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
Geodesic Acoustic Mode in Toroidally Axisymmetric Plasmas with Non-Circular Cross Sections
Institute of Scientific and Technical Information of China (English)
SHI Bing-Ren; LI Ji-Quan; DONG Jia-Qi
2005-01-01
@@ The geodesic acoustic mode in general toroidally axisymmetric plasmas such as Tokamak and spherical torus is studied in detail. The mode structure is found and the dispersion equation is derived and solved for arbitrary toroidally axi-symmetric plasmas. Besides the finite aspect ratio, effects of elongation and triangularity on this mode are clarified.
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
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 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...
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
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
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.
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.)
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.)
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
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
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
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.)
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
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
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.
Electron thermal transport in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Konings, J.A.
1994-11-30
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 (10{sup 8} 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).
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)
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
Energy Technology Data Exchange (ETDEWEB)
Ivanov, A. A., E-mail: aai@a5.kiam.ru; Martynov, A. A., E-mail: martynov@a5.kiam.ru; Medvedev, S. Yu., E-mail: medvedev@a5.kiam.ru; Poshekhonov, Yu. Yu., E-mail: naida@a5.kiam.ru [Russian Academy of Sciences, Keldysh Institute of Applied Mathematics (Russian Federation)
2015-03-15
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.
Coupled neoclassical-magnetohydrodynamic simulations of axisymmetric plasmas
Lyons, Brendan C.
2014-10-01
Neoclassical effects (e.g., the bootstrap current and neoclassical toroidal viscosity [NTV]) have a profound impact on many magnetohydrodynamic (MHD) instabilities, including tearing modes, edge-localized modes (ELMs), and resistive wall modes. High-fidelity simulations of such phenomena require a multiphysics code that self-consistently couples the kinetic and fluid models. We present the first results of the DK4D code, a dynamic drift-kinetic equation (DKE) solver being developed for this application. In this study, DK4D solves a set of time-dependent, axisymmetric DKEs for the non-Maxwellian part of the electron and ion distribution functions (fNM) with linearized Fokker-Planck-Landau collision operators. The plasma is formally assumed to be in the low- to finite-collisionality regimes. The form of the DKEs used were derived in a Chapman-Enskog-like fashion, ensuring that fNM carries no density, momentum, or temperature. Rather, these quantities are contained within the background Maxwellian and are evolved by an appropriate set of extended MHD equations. We will discuss computational methods used and benchmarks to other neoclassical models and codes. Furthermore, DK4D has been coupled to a reduced, transport-timescale MHD code, allowing for self-consistent simulations of the dynamic formation of the ohmic and bootstrap currents. Several applications of this hybrid code will be presented, including an ELM-like pressure collapse. We will also discuss plans for coupling to the spatially three-dimensional, extended MHD code M3D-C1 and generalizing to nonaxisymmetric geometries, with the goal of performing self-consistent hybrid simulations of tokamak instabilities and calculations of NTV torque. This work supported by the U.S. Department of Energy (DOE) under Grant Numbers DE-FC02-08ER54969 and DE-AC02-09CH11466.
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.
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.
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.
Axisymmetric equilibria of a gravitating plasma with incompressible flows
Throumoulopoulos, G N
2001-01-01
It is found that the ideal magnetohydrodynamic equilibrium of an axisymmetric gravitating magnetically confined plasma with incompressible flows is governed by a second-order elliptic differential equation for the poloidal magnetic flux function containing five flux functions coupled with a Poisson equation for the gravitation potential, and an algebraic relation for the pressure. This set of equations is amenable to analytic solutions. As an application, the magnetic-dipole static axisymmetric equilibria with vanishing poloidal plasma currents derived recently by Krasheninnikov, Catto, and Hazeltine [Phys. Rev. Lett. {\\bf 82}, 2689 (1999)] are extended to plasmas with finite poloidal currents, subject to gravitating forces from a massive body (a star or black hole) and inertial forces due to incompressible sheared flows. Explicit solutions are obtained in two regimes: (a) in the low-energy regime $\\beta_0\\approx \\gamma_0\\approx \\delta_0 \\approx\\epsilon_0\\ll 1$, where $\\beta_0$, $\\gamma_0$, $\\delta_0$, and $\\...
Axisymmetric Nonlinear Waves And Structures in Hall Plasmas
Islam, Tanim
2011-01-01
A Hall plasma consists of a plasma with not all species frozen into the magnetic field. In this paper, a general equation for the evolution of an axisymmetric magnetic field in a Hall plasma is derived, with an integral similar to the Grad-Shafranov equation. Special solutions arising from curvature -- whistler drift modes that propagate along the electron drift as a Burger's shock, and nonlinear periodic and soliton-like solutions to the generalized Grad-Shafranov integral -- are analyzed. We derive analytical and numerical solutions in an electron-ion Hall plasma, in which electrons and ions are the only species in the plasmas. Results may then be applied to electron-ion-gas Hall plasmas, in which the ions are coupled to the motion of gases in low ionized plasmas (lower ionosphere and protostellar disks), and to dusty Hall plasmas (such as molecular clouds), in which the much heavier charged dust may be collisionally coupled to the gas.
Simulating Plasma Turbulence in Tokamaks
Kepner, J V; Decyk, V; Kepner, Jeremy; Parker, Scott; Decyk, Viktor
1997-01-01
A challenging and fundamental research problem is the better understanding and control of the turbulent transport of heat in present-day tokamak fusion experiments. Recent developments in numerical methods along with enormous gains in computing power have made large-scale simulations an important tool for improving our understanding of this phenomena. Simulating this highly non-linear behavior requires solving for the perturbations of the phase space distribution function in five dimensions. We use a particle-in-cell approach to solve the equations. The code has been parallelized for a variety of architectures (C90, CM-5, T3D) using a 1-D domain decomposition along the toroidal axis, for which the number of particles in each cell remains approximately constant. The quasi-uniform distribution of particles, which minimizes load imbalance, coupled with the relatively small movement of particles across cells, which minimizes communications, makes this problem ideally suited to massively parallel architectures. We...
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.
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.
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
Energy Technology Data Exchange (ETDEWEB)
Jaspers, R.E.
1995-02-03
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).
Axisymmetric nonlinear waves and structures in Hall plasmas
Energy Technology Data Exchange (ETDEWEB)
Islam, Tanim [Lawrence Livermore National Laboratory, P. O. Box 808, Livermore, California 94551-0808 (United States)
2012-06-15
In this paper, a general equation for the evolution of an axisymmetric magnetic field in a Hall plasma is derived, with an integral similar to the Grad-Shafranov equation. Special solutions arising from curvature-whistler drift modes that propagate along the electron drift as a Burger's shock and nonlinear periodic and soliton-like solutions to the generalized Grad-Shafranov integral-are analyzed. We derive analytical and numerical solutions in a classical electron-ion Hall plasma, in which electrons and ions are the only species in the plasmas. Results may then be applied to the following low-ionized astrophysical plasmas: in protostellar disks, in which the ions may be coupled to the motion of gases; and in molecular clouds and protostellar jets, in which the much heavier charged dust in a dusty Hall plasma may be collisionally coupled to the gas.
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.
Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas
Haverkort, J.W.
2013-01-01
One of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma rotation, primarily
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
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
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.
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)
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
Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bremond, S.
1995-10-18
Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.
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
A simulation study of a controlled tokamak plasma
Fujii, N.; Niwa, Y.
1980-03-01
A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.
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
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
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
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).
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.
Tokamak magnetohydrodynamic equilibrium states with axisymmetric boundary and a 3D helical core.
Cooper, W A; Graves, J P; Pochelon, A; Sauter, O; Villard, L
2010-07-16
Magnetohydrodynamic (MHD) equilibrium states with imposed axisymmetric boundary are computed in which a spontaneous bifurcation develops to produce an internal three-dimensional (3D) configuration with a helical structure in addition to the standard axisymmetric system. Equilibrium states with similar MHD energy levels are shown to develop very different geometric structures. The helical equilibrium states resemble saturated internal kink mode structures.
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].
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
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
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
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
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)
Matsuoka, Seikichi; Kanno, Ryutaro; Sugama, Hideo
2015-01-01
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 are 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 $\\delta f$ Monte Carlo method for non-Hamiltonian system [G. Hu and J. A. Krommes, Phys. Plasmas $\\rm \\textbf{1}$, 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 transpo...
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.
Feedback Control for Plasma Position on HL-2A Tokamak
Institute of Scientific and Technical Information of China (English)
LIBo; SONGXianming; LILi; LIULi; WANGMinghong; FANMingjie; CHENLiaoyuan; YAOLieying; YANGQingwei
2003-01-01
HL-2A is a tokamak with closed divertor. It had been built at the end of 2002 and began to discharge from then on. To further study plasma discharges in HL-2A, a feedback control system (FBCS) for plasma position bad been developed in 2003.
High- Q plasmas in the TFTR tokamak
Energy Technology Data Exchange (ETDEWEB)
Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (USA))
1991-08-01
In the Tokamak Fusion Test Reactor (TFTR) (Plasma Phys. Controlled Fusion {bold 26}, 11 (1984)), the highest neutron source strength {ital S}{sub {ital n}} and D--D fusion power gain {ital Q}{sub DD} are realized in the neutral-beam-fueled and heated supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, {ital S}{sub {ital n}} increases approximately as {ital P}{sup 1.8}{sub {ital b}}. The highest-{ital Q} shots are characterized by high {ital T}{sub {ital e}} (up to 12 keV), {ital T}{sub {ital i}} (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad {ital T}{sub {ital e}} profiles, and lower {ital Z}{sub eff}. 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, {ital Q}{sub DD} increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness ({ital n}{sub {ital e}}(0)/{l angle}{ital n}{sub {ital e}}{r angle}) during the beam pulse. To date, the best fusion results are {ital S}{sub {ital n}}=5{times}10{sup 16} n/sec, {ital Q}{sub DD}=1.85{times}10{sup {minus}3}, and neutron yield=4.0{times}10{sup 16} n/pulse, obtained at {ital I}{sub {ital p}}=1.6--1.9 MA and beam energy {ital E}{sub {ital b}}=95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of {ital S}{sub {ital n}} arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with {ital P}{sub {ital b}}.
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.
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
Axisymmetric Toroidal Equilibrium with Sheared Toroidal Flows
Institute of Scientific and Technical Information of China (English)
SHIBingren
2002-01-01
Since the early 1960' s, the developments of the tokamak research make plasma flows a reality in many devices where neutral beam injections were used as heating in general and refueling in particular. Compared to the static axi-symmetric toroidal equilibrium that
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)
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.
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.
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
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
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)
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.
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.)
Self-consistent hybrid neoclassical-magnetohydrodynamic simulations of axisymmetric plasmas
Lyons, Brendan Carrick
Neoclassical effects (e.g., conductivity reduction and bootstrap currents) have a profound impact on many magnetohydrodynamic (MHD) instabilities in toroidally-confined plasmas, including tearing modes, edge-localized modes, and resistive wall modes. High-fidelity simulations of such phenomena require a multiphysics code that self-consistently couples the kinetic and fluid models. We review a hybrid formulation from the recent literatureAB that is appropriate for such studies. In particular, the formulation uses a set of time-dependent drift-kinetic equations (DKEs) to advance the non-Maxwellian part of the electron and ion distribution functions (fNM) with linearized Fokker-Planck-Landau collision operators. The form of the DKEs used were derived in a Chapman-Enskog-like fashion, ensuring that fNM carries no density, momentum, or temperature. Rather, these quantities are contained within the background Maxwellian and are evolved by a set of MHD equations which are closed by moments of fNM . We then present two DKE solvers based upon this formulation in axisymmetric toroidal geometries. The Neoclassical Ion-Electron Solver (NIES) solves the steady-state DKEs in the low-collisionality limit. Convergence and benchmark studies are discussed, providing a proof-of-principle that this new formulation can accurately reproduce results from the literature in the limit considered. We then present the DK4D code which evolves the finite-collisionality DKEs time-dependently. Computational methods used and successful benchmarks to other neoclassical models and codes are discussed. Furthermore, we couple DK4D to a reduced, transport-timescale MHD code. The resulting hybrid code is used to simulate the evolution of the current density in a large-aspect-ratio plasma in the presence of several different time-dependent pressure profiles. These simulations demonstrate the self-consistent, dynamic formation of the ohmic and bootstrap currents. In the slowly-evolving plasmas considered
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.
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.
Disruption avoidance through active magnetic feedback in tokamak plasmas
Paccagnella, Roberto; Zanca, Paolo; Yanovskiy, Vadim; Finotti, Claudio; Manduchi, Gabriele; Piron, Chiara; Carraro, Lorella; Franz, Paolo; RFX Team
2014-10-01
Disruptions avoidance and mitigation is a fundamental need for a fusion relevant tokamak. In this paper a new experimental approach for disruption avoidance using active magnetic feedback is presented. This scheme has been implemented and tested on the RFX-mod device operating as a circular tokamak. RFX-mod has a very complete system designed for active mode control that has been proved successful for the stabilization of the Resistive Wall Modes (RWMs). In particular the current driven 2/1 mode, unstable when the edge safety factor, qa, is around (or even less than) 2, has been shown to be fully and robustly stabilized. However, at values of qa (qa > 3), the control of the tearing 2/1 mode has been proved difficult. These results suggested the idea to prevent disruptions by suddenly lowering qa to values around 2 where the tearing 2/1 is converted to a RWM. Contrary to the universally accepted idea that the tokamaks should disrupt at low qa, we demonstrate that in presence of a well designed active control system, tokamak plasmas can be driven to low qa actively stabilized states avoiding plasma disruption with practically no loss of the plasma internal energy.
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.
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.
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...
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)
A quasi-linear gyrokinetic transport model for tokamak plasmas
Casati, Alessandro
2012-01-01
The development of a quasi-linear gyrokinetic transport model for tokamak plasmas, ultimately designed to provide physically comprehensive predictions of the time evolution of the thermodynamic relevant quantities, is a task that requires tight links among theoretical, experimental and numerical studies. The framework of the model here proposed, which operates a reduction of complexity on the nonlinear self-organizing plasma dynamics, allows in fact multiple validations of the current understanding of the tokamak micro-turbulence. The main outcomes of this work stem from the fundamental steps involved by the formulation of such a reduced transport model, namely: (1) the verification of the quasi-linear plasma response against the nonlinearly computed solution, (2) the improvement of the turbulent saturation model through an accurate validation of the nonlinear codes against the turbulence measurements, (3) the integration of the quasi-linear model within an integrated transport solver.
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.
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
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)
Plasma radiation in tokamak disruption simulation experiments
Energy Technology Data Exchange (ETDEWEB)
Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A. [Troitsk Inst. for Innovation and Fusion Research (Russian Federation); Wuerz, H. [Forschungszentrum Karlsruhe (Germany)
1995-12-31
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 {micro}s 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) 10{sub 6} 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
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.
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)
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)
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)
3D MHD disruptions simulations of tokamaks plasmas
Paccagnella, Roberto; Strauss, Hank; Breslau, Joshua
2008-11-01
Tokamaks Vertical Displacement Events (VDEs) and disruptions simulations in toroidal geometry by means of a single fluid visco-resistive magneto-hydro-dynamic (MHD) model are presented in this paper. The plasma model, implemented in the M3D code [1], is completed with the presence of a 2D homogeneous wall with finite resistivity. This allows the study of the relatively slowly growing magneto-hydro-dynamical perturbation, the resistive wall mode (RWM), which is, in this work, the main drive of the disruptions. Amplitudes and asymmetries of the halo currents pattern at the wall are also calculated and comparisons with tokamak experimental databases and predictions for ITER are given. [1] W. Park, E.V. Belova, G.Y. Fu, X.Z. Tang, H.R. Strauss, L.E. Sugiyama, Phys. Plasmas 6 (1999) 1796.
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.
Multi-field plasma sandpile model in tokamaks and applications
Peng, X. D.; Xu, J. Q.
2016-08-01
A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.
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
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.)
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)
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.
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
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.
Plasma Theory： Toroidal Field Ripple Induced Excursion of Banana Orbit in Tokamak Plasmas
Institute of Scientific and Technical Information of China (English)
GAOQingdi
2003-01-01
Magnetic confinement of thermonuclear plasma ions within a tokamak must be achieved with a finite number of toroidal field(TF) coils. This results in a rippled toroidal field structure, and consequent distortions in fast ion orbits with potentially rapid loss of the affected ions. The ripple loss is an important issue for the design of future tokamak reactors such as ITER because it results in reduced alpha heating as well as potentially severe localized wallreactors.
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.
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.
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...
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
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
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
Predicting High Harmonic Ion Cyclotron Heating Efficiency in Tokamak Plasmas
Green, D. L.; Berry, L. A.; Chen, G.; Ryan, P. M.; Canik, J. M.; Jaeger, E. F.
2011-09-01
Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (˜kVm-1) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.
The role of plasma rotation on MHD instabilities in tokamaks
Varadarajan, V.; Miley, G. H.
An improved analysis of the linear stage of the internal kink mode has been developed to include plasma rotation and finite aspect ratio effects. The linear instability growth rates are increased by the plasma rotation. A pseudo-variational, bilinear formalism is used to discretize the linear instability equations; Fourier decomposition is used in the periodic coordinate, and a mixed-finite element procedure is adopted in the radial direction. The numerical studies with the resulting PEST-like code can be used to predict the complex plasma eigenfrequencies. The finite aspect ratio results are similar to the large aspect ratio results for flow instability. The complex instability frequencies found in the 'fishbone' and TAE modes would be strong determined by the large plasma rotation velocities observed in present-day tokamak devices. These effects could be studied by using the computationally convenient bilinear form derived from the Frieman-Rotenberg equation.
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...
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.
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
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
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
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.
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)
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
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.)
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.)
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Green, David L [ORNL; Jaeger, E. F. [XCEL; Berry, Lee A [ORNL; Chen, Guangye [ORNL; Ryan, Philip Michael [ORNL; Canik, John [ORNL
2011-01-01
Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.
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.
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)
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.
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.
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
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
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.
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.
Remote network control plasma diagnostic system for Tokamak T-10
Troynov, V. I.; Zimin, A. M.; Krupin, V. A.; Notkin, G. E.; Nurgaliev, M. R.
2016-09-01
The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet.
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}.
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
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.
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.)
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.
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.
Initial Plasma Startup Test on SUNIST Spherical Tokamak
Institute of Scientific and Technical Information of China (English)
Wang Ying(王莹); Zeng Li(曾立); He Yexi(何也熙); SUMST Team
2003-01-01
The goal of the Sino-United Spherical Tokamak (SUNIST) at Tsinghua University is to extend the understanding of toroidal plasma physics at a low aspect ratio (R/a ≈ 1.3) and to demonstrate a maintainable target plasma by non-inductive startup. The SUNIST device is designed to operate with up to 13 kA of ohmic heating field current, and to 0.15 T of toroidal field at 10 kA of discharge current. All of the poloidal fields can provide 30 mVs of Volt-seconds transformer. Experimental results of plasma startup show that SUNIST has remarkable characteristics of high ramp rate (dIp/dt ≈ 50 MA/s ), high normalized current IN of about 2.8 (IN = Ip/aBT),and high-efficiency (Ip/IROD ≈ 0.4) production of plasma current while operating at a low toroidal field. Major disruption phenomena have not been observed from magnetic diagnostics of all testing shots. Initial discharges with 52 kA of plasma current (exceeding the designed value of 50 kA),2 ms of pulse length and 50 MA/s of ramp rate have been achieved easily with pre-ionized filament.
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.
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
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.
Modelisation of synchrotron radiation losses in realistic tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Albajar, F.; Johner, J.; Granata, G
2000-08-01
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)
In situ ``artificial plasma'' calibration of tokamak magnetic sensors
Shiraki, D.; Levesque, J. P.; Bialek, J.; Byrne, P. J.; DeBono, B. A.; Mauel, M. E.; Maurer, D. A.; Navratil, G. A.; Pedersen, T. S.; Rath, N.
2013-06-01
A unique in situ calibration technique has been used to spatially calibrate and characterize the extensive new magnetic diagnostic set and close-fitting conducting wall of the High Beta Tokamak-Extended Pulse (HBT-EP) experiment. A new set of 216 Mirnov coils has recently been installed inside the vacuum chamber of the device for high-resolution measurements of magnetohydrodynamic phenomena including the effects of eddy currents in the nearby conducting wall. The spatial positions of these sensors are calibrated by energizing several large in situ calibration coils in turn, and using measurements of the magnetic fields produced by the various coils to solve for each sensor's position. Since the calibration coils are built near the nominal location of the plasma current centroid, the technique is referred to as an "artificial plasma" calibration. The fitting procedure for the sensor positions is described, and results of the spatial calibration are compared with those based on metrology. The time response of the sensors is compared with the evolution of the artificial plasma current to deduce the eddy current contribution to each signal. This is compared with simulations using the VALEN electromagnetic code, and the modeled copper thickness profiles of the HBT-EP conducting wall are adjusted to better match experimental measurements of the eddy current decay. Finally, the multiple coils of the artificial plasma system are also used to directly calibrate a non-uniformly wound Fourier Rogowski coil on HBT-EP.
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.
Transport Bifurcation Induced by Sheared Toroidal Flow in Tokamak Plasmas
Highcock, E G; Parra, F I; Schekochihin, A A; Roach, C M; Cowley, S C
2011-01-01
First-principles numerical simulations are used to describe a transport bifurcation in a differentially rotating tokamak plasma. Such a bifurcation is more probable in a region of zero magnetic shear, where the component of the sheared toroidal flow that is perpendicular to the magnetic field has the strongest suppressing effect on the turbulence, than one of finite magnetic shear. Where the magnetic shear is zero, there are no growing linear eigenmodes at any finite value of flow shear. However, subcritical turbulence can be sustained, owing to the transient growth of modes driven by the ion temperature gradient (ITG) and the parallel velocity gradient (PVG). Nonetheless, in a parameter space containing a wide range of temperature gradients and velocity shears, there is a sizeable window where all turbulence is suppressed. Combined with the relatively low transport of momentum by collisional (neoclassical) mechanisms, this produces the conditions for a bifurcation from low to high temperature and velocity gr...
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)
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)
MINERVA: Ideal MHD stability code for toroidally rotating tokamak plasmas
Aiba, N.; Tokuda, S.; Furukawa, M.; Snyder, P. B.; Chu, M. S.
2009-08-01
A new linear MHD stability code MINERVA is developed for investigating a toroidal rotation effect on the stability of ideal MHD modes in tokamak plasmas. This code solves the Frieman-Rotenberg equation as not only the generalized eigenvalue problem but also the initial value problem. The parallel computing method used in this code realizes the stability analysis of both long and short wavelength MHD modes in short time. The results of some benchmarking tests show the validity of this MINERVA code. The numerical study with MINERVA about the toroidal rotation effect on the edge MHD stability shows that the rotation shear destabilizes the intermediate wavelength modes but stabilizes the short wavelength edge localized MHD modes, though the rotation frequency destabilizes both the long and the short wavelength MHD modes.
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.
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...
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.
Characteristics of Plasma Turbulence in the Mega Amp Spherical Tokamak
Ghim, Young-chul
2013-01-01
Turbulence is a major factor limiting the achievement of better tokamak performance as it enhances the transport of particles, momentum and heat which hinders the foremost objective of tokamaks. Hence, understanding and possibly being able to control turbulence in tokamaks is of paramount importance, not to mention our intellectual curiosity of it.
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
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.
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.
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)
Cremaschini, Claudio; Slaný, Petr; Stuchlík, Zdeněk; Karas, Vladimír
2013-01-01
The possible occurrence of equilibrium off-equatorial tori in the gravitational and electromagnetic fields of astrophysical compact objects has been recently proved based on non-ideal MHD theory. These stationary structures can represent plausible candidates for the modelling of coronal plasmas expected to arise in association with accretion discs. However, accretion disc coronae are formed by a highly diluted environment, and so the fluid description may be inappropriate. The question is posed of whether similar off-equatorial solutions can be determined also in the case of collisionless plasmas for which treatment based on kinetic theory, rather than fluid one, is demanded. In this paper the issue is addressed in the framework of the Vlasov-Maxwell description for non-relativistic multi-species axisymmetric plasmas subject to an external dominant spherical gravitational and dipolar magnetic field. Equilibrium configurations are investigated and explicit solutions for the species kinetic distribution functio...
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.
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...
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
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
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...
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...
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.)
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.
Energy Technology Data Exchange (ETDEWEB)
Meglicki, Z
1995-09-19
We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.
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
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.
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...
Impurity effects on trapped electron mode in tokamak plasmas
Du, Huarong; Wang, Zheng-Xiong; Dong, J. Q.
2016-07-01
The effects of impurity ions on the trapped electron mode (TEM) in tokamak plasmas are numerically investigated with the gyrokinetic integral eigenmode equation. It is shown that in the case of large electron temperature gradient ( η e ), the impurity ions have stabilizing effects on the TEM, regardless of peaking directions of their density profiles for all normalized electron density gradient R / L n e . Here, R is the major radius and L n e is the electron density gradient scale length. In the case of intermediate and/or small η e , the light impurity ions with conventional inwardly (outwardly) peaked density profiles have stabilizing effects on the TEM for large (small) R / L n e , while the light impurity ions with steep inwardly (outwardly) peaked density profiles can destabilize the TEM for small (large) R / L n e . Besides, the TEM driven by density gradient is stabilized (destabilized) by the light carbon or oxygen ions with inwardly (outwardly) peaked density profiles. In particular, for flat and/or moderate R / L n e , two independent unstable modes, corresponding respectively to the TEM and impurity mode, are found to coexist in plasmas with impurity ions of outwardly peaked density profiles. The high Z tungsten impurity ions play a stronger stabilizing role in the TEM than the low Z impurity ions (such as carbon and oxygen) do. In addition, the effects of magnetic shear and collision on the TEM instability are analyzed. It is shown that the collisionality considered in this work weakens the trapped electron response, leading to a more stable TEM instability, and that the stabilizing effects of the negative magnetic shear on the TEM are more significant when the impurity ions with outwardly peaked density profile are taken into account.
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
Plasma boundary determination in Damavand tokamak by using current filament method
Ghadiri, Rasoul; Sadeghi, Yahya; Esteki, Mohammad Hossein
2014-05-01
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.
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)
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
Identification of the ubiquitous Coriolis momentum pinch in JET tokamak plasmas
Weisen, H.; Camenen, Y.; Salmi, A.; Versloot, T. W.; de Vries, P. C.; Maslov, M.; Tala, T.; Beurskens, M.; Giroud, C.
2012-01-01
A broad survey of the experimental database of neutral beam heated plasmas in the JET tokamak has established the theoretically expected ubiquity, in rotating plasmas, of a convective transport mechanism which has its origin in the vertical particle drift resulting from the Coriolis force. This inwa
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
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)
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.
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.
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
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)
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.
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.
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)
Watanabe, Osamu
2016-09-01
Stationary direct current in the central solenoidal coil (DCCS) of tokamak devices can reduce the non-induction heating energy necessary for tokamak plasma formation. The magnetic field energy in the inner region of the central solenoidal coil (CS region) is expelled during the tokamak plasma formation, because the vertical magnetic field intensity generated by the central solenoidal coil and poloidal field coils is partly cancelled by the increase in the toroidal plasma current. Because this magnetic field energy expelled from the CS region is distributed to the tokamak plasma in accordance with the mutual inductance, this expelled energy can drive the toroidal plasma current inductively. This energy expulsion in the CS region can be enhanced by the DCCS without the modification of the tokamak plasma configuration, when the CS coil current has negligible leakage magnetic field in the plasma area. Because the drive of the toroidal plasma current by non-induction heating can be assisted by this inductive current drive mechanism, the non-induction heating energy necessary for the tokamak plasma formation can be reduced by the DCCS. If the non-induction heating is constant, the tokamak plasma formation time can be shorted by the DCCS.
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.
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.
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.
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)
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.
A discrete adaptive near-time optimum control for the plasma vertical position in a Tokamak
Scibile, L
2001-01-01
A nonlinear controller for the plasma vertical position in a Tokamak, based on a discrete-time adaptive near time optimum control algorithm (DANTOC) is designed to stabilize the system and to maximize the state-space region over which stability can be guaranteed. The controller is also robust to the edge localized modes (ELMs) and the 600 Hz noise from the thyristor power supplies that are the primary source of disturbances and measurement noise. The controller is tested in simulation for the JET Tokamak and the results confirm its efficacy in controlling the vertical position for different plasma configurations. The controller is also tested experimentally on a real Tokamak, COMPASS-D, and the results demonstrate the improvement with respect to a simple linear PD controller in the presence of disturbances and measurement noise. The emphasis of the is on the development of the design methodology. (38 refs).
Hofmann, F.; Dutch, M. J.; Favre, A.; Martin, Y.; Moret, J.-M.; Ward, D. J.
1998-03-01
A new vertical position control system, including an internal active coil, has become operational on TCV. The new system has made it possible to stabilize plasmas with open loop growth rates up to 4400 s-1, currents up to 1.0 MA and elongations up to 2.58. The closed loop stability of the new system has been analysed with a numerical model in which the plasma is assumed to be undeformable, and the power supply outputs are delayed with respect to their inputs. Model predictions agree with the main experimental results.
Energy Technology Data Exchange (ETDEWEB)
Hofmann, F.; Dutch, M.J.; Favre, A.; Martin, Y.; Moret, J.M.; Ward, D.J. [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)
1997-12-01
A new vertical position control system, including an internal active coil, has become operational on TCV. The new system has made it possible to stabilize plasmas with open-loop growth rates up to 4400 sec{sup -1}, currents up to 1.0 MA, and elongations up to 2.58. The closed-loop stability of the new system has been analyzed with a numerical model in which the plasma is assumed undeformable, and the power supply outputs are delayed with respect to their inputs. Model predictions agree with the main experimental results. (author) 9 figs., 1 tab., 18 refs.
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)
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.
Heat convection and transport barriers in low-magnetic-shear Rijnhuizen Tokamak Project plasmas
Mantica, P.; Gorini, G.; Hogeweij, G. M. D.; Cardozo, N. J. L.; Schilham, A.M.R.
2000-01-01
Layers of reduced electron heat transport ("transport barriers") have been observed in the Rijnhuizen Tokamak Project when the plasma is dominantly heated by electron cyclotron heating (ECH). Experiments into the properties of the transport barriers are reported. Modulation of the ECH powe
Development of real-time plasma analysis and control algorithms for the TCV tokamak using Simulink
Felici, F.; Le, H. B.; J. I. Paley,; Duval, B. P.; Coda, S.; Moret, J. M.; Bortolon, A.; L. Federspiel,; Goodman, T. P.; Hommen, G.; A. Karpushov,; Piras, F.; A. Pitzschke,; J. Romero,; G. Sevillano,; Sauter, O.; Vijvers, W.; TCV team,
2014-01-01
One of the key features of the new digital plasma control system installed on the TCV tokamak is the possibility to rapidly design, test and deploy real-time algorithms. With this flexibility the new control system has been used for a large number of new experiments which exploit TCV's powerful
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
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.
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.
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
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.
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.
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
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.
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Filipe F.P.W.; Bosco, Edson del
1994-12-31
This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.
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.
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.
3D MHD VDE and disruptions simulations of tokamaks plasmas including some ITER scenarios
Paccagnella, R.; Strauss, H. R.; Breslau, J.
2009-03-01
Tokamaks vertical displacement events (VDEs) and disruptions simulations in toroidal geometry by means of a single fluid visco-resistive magneto-hydro-dynamic (MHD) model are presented in this paper. The plasma model is completed with the presence of a 2D wall with finite resistivity which allows the study of the relatively slowly growing magnetic perturbation, the resistive wall mode (RWM), which is, in this paper, the main drive of the disruption evolution. Amplitudes and asymmetries of the halo currents pattern at the wall are also calculated and comparisons with tokamak experimental databases and predictions for ITER are given.
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
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
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
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
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
Energy Technology Data Exchange (ETDEWEB)
1993-12-01
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.
Resistive wall mode and neoclassical tearing mode coupling in rotating tokamak plasmas
McAdams, Rachel; Chapman, I T
2013-01-01
A model system of equations has been derived to describe a toroidally rotating tokamak plasma, unstable to Resistive Wall Modes (RWMs) and metastable to Neoclassical Tearing Modes (NTMs), using a linear RWM model and a nonlinear NTM model. If no wall is present, the NTM growth shows the typical threshold/saturation island widths, whereas a linearly unstable kink mode grows exponentially in this model plasma system. When a resistive wall is present, the growth of the linearly unstable RWM is accelerated by an unstable island: a form of coupled RWM-NTM mode. Crucially, this coupled system has no threshold island width, giving the impression of a triggerless NTM, observed in high beta tokamak discharges. In addition, increasing plasma rotation at the island location can mitigate its growth, but does not restore the threshold width.
Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M
2012-01-10
Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant 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)
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...
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
Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers
Energy Technology Data Exchange (ETDEWEB)
Maingi, R. [North Carolina State Univ., Raleigh, NC (United States)
1992-08-01
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.
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).
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).
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.
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
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
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.
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.
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.
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.
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.
Mitchell, N
2001-01-01
In recent proposals for next generation superconducting tokamaks, such as the ITER project, the nuclear burning plasma is confined by magnetic fields generated from a large set (up to 100 GJ stored energy) of superconducting magnets. These magnets suffer heat loads in operation from thermal and nuclear radiation from the surrounding components and plasma as well as eddy currents and AC losses generated within the magnets, together with the heat conduction through supports and resistive heat generated at the current lead transitions to room temperature. The initial cryoplant for such a tokamak is expected to have a steady state capacity of up to about 85 kW at 4.5 K, comparable to the system installed for LHC at CERN. Experimental tokamaks are expected to operate at least initially in a pulsed mode with 20-30 short plasma pulses and plasma burn periods each day. A conventional cryoplant, consisting of a cold box and a set of primary heat exchangers, is ill-suited to such a mode of operation as the instantaneou...
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).
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)
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.
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.
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)
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
Resistive MHD studies of high-beta tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Lynch, V.E.; Hicks, H.R.; Holmes, J.A.; Carreras, B.A.; Garcia, L.
1982-02-01
Numerical calculations have been performed to study the magnetohydrodynamic (MHD) activity in high-beta tokamaks such as ISX-B. These initial value calculations have been built on earlier low-beta techniques, but the beta 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 beta to predominantly pressure-driven modes at high beta is described. The nonlinear studies yield x-ray emissivity plots which are compared with experiment.
Resistive MHD studies of high-. beta. -tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Lynch, V.E.; Carreras, B.A.; Hicks, H.R.; Holmes, J.A.; Garcia, L.
1981-01-01
Numerical calculations have been performed to study the MHD activity in high-..beta.. tokamaks such as ISX-B. These initial value calculations built on earlier low ..beta.. techniques, but the ..beta.. 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 ..beta.. to predominantly pressure driven modes at high ..beta.. is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment.
The construction of an electrode biasing system for driving plasma rotation in J-TEXT tokamak
Zhu, T. Z.; Chen, Z. P.; Sun, Yue; Nan, J. Y.; Liu, H.; Zhuang, G.; Wang, Z. J.
2014-05-01
A newly designed electrode biasing system has been constructed for driving plasma rotation in J-TEXT tokamak. To reduce the influence to the plasma, the system contains a pneumatic driving system so that it can reciprocate in a single discharge, with a stroke of about 5 cm in 100 ms. The power supply of the system can provide stable and adjustable dc voltage in the range of 0-700 V, with adjustable duration of 10-200 ms; its instantaneous power output can reach up to more than 200 kW. In addition, the power supply can also provide a multi-cycle voltage waveform, with adjustable pulse width and voltage amplitude. When applying a positive bias to the plasma, both an improvement of plasma confinement and the speed-up of plasma-edge toroidal rotation in the same direction of plasma current are observed in the experiments.
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;
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 vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated...... the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas...
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
Chapman, I T; Scannell, R; Cooper, W A; Graves, J P; Hastie, R J; Naylor, G; Zocco, A
2010-12-17
Thomson scattering measurements with subcentimeter spatial resolution have been made during a sawtooth crash in a Mega Ampere Spherical Tokamak fusion plasma. The unparalleled resolution of the temperature profile has shed new light on the mechanisms that underlie the sawtooth. As magnetic reconnection occurs, the temperature gradient at the island boundary increases. The increased local temperature gradient is sufficient to make the helical core unstable to ideal magnetohydrodynamic instabilities, thought to be responsible for the rapidity of the collapse.
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)
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)
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.
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
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).
Characterization of plasma current quench during disruption in EAST tokamak
Institute of Scientific and Technical Information of China (English)
陈大龙; 沈飙; 杨飞; 钱金平; 肖炳甲
2015-01-01
Preliminary analysis of plasma current quench is presented in this paper based on the disruption database. It demon-strates that 26.8%discharges have disrupted in the last 2012 campaign, in addition, plasma disruptive rate grows with the increase of plasma current. Best-fit linear and instantaneous plasma current quench rate is extracted from the recent EAST disruptions, showing that 80%–30%interval of the maximum plasma current is well fit for EAST device. The lowest area-normalized current quench time is 3.33 ms/m2 with the estimated plasma electron temperature being 7.3 eV∼9.5 eV. In the disruption case the maximum eddy current goes up to 400 kA, and a fraction of currents are respectively driven on upper and lower outer plate with nearly 100 MPa–200 MPa stress in the leg.
Kim, Juhyung; Kim, S. S.; Jhang, Hogun
2016-09-01
Numerical studies are made of the effects of resistivity on linear plasma responses to resonant magnetic perturbations (RMPs) in tokamaks based on a reduced magnetohydrodynamic model. From a local two-field model, it is suggested that the ratio of the poloidal electron advection to the resistivity diffusion rate α m can be a figure of merit parameter in linear RMP penetration physics. The shielding efficiency is governed by α m , and when α m ≳ 1 , RMPs are effectively shielded. Global simulations using a four-field model [Hazeltine and Meiss, Phys. Rep. 121, 1 (1985)] show that there exists an effective threshold of the perpendicular electron flow ( Ve , ⊥ c ) beyond which RMPs cannot penetrate. Resistivity is found to determine Ve , ⊥ c which increases as resistivity becomes higher, making RMP penetration easier. At low resistivity, small Ve , ⊥ c renders the RMP penetration sensitive to ion collisionality and the change in q95. The kink response is observed to be closely related to the residual level of RMPs at rational surfaces and can be also strongly affected by resistivity.
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)
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)
The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak
Energy Technology Data Exchange (ETDEWEB)
Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)
2000-11-01
The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.
The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak
Energy Technology Data Exchange (ETDEWEB)
Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)
2000-07-01
The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.
Measurement requirements for the advanced tokamak operation of a burning plasma experiment
Energy Technology Data Exchange (ETDEWEB)
Boivin, R L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Casper, T [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Young, K M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States)
2004-05-01
The optimization of a tokamak towards steady state and high performance has been the focus of advanced tokamak (AT) research for the past decade. A central theme of AT research line is plasma control: control of the plasma shape; of the profiles of current, pressure, and rotation; of transport; and of MHD stability. To optimize the performance, measurements of crucial parameters such as the current density and the plasma pressure are required with appropriate spatial coverage and resolution. In addition, measurements of other parameters will be necessary to develop a fundamental understanding of the complex nonlinear interactions amongst the current density profile, the pressure profile and transport (e.g. turbulence) in high {beta} AT plasmas. Present day experiments are providing physics insight into what a burning plasma experiment (BPX) will require as measurements. Recent research has focused on MHD stability aspects such as the neoclassical tearing mode and resistive wall mode stabilization and control of the current profile. However, in burning plasmas, new factors such as alpha particles, with their heating contribution and their relationship to transport barriers, will be increasingly important. The close relationship between measurements and active control, and the resultant impact on the requirements, will be discussed.
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.
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.
Edge plasma pressure measurements using a mechanical force sensor on the tokamak ISTTOK
Energy Technology Data Exchange (ETDEWEB)
Lunt, T [Humboldt-Universitaet zu Berlin, Newtonstrasse 15, 12489, Berlin (Germany); Silva, C [Associaco Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Avenida Rovisco Pais, P-1049-001 Lisbon (Portugal); Fernandes, H [Associaco Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Avenida Rovisco Pais, P-1049-001 Lisbon (Portugal); Hidalgo, C [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Pedrosa, M A [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Duarte, P [Associaco Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Avenida Rovisco Pais, P-1049-001 Lisbon (Portugal); Figueiredo, H [Associaco Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Avenida Rovisco Pais, P-1049-001 Lisbon (Portugal); Pereira, T [Associaco Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Avenida Rovisco Pais, P-1049-001 Lisbon (Portugal)
2007-11-15
In the present paper we report on a novel mechanical probe, which is able to measure the plasma pressure directly. The probe consists of two pendulums whose heads are exposed to the tokamak edge plasma, while the deflection is measured very sensitively outside the plasma by means of semi-conductor strain gauges. The plasma pressure was successfully measured in the ISTTOK edge plasma, its value being in good agreement with that derived from the electrical probe data (p{sub p} = 1-10 Pa). Furthermore, we discuss the possibility of determining the ion temperature T{sub i} = p{sub p}/n - T{sub e} by combining the pressure measurement with those of n and T{sub e} from the electrical probes. Although the derived ion temperatures-besides that in the region close to the limiter-were reasonable, its uncertainty is still very large.
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
Plasma equilibrium and field diffusion during current rise phase of STP-2 screw pinch tokamak
International Nuclear Information System (INIS)
Plasma equilibrium and field diffusion during the current rise phase of the discharge have been investigated in STP-2 screw pinch tokamak. The plasma with maximum poloidal beta value βsub(p) of 3.0 has been obtained by compression and joule heating. However the maximum βsub(p) value without strong wall contacts was about 1.3. We observed that force-free current is formed in the periphery of the plasma and the penetration rate of the poloidal magnetic field is much faster than the penetration rate calculated from the classical resistivity. To understand the high-beta plasma equilibrium and the mechanism of fast penetration rate observed in STP-2 plasmas we have performed a numerical simulation using a 2-D MHD pinch code ''TOPICS''. It is demonstrated that the fast penetration rate can be explained by introducing the influx of neutral particles and the ion acoustic type anomalous resistivity. (author)
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
Wang, Shaojie
2016-07-01
Anomalous current pinch, in addition to the anomalous diffusion due to stochastic magnetic perturbations, is theoretically found, which may qualitatively explain the recent DIII-D experiment on resonant magnetic field perturbation. The anomalous current pinch, which may resolve the long-standing issue of seed current in a fully bootstrapped tokamak, is also discussed for the electrostatic turbulence.
Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco
Basic principles and recent findings of quasi-isotropic approximation (QIA) of a geometrical optics method are presented in a compact manner. QIA was developed in 1969 to describe electromagnetic waves in weakly anisotropic media. QIA represents the wave field as a power series in two small parameters, one of which is a traditional geometrical optics parameter, equal to wavelength ratio to plasma characteristic scale, and the other one is the largest component of anisotropy tensor. As a result, "" QIA ideally suits to tokamak polarimetry/interferometry systems in submillimeter range, where plasma manifests properties of weakly anisotropic medium.
Gyrokinetic full-torus simulations of ohmic tokamak plasmas in circular limiter configuration
Korpilo, T.; Gurchenko, A. D.; Gusakov, E. Z.; Heikkinen, J. A.; Janhunen, S. J.; Kiviniemi, T. P.; Leerink, S.; Niskala, P.; Perevalov, A. A.
2016-06-01
The gyrokinetic full 5D particle distribution code ELMFIRE has been extended to simulate circular tokamak plasmas from the magnetic axis to the limiter scrape-off-layer. The predictive power of the code in the full-torus configuration is tested via its ability to reproduce experimental steady-state profiles in FT-2 ohmic L-mode plasmas. The results show that the experimental profile solution is not reproduced numerically due to the difficulty of obtaining global power balance. This is verified by cross-comparison of ELMFIRE code versions, which shows also the impact of boundary conditions and grid resolution on turbulent transport.
Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas
Energy Technology Data Exchange (ETDEWEB)
Lepson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jernigan, J. Garrett [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2016-02-05
We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.
Plasmator. A numerical code for simulation of plasma transport in Tokamaks
International Nuclear Information System (INIS)
Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)
Edge Plasma Boundary Layer Generated By Kink Modes in Tokamaks
International Nuclear Information System (INIS)
This paper describes the structure of the electric current generated by external kink modes at the plasma edge using the ideally conducting plasma model. It is found that the edge current layer is created by both wall touching and free boundary kink modes. Near marginal stability, the total edge current has a universal expression as a result of partial compensation of the (delta)-functional surface current by the bulk current at the edge. The resolution of an apparent paradox with the pressure balance across the plasma boundary in the presence of the surface currents is provided.
Ballistic propagation of turbulence front in tokamak edge plasmas
International Nuclear Information System (INIS)
The flux-driven nonlinear simulation of resistive ballooning mode turbulence with tokamak edge geometry is performed to study the non-steady component in the edge turbulence. The large-scale and dynamical events in transport are investigated in a situation where the mean flow is suppressed. Two types of dynamics are observed. One is the radial propagation of the pulse of pressure gradient, the other is the appearance/disappearance of radially elongated global structure of turbulent heat flux. The ballistic propagation is observed in the pulse of pressure gradient, which is associated with the front of turbulent heat flux. We focus on this ballistic propagation phenomenon. Both of the bump of pressure gradient and the front of heat flux propagate inward and outward direction. It is confirmed that the strong fluctuation propagates with the pulse front. It is observed that the number of pulses going outward is close to those going inward. This ballistic phenomenon does not contradict to the turbulence spreading theory. Statistical characteristics of the ballistic propagation of pulses are evaluated and compared with scaling laws which is given by the turbulence spreading theory. It is found that they give qualitatively good agreement. (paper)
The Brunt-Vaisala frequency of rotating tokamak plasmas
Haverkort, J. W.; de Blank, H. J.; Koren, B.
2012-01-01
The continuous spectrum of analytical toroidally rotating magnetically confined plasma equilibria is investigated analytically and numerically. In the presence of purely toroidal flow, the ideal magnetohydrodynamic equations leave the freedom to specify which thermodynamic quantity is constant on th
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...
Studies of instability and transport in tokamak plasmas with very weak magnetic shear
Energy Technology Data Exchange (ETDEWEB)
Dong, J.Q.; Zhang, Y.Z. [Southwestern Inst. of Physics, Chengdu (China)]|[International Center for Theoretical Physics, Trieste (Italy); Mahajan, S.M. [Texas Univ., Austin, TX (United States). Inst. for Fusion Studies
1997-04-01
Ion temperature gradient (ITG or {eta}{sub i}) driven microinstabilities are studied, using kinetic theory, for tokamak plasmas with very weak (positive or negative) magnetic shear (VWS). The gradient of magnetic shear as well as the effects of parallel and perpendicular velocity shear (v{prime}{sub {parallel}} and v{prime}{sub E}) are included in the defining equations. Two eigenmodes: the double (D) and the global (G) are found to coexist. Parametric dependence of these instabilities, and of the corresponding quasilinear transport is systematically analyzed. It is shown that, in VWS plasmas, a parallel velocity shear (PVS) may stabilize or destabilize the modes, depending on the individual as well as the relative signs of PVS and of the gradient of magnetic shear. The quasilinear transport induced by the instabilities may be significantly reduced with PVS in VWS plasmas. The v{prime}{sub E} values required to completely suppress the instabilities are much lower in VWS plasmas than they are in normal plasmas. Possible correlations with tokamak experiments are discussed.
Interaction of CLAM Steel with Plasma in HT-7 Tokamak During High Parameter Operation
Institute of Scientific and Technical Information of China (English)
LI Chunjing; HUANG Qunying; FENG Yan; LI Jiangang; KONG Mingguang
2007-01-01
A Plasma Surface Interaction(PSI)experiment on China Low Activation Martensitic(CLAM)steel was done to check if CLAM steel could be used as a Plasma Facing Material (PFM).A specimen with a diameter of 45 mm was exposed to 897 shots of deuterium plasmas with a total duration of 712 sec at a minor radius of 30 cm in HT-7 tokamak.During the exposure experiment,no observable influence Was found on plasma performance.After exposure,the surface of the specimen seemed as smooth as before but with some colour change at the margin of the specimen.Even though some micro-damage,such as dense blisters,melting,splashing,depositions,and dust,Was found on local surfaces with Scanning Electron Microscopic(SEM)observation.The reflectivity of the specimen decreased only slightly.All of these shows CLAM steel has good stability and irradiation resistance.With further optimization,it could possibly be used as the first mirror material for plasma diagnostics in tokamaks.
Full Tokamak discharge simulation and kinetic plasma profile control for ITER
International Nuclear Information System (INIS)
Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode
Theoretical and Numerical Studies of Wave-packet Propagation in Tokamak Plasmas
International Nuclear Information System (INIS)
Full text: Theoretical and numerical studies of wave-packet propagation are presented to analyze the time varying 2D mode structures of electrostatic fluctuations in tokamak plasmas, using general flux coordinates. Instead of solving the 2D wave equations directly, the solution of the initial value problem is obtained, following the propagation of wave-packets generated by a source and reconstructing the time varying field. As application, the 2D WKB method is applied to investigate the shaping effects (elongation and triangularity) of tokamak geometry on the lower hybrid wave propagation and absorption. Meanwhile, the mode structure decomposition (MSD) method is used to handle the boundary conditions and simplify the 2D problem, casted into two nested 1D problems. The MSD method is related to that discussed in earlier works and reduces to the well-known 'ballooning formalism', when spatial scale separation applies. This method is used to investigate the time varying 2D electrostatic ITG mode structure with a mixed WKB-full wave technique. The time varying field pattern is reconstructed and the time asymptotic structure of the wave-packet propagation gives the 2D eigenmode and the corresponding eigenvalue. As a general approach to investigate 2D mode structures in tokamak plasmas, our method also applies for electromagnetic waves with general source/sink terms, either by an internal/external antenna or nonlinear wave interaction with zonal structures. (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
Naturally Occurring Velocity Shear Layer at the Plasma Edge of HT-7 Tokamak
Institute of Scientific and Technical Information of China (English)
徐国盛; 万宝年; 宋梅
2004-01-01
A naturally occurring velocity shear layer was observed at the plasma edge of HT-7 tokamak in regular ohmic heated discharges. One fast reciprocating Langmuir probe was used to measure all quantities in the radial force balance equation for main ion, which enables us to present the first report about the radial force balance in the boundary region of the HT-7 tokamak. The sharp gradient of radial electric field and the reduced fluctuation correlation and turbulent particle flux characterized the edge velocity shear layer. It was found that the shear of turbulence poloidal velocity was dominated by the E × B flow shear and the poloidal rotation determined the structure of radial electric field profile and as a result the E × B flows.
Fusion plasma theory. Task III. Auxiliary heating in tokamaks and tandem mirrors. Final report
International Nuclear Information System (INIS)
The research we have accomplished with this contract has focused on ICRF coupling, heating and breakeven studies for tokamaks and ECRF fundamental and second harmonic heating in tandem mirrors. The highlights include reviewed publication of ICRF Fokker-Planck heating and breakeven studies with international collaboration with the JET group, fundamental work on a differential equation for wave fields and a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas and a formulation and code development of slab matrix and differential equation solutions for ICRF waveguide coupling in tokamak edge regions. ECRF ray tracing studies have been carried out, and a reviewed paper published for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA
Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu; Sevryukov, O. N.; Dzhumaev, P. S.; Shumskiy, V. A.; Ivannikov, A. A.
2016-09-01
Capillary-Pore Systems (CPS) filled by liquid metals are considered as an alternative solution of materials choice for plasma facing component of tokamak reactor. Tin is viewed as one of the candidates for CPS because it has lower corrosiveness than gallium and lower saturated vapour pressure compared to lithium. The corrosion resistance of Mo, Nb and W in pure liquid tin was investigated. The corrosion tests were carried out in the static isothermal conditions at a temperature up to 1050°C. As a result of the corrosion study, it was found that Mo does not corrode in liquid Sn, as opposed to Nb and is compatible with liquid tin in temperatures of up to approx. 1000°C. This allows considering Mo as an alloy base material of the in-vessel tokamak elements based on liquid tin capillary pore systems.
Sawtooth stabilization by localized electron cyclotron heating in a tokamak plasma
International Nuclear Information System (INIS)
Sawtooth oscillations (STO) in the ohmically heated WT-3 tokamak are strongly modified or suppressed by localized electron cyclotron resonance heating (ECH) near the q = 1 surface, where q refers to the safety factor. The effect of ECH is much stronger when it is applied on the high field side (the inner side of the tokamak) as compared to the low field side (outer side). Complete suppression of the STO is achieved for the duration of the ECH when it is applied on the high field side, in a low density plasma, provided the ECH power exceeds a thresholds value. The STO stabilization is attributed to a modification of the current density profile by hot electrons generated by ECH, which reduces the shear in the q = region. 14 refs., 5 figs
Coherent structures in the boundary plasma of EAST Tokamak
DEFF Research Database (Denmark)
Yan, Ning
In recent years, with the application of fast camera in fusion plasma, as well as other diagnostic of spatial-temporal resolution such as Langmuir probe, it has become generally clear that the turbulence transport is mostly dominant by cross-field propagation of coherent structures, namely blobs...... or filaments in low-confinement mode (L-mode). Analogously, the fine structures associated with the edge-localized modes (ELMs), i.e., ELM filaments, have been shown to be the main carriers of the transport in the high-confinement mode (H-mode). The filaments carry particles and heat, impinging upon the plasma......-facing material, leading to intensive transient heat load and particle load on the local areas of both the divertor target plates and the first wall, which damages the material and causes enhanced recycling and impurity generation, then further pollutes the core plasma. In this project, we carried out experiment...
Finite beta effects on turbulent transport in tokamak plasmas
International Nuclear Information System (INIS)
The research on the transport properties of magnetically confined plasmas plays an essential role towards the achievement of practical nuclear fusion energy. An economically viable fusion reactor is expected to operate at high plasma pressure. This implies that the detailed study of the impact of electromagnetic effects, whose strength increases with increasing pressure, is of critical importance. In the present work, the electromagnetic effects on the particle, momentum and heat transport channels have been investigated, with both analytical and numerical calculations. Transport processes due to a finite plasma pressure have been identified, their physical mechanisms have been explained, and their contributions have been quantified, showing that they can be significant under experimentally relevant conditions.
Fast ions and momentum transport in JET tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Salmi, A.
2012-07-01
Fast ions are an inseparable part of fusion plasmas. They can be generated using electromagnetic waves or injected into plasmas as neutrals to heat the bulk plasma and to drive toroidal rotation and current. In future power plants fusion born fast ions deliver the main heating into the plasma. Understanding and controlling the fast ions is of crucial importance for the operation of a power plant. Furthermore, fast ions provide ways to probe the properties of the thermal plasma and get insight of its confinement properties. In this thesis, numerical code packages are used and developed to simulate JET experiments for a range of physics issues related to fast ions. Namely, the clamping fast ion distribution at high energies with RF heating, fast ion ripple torque generation and the toroidal momentum transport properties using NBI modulation technique are investigated. Through a comparison of numerical simulations and the JET experimental data it is shown that the finite Larmor radius effects in ion cyclotron resonance heating are important and that they can prevent fast ion tail formation beyond certain energy. The identified mechanism could be used for tailoring the fast ion distribution in future experiments. Secondly, ASCOT simulations of NBI ions in a ripple field showed that most of the reduction of the toroidal rotation that has been observed in the JET enhanced ripple experiments could be attributed to fast ion ripple torque. Finally, fast ion torque calculations together with momentum transport analysis have led to the conclusion that momentum transport in not purely diffusive but that a convective component, which increases monotonically in radius, exists in a wide range of JET plasmas. Using parameter scans, the convective transport has been shown to be insensitive to collisionality and q-profile but to increase strongly against density gradient. (orig.)
Poloidal asymmetries in the limiter shadow plasma of the Alcator C tokamak. Volume 1
International Nuclear Information System (INIS)
This thesis investigates conditions which exist in the limiter shadow plasma of the Alcator C tokamak. The understanding of this edge plasma region is approached from both experimental and theoretical points of view. First, a general overview of edge plasma physical processes is presented. Simple edge plasma models and conditions which can theoretically result in a poloidally asymmetric edge plasma are discussed. A review of data obtained from previous diagnostics in the Alcator C edge plasma is then used to motivate the development of a new edge plasma diagnostic system (DENSEPACK) to experimentally investigate poloidal asymmetries in this region. The bulk of this thesis focuses on the marked poloidal asymmetries detected by this poloidal probe array and possible mechanisms which might support such asymmetries on a magnetic flux surface. In processing the probe data, some important considerations on fitting Langmuir probe characteristics are identified. The remainder of this thesis catalogues edge versus central plasma parameter dependences. Regression analysis techniques are applied to characterize edge density for various central plasma parameters. Edge plasma conditions during lower hybrid radio frequency heating and pellet injection are also discussed
Exposure of W-TiC/Cu Functionally Graded Materials in the Edge Plasma of HT-7 Tokamak
Institute of Scientific and Technical Information of China (English)
刘洋; 朱大焕; 陈俊凌; 周张健; 鄢容
2012-01-01
Six-layered W-TiC/Cu functionally graded materials were fabricated by resistance sintering under ultra-high pressure and exposed in the edge plasma of HT-7 tokamak. Microstruc- ture morphologies show that the TiC particles distribute homogeneously in the W matrix, strength- ening the grain boundary, while gradient layers provide a good compositional transition from W- TiC to Cu. After about 360 shots in the HT-7 tokamak, clear surface modification can be observed after plasma exposure, and the addition of nano TiC particles is beneficial to the improvement of plasma loads resistance of W.
Influence of external resonant magnetic perturbation field on edge plasma of small tokamak HYBTOK-II
Energy Technology Data Exchange (ETDEWEB)
Hayashi, Y., E-mail: hayashi-yuki13@ees.nagoya-u.ac.jp [Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Suzuki, Y.; Ohno, N. [Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Okamoto, M. [Ishikawa National College of Technology, Kitachujo, Tsubata-cho, Kahoku-gun, Ishikawa 929-0392 (Japan); Kikuchi, Y. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Sakakibara, S.; Watanabe, K.; Takemura, Y. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292 (Japan)
2015-08-15
Radial profile of externally applied resonant magnetic perturbation (RMP) field with mode numbers of m = 6 and n = 2 in a small tokamak device HYBTOK-II have been investigated using a magnetic probe array, which is able to measure the radial profile of magnetic field perturbation induced by applying RMP. Results of RMP penetration into the plasma show that the RMP decreased toward the plasma center, while they were amplified around the resonant surface with a safety factor q = 3 due to the formation of magnetic islands. This suggests that RMP fields for controlling edge plasmas may trigger some kind of MHD instabilities. In addition, simulation results, based on a linearized four-field model, which agrees with the experimental ones, indicates that the penetration and amplification process of RMP strongly depend on a Doppler-shifted frequency between the RMP and plasma rotation.
International Nuclear Information System (INIS)
It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs
Stability of Tokamaks with respect to slip motions
International Nuclear Information System (INIS)
Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.)
Transport studies of tokamak plasmas in JT-60 device
International Nuclear Information System (INIS)
In order to study the magnetic flux exchange at the sawtooth crash, the nonlinear evolution of the tearing mode with poloidal mode number m=1 and toroidal mode number n=1 in the presence of two or three resonant surfaces satisfying q=m/n=1 is examined. It is clarified that whether the local magnetic flux exchange or the internal disruption occurs depends on the initial profile of helical flux function φ0. Light impurities such as carbon and oxygen and metal impurities such as titanium and molybdenum cause dominant radiation loss in the high density regime and give rise to major disruption. Increasing density by gas puffing while keeping the impurity content constant, the density limit of ohmically heated plasmas is studied. The numerically obtained operational diagram or Hugill diagram, qualitatively agrees well with that obtained in the experiment. Characteristics of global parameter dependence of JT-60 plasmas in both ohmic heating phase and neutral beam heating phase have been studied by using the models of electron and ion thermal diffusivities, χe and χi. The predicted Te profile and the energy confinement time, τE, show good agreement with experimental data in the medium line for both ohmic heating and neutral beam heating phase. The power degradation of energy confinement time (τEth ∝Pabs-1/2) is also reproduced. Ion temperature profiles of neutral beam heated phase of JT-60 have been simulated by using a χi model based on ηi mode turbulence and the drift wave turbulence. In case of TFTR L-mode plasmas, the simulated Ti profiles become flatter than experimental data. In the high Ti plasmas of JT-60 and supershot plasmas of TFTR, the predicted Ti profiles become broader compared with the experimental results. (J.P.N.)
Scattering of radio frequency waves by blobs in tokamak plasmas
International Nuclear Information System (INIS)
The density fluctuations and blobs present in the edge region of magnetic fusion devices can scatter radio frequency (RF) waves through refraction, reflection, diffraction, and coupling to other plasma waves. This, in turn, affects the spectrum of the RF waves and the electromagnetic power that reaches the core of the plasma. The usual geometric optics analysis of RF scattering by density blobs accounts for only refractive effects. It is valid when the amplitude of the fluctuations is small, of the order of 10%, compared to the background density. In experiments, density fluctuations with much larger amplitudes are routinely observed, so that a more general treatment of the scattering process is needed. In this paper, a full-wave model for the scattering of RF waves by a blob is developed. The full-wave approach extends the range of validity well beyond that of geometric optics; however, it is theoretically and computationally much more challenging. The theoretical procedure, although similar to that followed for the Mie solution of Maxwell's equations, is generalized to plasmas in a magnetic field. Besides diffraction and reflection, the model includes coupling to a different plasma wave than the one imposed by the external antenna structure. In the model, it is assumed that the RF waves interact with a spherical blob. The plasma inside and around the blob is cold, homogeneous, and imbedded in a uniform magnetic field. After formulating the complete analytical theory, the effect of the blob on short wavelength electron cyclotron waves and longer wavelength lower hybrid waves is studied numerically
Scattering of radio frequency waves by blobs in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Ram, Abhay K. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hizanidis, Kyriakos; Kominis, Yannis [School of Electrical and Computer Engineering, National Technical University of Athens, Association EURATOM-Hellenic Republic, Athens, GR-15773 (Greece)
2013-05-15
The density fluctuations and blobs present in the edge region of magnetic fusion devices can scatter radio frequency (RF) waves through refraction, reflection, diffraction, and coupling to other plasma waves. This, in turn, affects the spectrum of the RF waves and the electromagnetic power that reaches the core of the plasma. The usual geometric optics analysis of RF scattering by density blobs accounts for only refractive effects. It is valid when the amplitude of the fluctuations is small, of the order of 10%, compared to the background density. In experiments, density fluctuations with much larger amplitudes are routinely observed, so that a more general treatment of the scattering process is needed. In this paper, a full-wave model for the scattering of RF waves by a blob is developed. The full-wave approach extends the range of validity well beyond that of geometric optics; however, it is theoretically and computationally much more challenging. The theoretical procedure, although similar to that followed for the Mie solution of Maxwell's equations, is generalized to plasmas in a magnetic field. Besides diffraction and reflection, the model includes coupling to a different plasma wave than the one imposed by the external antenna structure. In the model, it is assumed that the RF waves interact with a spherical blob. The plasma inside and around the blob is cold, homogeneous, and imbedded in a uniform magnetic field. After formulating the complete analytical theory, the effect of the blob on short wavelength electron cyclotron waves and longer wavelength lower hybrid waves is studied numerically.
Optimisation of out-vessel magnetic diagnostics for plasma boundary reconstruction in tokamaks
Romero, J A
2013-01-01
To improve the low frequency spectrum of magnetic field measurements of future tokamak reactors such as ITER, several steady state magnetic sensor technologies have been considered. For all the studied technologies it is always advantageous to place the sensors outside the vacuum vessel and as far away from the reactor core to minimize radiation damage and temperature effects, but not so far as to compromise the accuracy of the equilibrium reconstruction. We have studied to what extent increasing the distance between out-vessel sensors and plasma can be compensated for sensor accuracy and/or density before the limit imposed by the degeneracy of the problem is reached. The study is particularized for the Swiss TCV tokamak, due to the quality of its magnetic data and its ability to operate with a wide range of plasma shapes and divertor configurations. We have scanned the plasma boundary reconstruction error as function of out-vessel sensor density, accuracy and distance to the plasma. The study is performed fo...
International Nuclear Information System (INIS)
The SMM wave laser scattering apparatus has been developed for the measurement of the waves and turbulences in the plasma. This apparatus will help greatly to clarify the physics of RF heating of the tokamak plasma. The present status of main parts of the apparatus, the SMM wave laser and the Schottky barrier diode mixer for the heterodyne receiver, are described. (author)
International Nuclear Information System (INIS)
It is shown how the properties of the electron cyclotron emission of a tokamak plasma can be used to measure the electron temperature. The design of a six channel Fabry-Perot interferometer is then described. This interferometer allows the measurement of the time evolution of the electron temperature profile of the plasma in the TFR tokamak. Using this technique interesting results have been obtained concerning the current penetration during the start up phase of a tokamak discharge
Characteristics of disruptive plasma current decay in the low loop resistance tokamak HT-2
Energy Technology Data Exchange (ETDEWEB)
Abe, Mitsushi; Doi, Akira; Otsuka, Michio [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.
1995-07-01
Experiments with two different vacuum vessel loop resistances {Omega}{sub V} were carried out to examine the characteristics of disruptive plasma current decay in the HT-2 tokamak. The resistance {Omega}{sub V} was originally high (14m{Omega}), but {Omega}{sub V} was modified to have a low {Omega}{sub V} of 0.3m{Omega}, which was less than the plasma loop resistance calculated at electron temperature 5eV. The disruptive plasma current decay can be divided into two phases. During the first phase, the plasma has a magnetic axis and closed flux surfaces. Modification of the resistance does not affect disruptive characteristics in this phase. During the second phase, the plasma has no closed flux surface and a rapid thermal energy loss occurs from the plasma. Consequently the plasma loop resistance becomes larger than the modified {Omega}{sub V} and replacement of the toroidal current from the plasma to the vacuum vessel occurs at the start of this phase. The {Omega}{sub V} modification does not affect the total force on the plasma, but it does change the spatial distribution and local force direction. The main toroidal vessel current component is a uniform component in the poloidal distribution. (author).
New regime of low ion collisionality in the neoclassical equilibrium of tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Ramos, J. J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139-4307 (United States)
2015-07-15
The neoclassical description of an axisymmetric toroidal plasma equilibrium is formulated for an unconventionally low ordering of the collisionality that suits realistic thermonuclear fusion conditions. This requires a drift-kinetic analysis to the second order of the ion Larmor radius, which yields a new contribution to the leading solution for the non-Maxwellian part of the ion distribution function if the equilibrium geometry is not up-down symmetric. An explicit geometrical factor weighs this second Larmor-radius order, low-collisionality effect that modifies the neoclassical ion parallel flow, and the ion contribution to the bootstrap current.
Optimal control of tokamak and stellarator plasma behaviour
Energy Technology Data Exchange (ETDEWEB)
Rastovic, Danilo [Control Systems Group, Nehajska 62, 10000 Zagreb (Croatia)]. E-mail: drastovi@tesla.vtszg.hr
2007-04-15
The control of plasma transport, laminar and turbulent, is investigated, using the methods of scaling, optimal control and adaptive Monte Carlo simulations. For this purpose, the asymptotic behaviour of kinetic equation is considered in order to obtain finite-dimensional invariant manifolds, and in this way the finite-dimensional theory of control can be applied. We imagine the labyrinth of open doors and after applying self-similarity, the motion moved through all the desired doors in repeatable ways as Brownian motions. We take local actions for each piece of contractive ergodic motion, and, after self-organization in adaptive invariant measures, the optimum movement of particles is obtained according to the principle of maximum entropy. This is true for deterministic and stochastic cases that serve as models for plasma dynamics.
Kelvin-Helmholtz instabilities in tokamak edge plasmas
Energy Technology Data Exchange (ETDEWEB)
Garbet, X.; Fenzi, C.; Capes, H.; Devynck, P.; Antar, G
1999-07-15
The parallel Kelvin-Helmholtz instability is investigated as a possible explanation for poloidal asymmetries of density fluctuations which reverse with the plasma current direction. It is shown that these modes are localised around the position where the radial gradient of parallel velocity is maximum. Two mechanisms lead to unstable Kelvin-Helmholtz modes: the acceleration of ions in a presheath and the anomalous Stringer spin-up due to asymmetries of the particle flux. Up-down asymmetries are explained by combining these two effects. Depending on the limiter configuration, the Stringer effect amplifies or weakens the flow due to presheath acceleration. This type of asymmetry reverses with the plasma current direction. (authors)
Time dependent neutral gas transport in tokamak edge plasmas
Energy Technology Data Exchange (ETDEWEB)
Reiter, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-KFA Postfach 1913, D-52425, Juelich (Germany); May, C. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-KFA Postfach 1913, D-52425, Juelich (Germany); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85740, Garching (Germany); Schneider, R. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85740, Garching (Germany)
1995-04-01
The effects of neutral particles on the edge plasma conditions play a key role in divertor and limiter physics. In computational models they are usually treated in steady state approximation (instantaneous relaxation). However, the characteristic transport time scale is comparable to the ion acustic time scale. Thus neutral atoms relax to their steady state distributions much slower than electron temperature profiles along the fieldlines are established. A computational assessment of divertor or limiter dynamics requires ultimately an extension to time dependent algorithms. The numerical procedure in the EIRENE Monte Carlo code is presented. A first numerical study of ELM`s in the ASDEX-Upgrade divertor plasma has been carried out and the results are briefly discussed. ((orig.)).
International Nuclear Information System (INIS)
An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design
Trapped Electron Mode Turbulence Driven Intrinsic Rotation in Tokamak Plasmas
International Nuclear Information System (INIS)
Recent progress from global gyrokinetic simulations in understanding the origin of intrinsic rotation in toroidal plasmas is reported with emphasis on electron thermal transport dominated regimes. The turbulence driven intrinsic torque associated with nonlinear residual stress generation by the fluctuation intensity and the intensity gradient in the presence of zonal flow shear induced asymmetry in the parallel wavenumber spectrum is shown to scale close to linearly with plasma gradients and the inverse of the plasma current. These results qualitatively reproduce empirical scalings of intrinsic rotation observed in various experiments. The origin of current scaling is found to be due to enhanced kll symmetry breaking induced by the increased radial variation of the safety factor as the current decreases. The physics origin for the linear dependence of intrinsic torque on pressure gradient is that both turbulence intensity and the zonal flow shear, which are two key ingredients for driving residual stress, increase with the strength of turbulence drive, which is R0/LTe and R0/Lne for the trapped electron mode.
Scattering of radio frequency waves by cylindrical density filaments in tokamak plasmas
Ram, Abhay K.; Hizanidis, Kyriakos
2016-02-01
In tokamak fusion plasmas, coherent fluctuations in the form of blobs or filaments are routinely observed in the scrape-off layer. Radio frequency (RF) electromagnetic waves, excited by antenna structures placed near the wall of a tokamak, have to propagate through the scrape-off layer before reaching the core of the plasma. While the effect of fluctuations on the properties of RF waves has not been quantified experimentally, it is of interest to carry out a theoretical study to determine if fluctuations can affect the propagation characteristics of RF waves. Usually, the difference between the plasma density inside the filament and the background plasma density is sizable, the ratio of the density difference to the background density being of order one. Generally, this precludes the use of geometrical optics in determining the effect of fluctuations, since the relevant ratio has to be much less than one, typically, of the order of 10% or less. In this paper, a full-wave, analytical model is developed for the scattering of a RF plane wave by a cylindrical plasma filament. It is assumed that the plasma inside and outside the filament is cold and uniform and that the major axis of the filament is aligned along the toroidal magnetic field. The ratio of the density inside the filament to the density of the background plasma is not restricted. The theoretical framework applies to the scattering of any cold plasma wave. In order to satisfy the boundary conditions at the interface between the filament and the background plasma, the electromagnetic fields inside and outside the filament need to have the same k∥ , the wave vector parallel to the ambient magnetic field, as the incident plane wave. Consequently, in contrast to the scattering of a RF wave by a spherical blob [Ram et al., Phys. Plasmas 20, 056110-1-056110-10 (2013)], the scattering by a field-aligned filament does not broaden the k∥ spectrum. However, the filament induces side-scattering leading to surface
Energy Technology Data Exchange (ETDEWEB)
Wang, W. X. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Ethier, S. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Ren, Y. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Kaye, S. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Chen, J. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Startsev, E. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543, USA; Lu, Z. [University of California, San Diego, La Jolla, California 92093, USA; Li, Z. Q. [Zhejiang University, Hangzhou, People' s Republic of China
2015-10-01
Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around k(theta)rho(s) similar to 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport
Study of plasma turbulence by ultrafast sweeping reflectometry on the Tore Supra Tokamak
International Nuclear Information System (INIS)
The performance of a fusion reactor is closely related to the turbulence present in the plasma. The latter is responsible for anomalous transport of heat and particles that degrades the confinement. The measure and characterization of turbulence in tokamak plasma is therefore essential to the understanding and control of this phenomenon. Among the available diagnostics, the sweeping reflectometer installed on Tore Supra allows to access the plasma density fluctuations from the edge to the centre of the plasma discharge with a fine spatial (mm) and temporal resolution (μs), that is of the order of the characteristic turbulence scales.This thesis consisted in the characterization of plasma turbulence in Tore Supra by ultrafast sweeping reflectometry measurements. Correlation analyses are used to quantify the spatial and temporal scales of turbulence as well as their radial velocity. In the first part, the characterization of turbulence properties from the reconstructed plasma density profiles is discussed, in particular through a comparative study with Langmuir probe data. Then, a parametric study is presented, highlighting the effect of collisionality on turbulence, an interpretation of which is proposed in terms of the stabilization of trapped electron turbulence in the confined plasma. Finally, it is shown how additional heating at ion cyclotron frequency produces a significant though local modification of the turbulence in the plasma near the walls, resulting in a strong increase of the structure velocity and a decrease of the correlation time. The supposed effect of rectified potentials generated by the antenna is investigated via numerical simulations. (author)
Wave heating of an ion beam in a tokamak plasma
International Nuclear Information System (INIS)
Heating of a 26-keV trapped-ion beam by interaction with incident lower-hybrid RF power (as low as 5kW) was observed in the ATC plasma. We suggest that ion-cyclotron damping of lower hybrid waves by beam ions can account for the increase in beam energy. This process can explain the main features of the experiment: (1) preferential absorption in the perpendicular direction, (2) lack of absorption by the background plasma ions, and (3) low power requirement for absorption. Theory requires ksub(perpendicular)rhosub(i) approximately > (ω/ωsub(ci))sup(1/2). The relatively high perpendicular temperature of the beam ions (approximately 1keV), combined with one of several possibilities for RF energy at large ksub(perpendicular), allows the condition on ksub(perpendicular)rhosub(i) to be satisfied. Moreover, the large parallel energy of the beam ions plays a major role in broadening the harmonic resonances, thus making it possible for a large number of beam ions to resonate with the wave. Though the primary process is perpendicular absorption, there is also a net gain in parallel energy during a collision time due to pitch-angle scattering. The importance of this heating mechanism for large machines such as TFTR is discussed. For these machines, ions will be injected with large rhosub(i), making ion heating possible even with moderate values of ksub(perpendicular). (author)
Effect of biasing on plasma rotation in the edge of IR-T1 Tokamak
Energy Technology Data Exchange (ETDEWEB)
Mohammadi, S.; Ghoranneviss, M.; Arvin, R.; Gheydi, M.; Nikmohammadi, A. [Plasma physics Research Center, Science and Research Branch, Islamic Azad University, P.O.Box: 14665-768 Tehran (Iran, Islamic Republic of); Khorshid, P.; Bolourian, H. [Department of Physics, Islamic Azad University, Mashhad Branch, Mashhad (Iran, Islamic Republic of)
2011-07-01
Full text of publication follows: Electrode biasing experiments were carried out on the IR-T1 Tokamak. The effects of radial electric field (Er) on plasma fluid velocity and magnetic island rotation investigated by a Mach/Langmuir electric probe and an array of 12 Mirnov coils. The Results have shown a change in the fluid velocity during biasing regime. References: [1] Van Oost G. et al. 2001 Czech. J. of Phys. 51 957; [2] Effect of Plasma Biasing on Suppression of Electrostatic Fluctuation in the Edge Region of STP-3(M) Reversed Field Pinch J. Phys. Soc. Jpn. 74 (2005) pp.605-612; [3] Weynants R. R. and Van Oost G. 1993 Plasma Phys. Contr. Fusion 35 B177. (authors)
Effects of high Z probe on plasma behavior in HT-6M tokamak
International Nuclear Information System (INIS)
Molybdenum and tungsten probes have been tested in HT-6M tokamak under various discharge conditions aiming to find out the conditions in which high Z PFC can be used without serious degradation of core plasma performance. In normal OH discharges, the degradation of core plasma performance was found only when the probe was inserted beyond 3.0 cm inside the last closed flux surface (LCFS). The plasma performance did not change with positive biasing to the probe, whereas central Te degraded during negative biasing of -100 V. The insertion of the Mo probe to 1.5 cm inside the LCFS made a change in the threshold power of the L-H transition in EOH discharges. These results suggest a certain operation range of the H-mode in the EOH discharge with the Mo probe in HT-6M. (author)
Effects of high Z probe on plasma behavior in HT-6M tokamak
International Nuclear Information System (INIS)
Molybdenum and tungsten probes have been tested in HT-6M tokamak under various discharge conditions aiming to find out the conditions in which high Z PFC can be used without serious degradation of core plasma performance. In normal OH discharges, the degradation of core plasma performance was found only when the probe was inserted beyond 3.0 cm inside the last closed flux surface (LCFS). The plasma performance did not change with positive biasing to the probe, whereas central Te degraded during negative biasing of -100 V. The insertion of the Mo probe to 1.5 cm inside the LCFS made a change in the threshold power of the L-H transition in EOH discharges. These results suggest a certain operation range of the H-mode in the EOH discharge with the Mo probe in HT-6M. (orig.)
Effects of high Z probe on plasma behavior in HT-6M tokamak
Li, J.; Gong, X.; Luo, L.; Yin, F. X.; Noda, N.; Wan, B.; Xu, W.; Gao, X.; Yin, F.; Jiang, J. G.; Wu, Z.; Zhao, J. Y.; Wu, M.; Liu, S.; Han, Y.
1997-02-01
Molybdenum and tungsten probes have been tested in HT-6M tokamak under various discharge conditions aiming to find out the conditions in which high Z PFC can be used without serious degradation of core plasma performance. In normal OH discharges, the degradation of core plasma performance was found only when the probe was inserted beyond 3.0 cm inside the last closed flux surface (LCFS). The plasma performance did not change with positive biasing to the probe, whereas central Te degraded during negative biasing of -100 V. The insertion of the Mo probe to 1.5 cm inside the LCFS made a change in the threshold power of the L-H transition in EOH discharges. These results suggest a certain operation range of the H-mode in the EOH discharge with the Mo probe in HT-6M.
Hsu, S C
2006-01-01
A spherical tokamak (ST) with a plasma center column (PCC) can be formed via driven magnetic relaxation of a screw pinch plasma. An ST-PCC could in principle eliminate many problems associated with a material center column, a key weakness of the ST reactor concept. This work summarizes the design space for an ST-PCC in terms of flux amplification, aspect ratio, and elongation, based on the zero-beta Taylor-relaxed analysis of Tang & Boozer [Phys. Plasmas 13, 042514 (2006)]. The paper will discuss (1) equilibrium and stability properties of the ST-PCC, (2) issues for an engineering design, and (3) key differences between the proposed ST-PCC and the ongoing Proto-Sphera effort in Italy.
International Nuclear Information System (INIS)
In future fusion reactors, density control, such as fueling by pellet injection, is an effective method to control the formation of the internal transport barrier (ITB) in reversed magnetic shear plasma, which can improve plasma performance. On the other hand, an operation with ITB can cause accumulation of impurities inside the core ITB region. We studied the relation between pellet injection and ITB formation and the effect of impurity transport on the core of ITB for tokamak plasmas by using the toroidal transport analysis linkage. For ITB formation, we showed that the pellet has to be injected beyond the position where the safety factor q takes the minimum value. We confirmed that the accumulation of impurities causes the attenuation of ITB owing to radiation loss inside the ITB region. Moreover, in terms of the divertor heat flux reduction by impurity gas, the line radiation loss is high for high-Z noble gas impurities, such as Kr, whereas factor Q decreases slightly. (author)
Electron heating by mode-converted ion-Bernstein waves in ICRF heating of tokamak plasmas
International Nuclear Information System (INIS)
In a tokamak plasma, ion-Bernstein waves (IBW) can be excited by mode-conversion of the externally launched fast wave for ICRF heating. This conversion process is known to be efficient for low k/sub parallel/'s which carry substantial power from a single loop antenna. A detailed numerical analysis of the propagation of the IBW shows that the initial small k/sub parallel/ are significantly enhanced along the rays due to toroidal effects. The upshift can occur for short radial distances of propagation and is large enough so that the IBW can Landau damp onto the electrons. This could help explain the observed strong electron heating by ICRF waves in tokamak plasmas. The numerical ray trajectory analysis is done in toroidal geometry for a hot Maxwellian plasma with gradients in temperature, density, toroidal and poloidal magnetic fields included in a WKB sense. A simple analytical model is derived which explains the upshift in k/sub parallel/ and gives results very close to the numerically obtained values. Approximate analytical conditions for appreciable electron Landau damping of the IBW are also given
Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma
Energy Technology Data Exchange (ETDEWEB)
Xu, Liqing; Zhang, Jizong; Chen, Kaiyun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Hu, Liqun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhu, Yubao [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)
2015-12-15
Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.
Performance of a New Ion Source for KSTAR Tokamak Plasma Heating
Tae-Seong, Kim; Seung, Ho Jeong; Doo, Hee Chang; Kwang, Won Lee; Sang-Ryul, In
2014-06-01
In the experimental campaign of 2010 and 2011 on KSTAR, the NBI-1 system was equipped with one prototype ion source and operated successfully, providing a neutral beam power of 0.7-1.6 MW to the tokamak plasma. The new ion source planned for the 2012 KSTAR campaign had a much more advanced performance compared with the previous one. The target performance of the new ion source was to provide a neutral deuterium beam of 2 MW to the tokamak plasma. The ion source was newly designed, fabricated, and assembled in 2011. The new ion source was then conditioned up to 64 A/100 keV over a 2-hour beam extraction and performance tested at the NB test stand (NBTS) at the Korea Atomic Energy Research Institute (KAERI) in 2012. The measured optimum perveance at which the beam divergence is a minimum was about 2.5 μP, and the minimum beam divergent angle was under 1.0° at 60 keV. These results indicate that the 2.0 MW neutral beam power at 100 keV required for the heating of plasma in KSTAR can be delivered by the installation of the new ion source in the KSTAR NBI-1 system.
Calculated radiative power losses from mid- and high-Z impurities in Tokamak plasmas
Fournier, Kevin B.; May, M. J.; Pacella, D.; Gregory, B. C.; Rice, J. E.; Terry, J. L.; Finkenthal, M.; Goldstein, W. H.
1998-09-01
This paper summarizes recent calculations of the radiative cooling coefficient for molybdenum (Z=42), krypton (Z=36) and argon (Z=18). The radiative processes considered are collisional-radiative line emission, dielectronic recombination line emission, and radiative recombination and bremsstrahlung continuum emission. Collisional-radiative line emission dominates the power loss channels for a given impurity at all but the highest plasma electron temperatures. The atomic data for the line emission are computed ab initio with the HULLAC atomic physics suite of codes. Relativistic, ab initio atomic physics data are used to compute ionization and recombination rate coefficients; the resulting charge state distribution and recombination rates are used to estimate the radiative power from recombination processes. The calculations in the present work are benchmarked against absolute measurements of ion brightness profiles in the Frascati Tokamak Upgrade plasma. Integrated measurements from tokamak plasmas such as bolometry are then simulated. The atomic physics data used to predict the emissivity of individual ions is validated; the calculated cooling coefficients agree well with bolometric measurements.
Calculated radiative power losses from mid- and high-Z impurities in Tokamak plasmas
International Nuclear Information System (INIS)
This paper summarizes recent calculations of the radiative cooling coefficient for molybdenum (Z=42), krypton (Z=36) and argon (Z=18). The radiative processes considered are collisional-radiative line emission, dielectronic recombination line emission, and radiative recombination and bremsstrahlung continuum emission. Collisional-radiative line emission dominates the power loss channels for a given impurity at all but the highest plasma electron temperatures. The atomic data for the line emission are computed ab initio with the HULLAC atomic physics suite of codes. Relativistic, ab initio atomic physics data are used to compute ionization and recombination rate coefficients; the resulting charge state distribution and recombination rates are used to estimate the radiative power from recombination processes. The calculations in the present work are benchmarked against absolute measurements of ion brightness profiles in the Frascati Tokamak Upgrade plasma. Integrated measurements from tokamak plasmas such as bolometry are then simulated. The atomic physics data used to predict the emissivity of individual ions is validated; the calculated cooling coefficients agree well with bolometric measurements
Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma
International Nuclear Information System (INIS)
Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well
Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas
DEFF Research Database (Denmark)
Militello, F; Fundamenski, W; Naulin, Volker;
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...... of the edge/SOL density and temperature. In addition, we also discuss how the system changes when the length of the divertor leg is modified. This allows one to better understand the regime of operation of the Super-X divertor which will be implemented on MAST-Upgrade. The results obtained qualitatively agree...
Thermonuclear-driven fast magnetosonic-wave heating in tokamak plasmas
International Nuclear Information System (INIS)
A thermonuclear driven fast magnetosonic wave instability is investigated in tokamak plasmas for propagation transverse to the external magnetic field at frequencies of several times the alpha particle gyro rate: ω approx. = L Ω/sub α/ = k/sub perpendicular/ v/sub A/, L approx. 4 to 8, k/sub parallel/ << k/sub perpendicular/. The 2-D differential quasi-linear diffusion equation is derived in circular cylindrical, v/sub perpendicular/-v/sub parallel/ geometry. We perform an expansion in the small parameter k/sub parallel/k/sub perpendicucular/ of the quasi-linear diffusion coefficients
Plasma Turbulence in the Scrape-off Layer of the ISTTOK Tokamak
Jorge, Rogerio; Halpern, Federico D; Loureiro, Nuno F; Silva, Carlos
2016-01-01
The properties of plasma turbulence in a poloidally limited scrape-off layer (SOL) are addressed, with focus on ISTTOK, a large aspect ratio tokamak with a circular cross section. Theoretical investigations based on the drift-reduced Braginskii equations are carried out through linear calculations and non-linear simulations, in two- and three-dimensional geometries. The linear instabilities driving turbulence and the mechanisms that set the amplitude of turbulence as well as the SOL width are identified. A clear asymmetry is shown to exist between the low-field and the high-field sides of the machine. A comparison between experimental measurements and simulation results is presented.
Production and maintenance of high poloidal beta tokamak plasmas by means of rf current drive
Energy Technology Data Exchange (ETDEWEB)
Luckhardt, S.C.; Chen, K.; Coda, S.; Kesner, J.; Kirkwood, R.; Lane, B.; Porkolab, M.; Squire, J.
1989-03-27
It is shown that in tokamak plasmas sustained by rf current drive, the contribution of the suprathermal rf-driven electron population to the poloidal beta (..beta../sub p/) can be substantial if the total current is comparable to the Alfven critical current, I/sub A/ = (4..pi..mcv/..mu../sub 0/ec)..gamma... Equilibria with values of epsilon..beta../sub p/ up to approximately 1.3 were obtained, and no equilibrium or gross stability limits were observed.
Geodesic acoustic modes in tokamak plasmas with a radial equilibrium electric field
Energy Technology Data Exchange (ETDEWEB)
Zhou, Deng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People' s Republic of China and Centre for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)
2015-09-15
The dispersion relation of geodesic acoustic modes in the tokamak plasma with an equilibrium radial electric field is derived and analyzed. Multiple branches of eigenmodes have been found, similar to the result given by the fluid model with a poloidal mass flow. Frequencies and damping rates of both the geodesic acoustic mode and the sound wave increase with respect to the strength of radial electric field, while the frequency and the damping rate of the lower frequency branch slightly decrease. Possible connection to the experimental observation is discussed.
Kinetic theory and simulation of multi-species plasmas in tokamaks excited with ICRF microwaves
International Nuclear Information System (INIS)
This paper presents a description of a bounce-averaged Fokker-Planck quasilinear model for the kinetic description of tokamak plasmas. The non-linear collision and quasilinear resonant diffusion operators are represented in a form conducive to numerical solution with specific attention to the treatment of the boundary layer separating trapped and passing orbit regions of velocity space. The numerical techniques employed are detailed in so far as they constitute significant departure from those used in the conventional uniform magnetic field case. Examples are given to illustrate the combined effects of collisional and resonant diffusion
Strong Scattering of High Power Millimeter Waves in Tokamak Plasmas with Tearing Modes
DEFF Research Database (Denmark)
Westerhof, E.; Nielsen, Stefan Kragh; Oosterbeek, J.W.;
2009-01-01
In tokamak plasmas with a tearing mode, strong scattering of high power millimeter waves, as used for heating and noninductive current drive, is shown to occur. This new wave scattering phenomenon is shown to be related to the passage of the O point of a magnetic island through the high power...... heating beam. The density determines the detailed phasing of the scattered radiation relative to the O-point passage. The scattering power depends strongly nonlinearly on the heating beam power. ©2009 The American Physical Society...
Measurements of Reynolds stress and turbulence in the boundary plasma of the HT-7 tokamak
Institute of Scientific and Technical Information of China (English)
Song Mei; Wan Bao-Nian; Xu Guo-Sheng
2004-01-01
Measurements of electric field fluctuations, Reynolds stress and poloidal flow have been performed in the boundary region of the HT-7 tokamak using a Langmuir probe array. Sheared radial electric field and poloidal flow have been found in the vicinity of the limiter and the turbulence has been clearly modified in this region. Furthermore, the electrostatic Reynolds stress component shows a radial gradient close to the velocity shear layer location. All results here indicate that the radial gradient of Reynolds stress may play an important role in the driving of poloidal flows in the plasma boundary region.
Renewed first wall coating in plasma shots at the T-11M tokamak
Energy Technology Data Exchange (ETDEWEB)
Buzhinskij, O.I., E-mail: buzh@triniti.ru [Troitsk Institute for Innovation and Fusion Research, Troitsk, Moscow Reg. 142190 (Russian Federation); Barsuk, V.A.; Otroshchenko, V.G. [Troitsk Institute for Innovation and Fusion Research, Troitsk, Moscow Reg. 142190 (Russian Federation)
2010-12-15
Experimental results on boronization in situ in the tokamak T-11M plasma shots using non-toxic and not explosive metacarborane C{sub 2}H{sub 12}B{sub 10} are presented. As a result of boronization, the film with thickness up to 0.2 {mu}m at deposition rate {approx}25 nm/s was produced. The microhardness of the formed boron containing film H{sub 10} - 600, which indicates on structuredness of coating (the microhardness of the CVD B{sub 4}C films was H{sub 100} - 1800). Injection of carborane in the plasma shots has improved a stabilization of plasma filament. The impurities in wall area have been suppressed, high-vacuum characteristics of the discharge chamber were stabilized. Plasma shot duration without disruption increased essentially. At the density of n{sub e} = 1.3 x 10{sup 13} cm{sup -3}, I{sub p} = 70 kA a shot duration was 350 ms and at the density of n{sub e} = 4.65 x 10{sup 13} cm{sup -3}, I{sub p} = 70 kA was {approx}250 ms. High repeatability of experimental results has appeared. Boronization results in to an essential decrease of the volt-second consumption rate and, correspondingly, to an increase of shot duration. Developed technology opens an opportunity of practical production of renewed structured boron-carbon coatings using of plasma shots in existing large-scale tokamaks and plasma devices.
On the non-stiffness of edge transport in L-mode tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Sauter, O.; Brunner, S.; Kim, D.; Merlo, G.; Behn, R.; Coda, S.; Duval, B. P.; Federspiel, L.; Goodman, T. P.; Karpushov, A.; Merle, A.; Team, TCV, E-mail: olivier.sauter@epfl.ch [Centre de Recherches en Physique des Plasmas, Association EURATOM-Confédération Suisse, EPFL, PPB-Ecublens, 1015 Lausanne (Switzerland); Camenen, Y. [CNRS, UMR 7345, Aix-Marseille Université, Marseille (France)
2014-05-15
Transport analyses using first-principle turbulence codes and 11/2 -D transport codes usually study radial transport properties between the tokamak plasma magnetic axis and a normalized minor radius around 0.8. In this region, heat transport shows significantly stiff properties resulting in temperature scalelength values (R∕L{sub T}) that are relatively independent of the level of the radial heat flux. We have studied experimentally in the tokamak à configuration variable [F. Hofmann et al., Plasma Phys. Controlled Fusion 36, B277 (1994)] the radial electron transport properties of the edge region, close to the last closed flux surface, namely, between ρ{sub V}=√(V/V{sub edge})=0.8 to 1. It is shown that electron transport is not stiff in this region and high R∕L{sub Te} values (∼20–40) can be attained even for L-mode confinement. We can define a “pedestal” location, already in L-mode regimes, where the transport characteristics change from constant logarithmic gradient, inside ρ{sub V} = 0.8, to constant gradient between 0.8 and 1.0. In particular, we demonstrate, with well resolved T{sub e} and n{sub e} profiles, that the confinement improvement with plasma current I{sub p}, with or without auxiliary heating, is due to this non-stiff edge region. This new result is used to explain the significant confinement improvement observed with negative triangularity, which could not be explained by theory to date. Preliminary local gyrokinetic simulations are now consistent with an edge, less stiff, region that is more sensitive to triangularity than further inside. We also show that increasing the electron cyclotron heating power increases the edge temperature inverse scalelength, in contrast to the value in the main plasma region. The dependence of confinement on density in ohmic plasmas is also studied and brings new insight in the understanding of the transition between linear and saturated confinement regimes, as well as of the density limit and
Energy Technology Data Exchange (ETDEWEB)
Kenneth M. Young
2010-02-22
A Demonstration tokamak (Demo) is an essential next step toward a magnetic-fusion based reactor. One based on advanced-tokamak (AT) plasmas is especially appealing because of its relative compactness. However, it will require many plasma measurements to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.
Estimation de la diffusion thermique dans les plasmas de Tokamak
Mechhoud, Sarah; Witrant, Emmanuel; Dugard, Luc; Moreau, Didier
2012-01-01
6 pages International audience Ce travail concerne l'étude du profil de transport de la température des électrons du plasma. Une approximation numérique basée sur la méthode de Galerkin est proposée. Le coefficient de diffusion est estimé grâce à la projection spatiale qui réduit le problème à une dimension finie. Le filtre de Kalman étendu est proposé pour cette identification. Le travail est accompagné par un ensemble de simulations et de comparaisons avec les données expérimentales....
Runaway electron dynamics in tokamak plasmas with high impurity content
Martín-Solís, J. R.; Loarte, A.; Lehnen, M.
2015-09-01
The dynamics of high energy runaway electrons is analyzed for plasmas with high impurity content. It is shown that modified collision terms are required in order to account for the collisions of the relativistic runaway electrons with partially stripped impurity ions, including the effect of the collisions with free and bound electrons, as well as the scattering by the full nuclear and the electron-shielded ion charge. The effect of the impurities on the avalanche runaway growth rate is discussed. The results are applied, for illustration, to the interpretation of the runaway electron behavior during disruptions, where large amounts of impurities are expected, particularly during disruption mitigation by massive gas injection. The consequences for the electron synchrotron radiation losses and the resulting runaway electron dynamics are also analyzed.
Institute of Scientific and Technical Information of China (English)
朱大焕; 刘洋; 陈俊凌; 鄢容
2012-01-01
Thermo-mechanical simulation of the vacuum plasma spraying tungsten (VPS-W) coating on the actively cooled CuCrZr substrate under the relevant quasi-stationary heat load and transient heat flux for tokamak device, is conducted by finite element analysis (FEA). It is shown that the failure of copper softening is likely to occur at the W/Cu compliant interlayer under a typical quasi-stationary heat load and the surface failure of plastic yield damage to occur at the surface edge under a transient heat flux. In addition, the critical transient heat flux for melting is approximately 0.75 MJ/m2 for about 0.5 ms. All these results are useful for the design of the plasma facing components (PFCs) and the plasma operation in the future.
Spectra of germanium and selenium in the 50-350 A region from the PLT tokamak plasma
International Nuclear Information System (INIS)
Spectra of germanium and selenium injected into the PLT tokamak plasma were observed in the 50 to 350 A region for GeXIV-XXV (KI to OI-like) and SeXVI-XXIV (KI to NaI-like). A number of 3p/sup k/-3p/sup k-1/3d transitions predicted by isoelectronic sequence extrapolation have been identified. Also, previously identified lines from ions in the AlI to OI-like and KI-like isoelectronic sequences have been observed in the tokamak plasma
The O-X-B mode conversion scheme for ECRH of a high-density Tokamak plasma
DEFF Research Database (Denmark)
Hansen, F. R.; Lynov, Jens-Peter; Michelsen, Poul
1985-01-01
A method to apply electron cyclotron resonance heating (ECRH) to a Tokamak plasma with central density higher than the critical density for cut-off of the ordinary mode (O-mode) has been investigated. This method involves two mode conversions, from an O-mode via an extraordinary mode (X-mode) int......A method to apply electron cyclotron resonance heating (ECRH) to a Tokamak plasma with central density higher than the critical density for cut-off of the ordinary mode (O-mode) has been investigated. This method involves two mode conversions, from an O-mode via an extraordinary mode (X...
Multi-parameter gradient procedure for polarimetry data inversion in tokamak plasma
Energy Technology Data Exchange (ETDEWEB)
Chrzanowski, J., E-mail: j.chrzanowski@am.szczecin.pl [Maritime University, Szczecin Wały Chrobrego 1/2 (Poland); Kravtsov, Yu. A. [Maritime University, Szczecin Wały Chrobrego 1/2 (Poland); Mazon, D. [Association Euratom/CEA, CEA Cadarache DSM/IRFM, 13108 St. Paul lez Durance Cedex (France); JET, Culham (United Kingdom)
2013-10-15
Highlights: ► We use gradient procedure to fit plasma parameters to polarimetric data. ► Calculations are performed in developed by authors angular variables technique. ► Numerical results are compared with experimentally measured angular parameters. ► We observe satisfied accuracy of inversion procedure after several iterations. -- Abstract: Multi-parameter gradient procedure is suggested which allows fitting tokamak plasma model to polarimetric data. One of the simplest version of gradient procedure deals with four parameters model: maximum values of electron density, maximum value of electric current density in plasma, common radius of electron density, electric current distributions and increment of the safety factor inside plasma. Using recently developed by authors angular variables technique (AVT) in plasma polarimetry we may compute angular parameters of polarization ellipse for a given set of four plasma parameters and compare them with experimentally measured angular parameters. With angular parameters, measured in two channels polarimetric system (two azimuthal and two ellipticity angles, totally four experimental values). Applying then gradient procedure for squared difference between computed and measured parameters, we find four parameters of plasma model and thereby perform inversion of polarimetric data. Numerical simulations have approved that gradient procedure provides acceptable accuracy of inversion already after several iterations.
Advanced antenna system for Alfven wave plasma heating and current drive in TCABR tokamak
International Nuclear Information System (INIS)
An advanced antenna system that has been developed for investigation of Alfven wave plasma heating and current drive in the TCABR tokamak is described. The main goal was the development of such a system that could insure the excitation of travelling single helicity modes with predefined wave mode numbers M and N. The system consists of four similar modules with poloidal windings. The required spatial spectrum is formed by proper phasing of the RF feeding currents. The impedance matching of the antenna with the four-phase oscillator is accomplished by resonant circuits which form one assembly unit with the RF feeders. The characteristics of the antenna system design with respect to the antenna-plasma coupling and plasma wave excitation, for different phasing of the feeding currents, are summarised. The antenna complex impedance Z=ZR+ZI is calculated taking into account both the plasma response to resonant excitation of fast Alfven waves and the nonresonant excitation of vacuum magnetic fields in conducting shell. The matching of the RF generator with the antenna system during plasma heating is simulated numerically, modelling the plasma response with mutually coupled effective inductances with corresponding active ZR and reactive ZI impedances. The results of the numerical simulation of the RF system performance, including both the RF magnetic field spectrum analysis and the modeling of the RF generator operation with plasma load, are presented. (orig.)
Martin Yves; Duval Basil P.; Karpushov Alexander N.; Labit Benoit; Reimerdes Holger
2014-01-01
To realise the potential of fusion as an abundant energy source, several challenges remain. The TCV tokamak, featuring high shaping capability and a flexible heating system, is strongly contributing to solving these challenges. A fundamental challenge remains in controlling heat exhaust from the plasma. ITER's currently foreseen operational regime implies heat flows to the plasma facing materials that are not compatible with a commercial fusion reactor. TCV has demonstrated alternative plasma...
Klimanov, Igor; Fasoli, Ambrogio
2007-01-01
The performance of tokamaks is usually described in terms of plasma temperature, density and confinement time. The term temperature implies that the plasma is in thermal equilibrium and its particles have maxwellian (normal) velocity distribution. However, in several conditions, the plasma contains a significant number of suprathermal or 'fast' particles, whose energy is several times higher than thermal energy. The number of such particles can be significantly higher than that corresponding ...
Four-field model for tokamak plasma dynamics
International Nuclear Information System (INIS)
A generalization of reduced magnetohydrodynamics is constructed from moments of the Fokker-Planck equation. The new model uses familiar aspect-ratio approximations but allows for (1) evolution as slow as the diamagnetic drift frequency, thereby including certain finite Larmor radius effects; (2) pressure gradient terms in a generalized Ohm's law, thus making accessible the adiabatic electron limit; and (3) plasma compressibility, including the divergence of both parallel and perpendicular flows. The system is isothermal and surprisingly simple, involving only one additional field variable, i.e., four independent fields replace the three fields of reduced magnetohydrodynamics. It possesses a conserved energy. The model's equilibrium limit is shown to reproduce not only the large aspect-ratio Grad-Shafranov equation, but also such finite Larmor radius effects as the equilibrium ion parallel flow. Its linearized version reproduces, inter alia, crucial physics of the long mean-free-path electron response. Nonlinearly, the four-field model is shown to describe diffusion in stochastic magnetic fields with good qualitative accuracy
Wahlberg, C.
2009-08-01
This paper analyses low-frequency magnetohydrodynamic (MHD) modes, especially the geodesic acoustic modes (GAMs), in toroidal plasmas with large aspect ratio and circular cross section, including the effects of toroidal plasma rotation. A system of equations describing MHD modes with frequency of the order of the sound frequency in such plasmas is derived from the Frieman-Rotenberg equation, using a technique where the plasma perturbation ξ and the perturbed magnetic field Q are expanded separately in the inverse aspect ratio ɛ = r/R, where r and R denote the minor and major radii of the plasma torus, respectively. The large-scale, ideal MHD properties of the GAM induced by toroidal rotation (Wahlberg 2008 Phys. Rev. Lett. 101 115003) are thereafter analysed in more detail employing this system of equations. It is shown that both the axisymmetric GAMs existing in rotating plasmas are localized on a specific magnetic surface only to leading order in ɛ, and that a 'halo' consisting of finite components of both ξ and Q with dominant poloidal mode numbers m = ±2 appears outside this magnetic surface to higher orders in ɛ.
Modeling of the equilibrium of a tokamak plasma; Modelisation de l'equilibre d'un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Grandgirard, V
1999-12-01
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.)
Recent Advances in the Design of Quasi-axisymmetric Stellarator Plasma Configurations
Energy Technology Data Exchange (ETDEWEB)
Reiman, A.; Ku, L.; Monticello, D.; Hirschman, S.; Hudson, S.; Kessel, C. [and others
2001-01-30
Strategies for the improvement of quasi-axisymmetric stellarator configurations are explored. Calculations of equilibrium flux surfaces for candidate configurations are also presented. One optimization strategy is found to generate configurations with improved neoclassical confinement, simpler coils with lower current density, and improved flux surface quality relative to previous designs. The flux surface calculations find significant differences in the extent of islands and stochastic regions between candidate configurations. (These calculations do not incorporate the predicted beneficial effects of perturbed bootstrap currents.) A method is demonstrated for removing low-order islands from candidate configurations by relatively small modifications of the configuration. One configuration is identified as having particularly desirable properties for a proposed experiment.
International Nuclear Information System (INIS)
The Topolotron is an axisymmetric, toroidal magnetic fusion concept in which two-dimensional effects are important, as well as all three magnetic field components. The particular MHD model employed is basically the one-fluid, two-temperature model using classical Braginskii transport with viscous effects ignored. The model is augmented by Saha-Boltzmann dissociation and partial ionization physics, a simple radiation loss mechanism, and an additional resistivity due to electron-neutral collisions. While retaining all velocity and magnetic field components, the assumption of axisymmetry is made, and the resulting equations are expanded in cylindrical coordinates. The major approximation technique is then applied: spline collocation, which reduces these equations to a set of ordinary differential equations
Energy Technology Data Exchange (ETDEWEB)
Cook, G.O. Jr.
1982-12-01
The Topolotron is an axisymmetric, toroidal magnetic fusion concept in which two-dimensional effects are important, as well as all three magnetic field components. The particular MHD model employed is basically the one-fluid, two-temperature model using classical Braginskii transport with viscous effects ignored. The model is augmented by Saha-Boltzmann dissociation and partial ionization physics, a simple radiation loss mechanism, and an additional resistivity due to electron-neutral collisions. While retaining all velocity and magnetic field components, the assumption of axisymmetry is made, and the resulting equations are expanded in cylindrical coordinates. The major approximation technique is then applied: spline collocation, which reduces these equations to a set of ordinary differential equations.
New non-axisymmetric eigenmodes associated with an edge plasma layer
International Nuclear Information System (INIS)
Effects of a rarefied plasma layer surrounding a cylindrical main plasma on Alfven waves are investigated. The plasma is approximated with a two-step density profile and is assumed to be surrounded with a conducting wall. When the Alfven resonance exists inside the rarefied plasma layer, two new modes are generated. One has its maximum of the wave intensity at the wall, is thus similar to a surface wave and the other is a short of a coaxial mode. These results are re-examined in a diffuse boundary plasma and the presence of these modes is confirmed. (author)
International Nuclear Information System (INIS)
The magnetic confinement in tokamaks is for now the most advanced way towards energy production by nuclear fusion. Both theoretical and experimental studies showed that rotation generation can increase its performance by reducing the turbulent transport in tokamak plasmas. The rotation influence on the heat and particle fluxes is studied along with the angular momentum transport with the quasi-linear gyro-kinetic eigenvalue code QuaLiKiz. For this purpose, the QuaLiKiz code is modified in order to take the plasma rotation into account and compute the angular momentum flux. It is shown that QuaLiKiz framework is able to correctly predict the angular momentum flux including the E*B shear induced residual stress as well as the influence of rotation on the heat and particle fluxes. The major approximations of QuaLiKiz formalisms are reviewed, in particular the ballooning representation at its lowest order and the eigenfunctions calculated in the hydrodynamic limit. The construction of the quasi-linear fluxes is also reviewed in details and the quasi-linear angular momentum flux is derived. The different contributions to the turbulent momentum flux are studied and successfully compared both against non-linear gyro-kinetic simulations and experimental data. (author)
TEMPEST simulations of the plasma transport in a single-null tokamak geometry
Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.
2010-06-01
We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.
D-T burning plasma characteristics in an A=2 tokamak reactor
Institute of Scientific and Technical Information of China (English)
石秉仁
2005-01-01
The deuterium-tritium (D-T) burning plasma characteristic in an aspect ratio A=2 tokamak reactor is studied based on a simple equilibrium configuration, the Soloviev-type configuration. Operation limits for the Troyon beta value and for the Greenwald density value as well as for the ITER H-mode confinement scaling are used as the benchmark.It is found that in addition to suitable elongation, large triangularity has advantage for arriving at high beta value and obtaining high fusion power output. Compared to the present ITER design, the A=2 system can have very good merit for the avoidance of disruptions by setting rather high edge q value while keeping relatively large total toroidal current.The main disadvantage of decreasing the aspect ratio is due to the loss of more useful space in the inward region that leads to the decrease of toroidal magnetic field in the plasma region, then worsening the fusion merit. Our analysis and calculation also present a trade-off in this respect. Due to simple equilibrium configuration assumed, some other important issues such as the bootstrap current alignment cannot be optimized. However, the present analysis can offer an insight into the advantages of the medium aspect ratio reactor system that is a blank in present-day tokamak study.
Plasma current start-up using the lower hybrid wave on the TST-2 spherical tokamak
Takase, Y.; Ejiri, A.; Inada, T.; Moeller, C. P.; Shinya, T.; Tsujii, N.; Yajima, S.; Furui, H.; Homma, H.; Imamura, K.; Nakamura, K.; Nakamura, K.; Sonehara, M.; Takeuchi, T.; Togashi, H.; Tsuda, S.; Yoshida, Y.
2015-12-01
Non-inductive plasma current start-up, ramp-up and sustainment by waves in the lower hybrid wave (LHW) frequency range at 200 MHz were investigated on the TST-2 spherical tokamak (R0 ≤ 0.38 m, a ≤ 0.25 m, Bt0 ≤ 0.3T, Ip ≤ 0.14 MA). Experimental results obtained using three types of antenna were compared. Both the highest plasma current (Ip = 18 kA) and the highest current drive figure of merit ηCD≡n¯eIpR0/PRF=1.4 ×1017 A/W/m2 were achieved using the capacitively-coupled combline (CCC) antenna, designed to excite the LHW with a sharp and highly directional wavenumber spectrum. For Ip greater than about 5 kA, high energy electrons accelerated by the LHW become the dominant carrier of plasma current. The low value of ηCD observed so far are believed to be caused by a rapid loss of energetic electrons and parasitic losses of the LHW energy in the plasma periphery. ηCD is expected to improve by an order of magnitude by increasing the plasma current to improve energetic electron confinement. In addition, edge power losses are expected to be reduced by increasing the toroidal magnetic field to improve wave accessibility to the plasma core, and by launching the LHW from the inboard upper region of the torus to achieve better single-pass absorption.
International Nuclear Information System (INIS)
Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 1014 cm-3 and 2.5 x 1014 cm-3. Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments
Influence of plasma opacity on current decay after disruptions in tokamaks
International Nuclear Information System (INIS)
Current decays after disruptions as well as after noble gas injections in tokamaks are examined. As is shown, the cooled plasmas at the stage of current decay are partially opaque for radiation in lines giving the main impact into total thermal losses. The thermal balance is supposed to be determined by Ohmic heating and radiative losses. A zero-dimensional model for radiation losses and temperature distribution over minor radius is used. Plasma current evolution is simulated with DIMRUN and DINA codes. Impurity distribution over ionization states is calculated from the time-dependent set of differential equations. The opacity effects are found to be most important for simulation of JET disruption experiments with beryllium- and carbon-seeded plasmas. The decay times calculated are in good agreement with the experimental values. Current decays in beryllium-, carbon-, argon- and neon-seeded plasmas for ITER parameters are simulated. The temperatures after thermal quench are shown to be significantly higher in comparison with the model of transparent plasmas. Opacity effects are found to be most important for Be- and C-seeded plasmas. Runaway electron currents are damped significantly if opacity effects are taken into account in any case examined
Noori, Ehsanallah; Sadeghi, Yahya; Ghoranneviss, Mahmood
2016-10-01
In this study, magnetic measurement of poloidal fields were used to determine poloidal beta and plasma internal inductance of Damavand tokamak combination of poloidal beta and plasma internal inductance (β _p+{l_i}/{2} ), known as Shafranov parameter, was obtained experimentally in terms of normal and tangential components of the magnetic field. Plasma internal inductance and poloidal beta were obtained using parametrization method based on analytical solution of Grad-Shafranov equation (GSE) and compared with parabolic-like profile of toroidal current density approach for determination of the plasma internal inductance. Finding evolution of β _p+{l_i}/{2} and plasma internal inductance. Finding poloidal beta (Shafranov parameter and internal inductance) and using energy balance equation, thermal energy and energy confinement were determined qualitatively in terms of poloidal beta during a regular discharge of Damavand tokamak.
Tokamak edge plasma rotation in the presence of the biased electrode
Energy Technology Data Exchange (ETDEWEB)
Ghoranneviss, M.; Mohammadi, S. [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Elahi, A. Salar, E-mail: Salari_phy@yahoo.com [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Arvin, R. [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of)
2013-02-15
Electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then biasing experiments were carried out. Also, using a Mach probes the effects of radial electric field (produced by biased electrode) on the poloidal and toroidal components of the edge plasma velocity were investigated. The results showed an increase in both toroidal and poloidal components of the edge plasma velocity during biasing regime. Results compared and discussed. During positive biasing, increased E{sub r} tends to slow the poloidal rotation in the electron diamagnetic drift direction, i.e., to speed up rotation in the ion diamagnetic drift direction. An increased toroidal rotation velocity has the opposite effect on the poloidal rotation.
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Energy Technology Data Exchange (ETDEWEB)
Wang, W. X. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ethier, S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ren, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kaye, S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Chen, J. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Startsev, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lu, Z. [Univ. of California, San Diego, CA (United States). La Jolla, CA
2015-10-30
New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong ExB shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offering one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. This predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.
Generation of a magnetic island by edge turbulence in tokamak plasmas
Poyé, A.; Agullo, O.; Muraglia, M.; Garbet, X.; Benkadda, S.; Sen, A.; Dubuit, N.
2015-03-01
We investigate, through extensive 3D magneto-hydro-dynamics numerical simulations, the nonlinear excitation of a large scale magnetic island and its dynamical properties due to the presence of small-scale turbulence. Turbulence is induced by a steep pressure gradient in the edge region [B. D. Scott, Plasma Phys. Controlled Fusion 49, S25 (2007)], close to the separatrix in tokamaks where there is an X-point magnetic configuration. We find that quasi-resonant localized interchange modes at the plasma edge can beat together and produce extended modes that transfer energy to the lowest order resonant surface in an inner stable zone and induce a seed magnetic island. The island width displays high frequency fluctuations that are associated with the fluctuating nature of the energy transfer process from the turbulence, while its mean size is controlled by the magnetic energy content of the turbulence.
Generation of a magnetic island by edge turbulence in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Poyé, A. [Aix-Marseille Université, CNRS, PIIM, UMR 7345, Marseille (France); Université de Bordeaux, CELIA Laboratory, Talence 33405 (France); Agullo, O.; Muraglia, M.; Benkadda, S.; Dubuit, N. [Aix-Marseille Université, CNRS, PIIM, UMR 7345, Marseille (France); France-Japan Magnetic Fusion Laboratory, LIA 336 CNRS, Marseille (France); Garbet, X. [IRFM, CEA, St-Paul-Lez-Durance 13108 (France); Sen, A. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)
2015-03-15
We investigate, through extensive 3D magneto-hydro-dynamics numerical simulations, the nonlinear excitation of a large scale magnetic island and its dynamical properties due to the presence of small-scale turbulence. Turbulence is induced by a steep pressure gradient in the edge region [B. D. Scott, Plasma Phys. Controlled Fusion 49, S25 (2007)], close to the separatrix in tokamaks where there is an X-point magnetic configuration. We find that quasi-resonant localized interchange modes at the plasma edge can beat together and produce extended modes that transfer energy to the lowest order resonant surface in an inner stable zone and induce a seed magnetic island. The island width displays high frequency fluctuations that are associated with the fluctuating nature of the energy transfer process from the turbulence, while its mean size is controlled by the magnetic energy content of the turbulence.
Present and perspective roles of soft X-ray tomography in tokamak plasma position measurements
International Nuclear Information System (INIS)
This paper shows the importance and feasibility of real-time tomography in fusion experiments for the example of soft X-ray (SXR) position measurements. The requirement of non-magnetic real-time diagnostics in low frequencies for ITER is discussed. This is illustrated by recent results of rapid tomographic inversion of SXR measurements on tokamak TCV. Comparison with the magnetic reconstruction data has not only shown the valuable resolution capabilities of both techniques, but also revealed a slight dependence of magnetic measurements on toroidal magnetic field and an unnoticed drift of plasma position observer. A feasibility study using current hardware capacities for programmable real-time tomographic system with plasma position feedback output was carried out. A compact solution is found to be tractable opening wide perspectives for development
Nonlinear Transport Processes in Tokamak Plasmas. Part I: The Collisional Regimes
Sonnino, Giorgio
2008-01-01
An application of the thermodynamic field theory (TFT) to transport processes in L-mode tokamak plasmas is presented. The nonlinear corrections to the linear (Onsager) transport coefficients in the collisional regimes are derived. A quite encouraging result is the appearance of an asymmetry between the Pfirsch-Schlueter (P-S) ion and electron transport coefficients: the latter presents a nonlinear correction, which is absent for the ions, and makes the radial electron coefficients much larger than the former. Explicit calculations and comparisons between the neoclassical results and the TFT predictions for JET plasmas are also reported. We found that the nonlinear electron P-S transport coefficients exceed the values provided by neoclassical theory by a factor, which may be of the order 100. The nonlinear classical coefficients exceed the neoclassical ones by a factor, which may be of order 2. The expressions of the ion transport coefficients, determined by the neoclassical theory in these two regimes, remain...
Multidirectional plasma flow measurement by Gundestrup Probe in scrape-off layer of ADITYA tokamak
Energy Technology Data Exchange (ETDEWEB)
Sangwan, Deepak; Jha, Ratneshwar; Tanna, Rakesh L. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India)
2015-11-15
Multidirectional plasma flow measurements by using Gundestrup Probe in the scrape-off layer of ADITYA tokamak are presented. The ADITYA Gundestrup Probe-head consists of eight plates arranged around the ceramic rod and three pins normal to side plates. Plates are used to measure both parallel and perpendicular flows simultaneously and pins are used to measure plasma density and floating potential. A comparison of direct perpendicular flow measurement and by two other plates of Gundestrup Probe is presented. Possible causes of perpendicular flows are identified and compared with the measured flows. It is observed that the mechanism of the parallel flow and the perpendicular flow is different only at high parallel Mach number. A puff of the working gas is used to study its effect on the perpendicular flows and its reversal with the gas puff is observed.
Shielding of External Magnetic Perturbations By Torque In Rotating Tokamak Plasmas
Energy Technology Data Exchange (ETDEWEB)
Park, Jong-Kyu; Boozer, Allen H.; Menard, Jonathan E.; Gerhardt, Stefan P.; Sabbagh, Steve A.
2009-08-24
The imposition of a nonaxisymmetric magnetic perturbation on a rotating tokamak plasma requires energy and toroidal torque. Fundamental electrodynamics implies that the torque is essentially limited and must be consistent with the external response of a plasma equilibrium ƒ = j x B. Here magnetic measurements on National Spherical Torus eXperiment (NSTX) device are used to derive the energy and the torque, and these empirical evaluations are compared with theoretical calculations based on perturbed scalar pressure equilibria ƒ = ∇p coupled with the theory of nonambipolar transport. The measurement and the theory are consistent within acceptable uncertainties, but can be largely inconsistent when the torque is comparable to the energy. This is expected since the currents associated with the torque are ignored in scalar pressure equilibria, but these currents tend to shield the perturbation.
Measurements of Boundary Plasma in Synergy Discharges of IBW and LHCD on the HT-7 Tokamak
Institute of Scientific and Technical Information of China (English)
宋梅; 万宝年; 徐国盛; 陈忠勇; 刘海庆; 凌必利; 李成富
2003-01-01
By applying ion Bernstein wave (IBW) heating into the lower hybrid current drive (LHCD) plasma, improyed confinements have been obtained in the HT-7 tokamak. The central electron temperaturewas doubled and the storage energy was increased significantly. The core electron density and temperature were broadened and their profiles near the edge were steepened. A transport barrier has been formed in the vicinity of the limiter radial location. An enhanced shear in poloidal phase velocity was found in the same region with reduction of the fluctuation levels and the coherences between fluctuations. The results suggest that the improved confinement in the IBW and LHCD plasma is at least partially due to the modification of shear in poloidal velocity and then the suppression of fluctuations and fluctuation induced fluxes via de-correlation effect.
Fokker--Planck/transport analyses of fusion plasmas in contemporary beam-driven tokamaks
Energy Technology Data Exchange (ETDEWEB)
Mirin, A.A.; McCoy, M.G.; Killeen, J.; Rensink, M.E.; Shumaker, D.E.; Jassby, D.L.; Post, D.E.
1978-04-01
The properties of deuterium plasmas in experimental tokamaks heated and fueled by intense neutral-beam injection are evaluated with a Fokker-Planck/radial transport code coupled with a Monte Carlo neutrals treatment. Illustrative results are presented for the Poloidal Divertor Experiment at PPPL as a function of beam power and plasma recycling coefficient, R/sub c/. When P/sub beam/ = 8 MW at E/sub b/ = 60 keV, and R/sub c/ = 0.2, then
International Nuclear Information System (INIS)
Tokamak disruption simulation experiments have been conducted at the University of New Mexico using the PLADIS I plasma gun system. Earlier work had characterized the plasma-surface interaction in terms of parameters such as incident energy from bucket calorimeter measurements and rough measurements of beam area from flat damage targets. A variety of new plasma diagnostics have been used to further investigate the characteristics of the incident plasma beam and vapor shield plasma in a simulated tokamak disruption. These diagnostics have included laser interferometry, two-color pyrometry, emission spectroscopy, and other methods to quantify the characteristics of the incident and vapor shield plasmas of a simulated tokamak disruption. The synthesis of different beam area measurement techniques is used to determine the radial structure of the plasma beam. Vacuum ultra violet spectroscopy is used to determine the thickness and internal structure of the vapor shield plasma. Results from two-color optical pyrometry and surface pressure measurements are used to determine the dynamics of vapor shield formation
Tokamak D-T neutron source models for different plasma physics confinement modes
Energy Technology Data Exchange (ETDEWEB)
Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Puma, Antonella Li; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)
2012-08-15
Highlights: Black-Right-Pointing-Pointer HCLL DEMO neutronics is based on plasma physics L-mode, but may use H or A mode. Black-Right-Pointing-Pointer Based on Plasma Physics 0D code, H and A-mode D-T neutron sources formulae are proposed. Black-Right-Pointing-Pointer TRANSGEN code is built to create 2D source maps as input for Monte-Carlo codes. Black-Right-Pointing-Pointer A-mode neutronic impact is compared to L-mode at same power on a HCLL DEMO design. Black-Right-Pointing-Pointer Results show TBR and Me slight changes, contrary to NWL profile: from -22% to +11%. - Abstract: Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H and A-mode). Based on analytical formulae used in plasma physics, H and A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium-lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from -22% to +11% for NWL.
Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas
Ribeiro, Celso
2015-11-01
The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.
Spectral analysis of turbulence propagation mechanisms in solar wind and tokamaks plasmas
International Nuclear Information System (INIS)
This thesis takes part in the study of spectral transfers in the turbulence of magnetized plasmas. We will be interested in turbulence in solar wind and tokamaks. Spacecraft measures, first principle simulations and simple dynamical systems will be used to understand the mechanisms behind spectral anisotropy and spectral transfers in these plasmas. The first part of this manuscript will introduce the common context of solar wind and tokamaks, what is specific to each of them and present some notions needed to understand the work presented here. The second part deals with turbulence in the solar wind. We will present first an observational study on the spectral variability of solar wind turbulence. Starting from the study of Grappin et al. (1990, 1991) on Helios mission data, we bring a new analysis taking into account a correct evaluation of large scale spectral break, provided by the higher frequency data of the Wind mission. This considerably modifies the result on the spectral index distribution of the magnetic and kinetic energy. A second observational study is presented on solar wind turbulence anisotropy using autocorrelation functions. Following the work of Matthaeus et al. (1990); Dasso et al. (2005), we bring a new insight on this statistical, in particular the question of normalisation choices used to build the autocorrelation function, and its consequence on the measured anisotropy. This allows us to bring a new element in the debate on the measured anisotropy depending on the choice of the referential either based on local or global mean magnetic field. Finally, we study for the first time in 3D the effects of the transverse expansion of solar wind on its turbulence. This work is based on a theoretical and numerical scheme developed by Grappin et al. (1993); Grappin and Velli (1996), but never used in 3D. Our main results deal with the evolution of spectral and polarization anisotropy due to the competition between non-linear and linear (Alfven coupling
Energy confinement and transport of H-mode plasmas in tokamak
International Nuclear Information System (INIS)
A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each
Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas
Horacek, J.; Pitts, R. A.; Adamek, J.; Arnoux, G.; Bak, J.-G.; Brezinsek, S.; Dimitrova, M.; Goldston, R. J.; Gunn, J. P.; Havlicek, J.; Hong, S.-H.; Janky, F.; LaBombard, B.; Marsen, S.; Maddaluno, G.; Nie, L.; Pericoli, V.; Popov, Tsv; Panek, R.; Rudakov, D.; Seidl, J.; Seo, D. S.; Shimada, M.; Silva, C.; Stangeby, P. C.; Viola, B.; Vondracek, P.; Wang, H.; Xu, G. S.; Xu, Y.; Contributors, JET
2016-07-01
As in many of today’s tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, {{q}||} in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as {{q}||}={{q}0}\\text{exp} ~≤ft(-r/λ q\\text{omp}\\right) , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, λ q\\text{omp} . The initial choice of λ q\\text{omp} , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with R=\\text{0}\\text{.4--2}\\text{.8} \\text{m}, {{B}0}=\\text{1}\\text{.2--7}\\text{.5} \\text{T}, {{I}\\text{p}}=\\text{9--2500} \\text{kA}. Measurements of λ q\\text{omp} in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar
Institute of Scientific and Technical Information of China (English)
SHI Bing-Ren; LI Ji-Quan
2007-01-01
An exact ballooning mode eigen-equation is derived to study stability of axi-symmetric toroidal plasma with arbitrary aspect ratio, including the tokamak, the finite aspect ratio and the spherical torus plasmas. For comparison with the widely used ( s - α) model, an analytic exact equilibrium configuration with circular magnetic surfaces is analysed in detail. It is indicated that the (s - α) model needs to be improved for more realistic configurations.
An optimal real-time controller for vertical plasma stabilization
Cruz, N; Coda, S; Duval, B P; Le, H B; Rodrigues, A P; Varandas, C A F; Correia, C M B A; Goncalves, B S
2014-01-01
Modern Tokamaks have evolved from the initial axisymmetric circular plasma shape to an elongated axisymmetric plasma shape that improves the energy confinement time and the triple product, which is a generally used figure of merit for the conditions needed for fusion reactor performance. However, the elongated plasma cross section introduces a vertical instability that demands a real-time feedback control loop to stabilize the plasma vertical position and velocity. At the Tokamak \\`a Configuration Variable (TCV) in-vessel poloidal field coils driven by fast switching power supplies are used to stabilize highly elongated plasmas. TCV plasma experiments have used a PID algorithm based controller to correct the plasma vertical position. In late 2013 experiments a new optimal real-time controller was tested improving the stability of the plasma. This contribution describes the new optimal real-time controller developed. The choice of the model that describes the plasma response to the actuators is discussed. The ...
Application of reversed ultra-fast theta-pinch to a tokamak plasma
International Nuclear Information System (INIS)
A reversed ultra-fast theta-pinch using a Blumlein circuit is applied to an ordinary tokamak plasma to provide of further heating. A large amplitude magnetic wave is excited and propagates radially into the plasma and is expected to heat it turbulently by its passage, the plasma cross-section remaining unchanged. A preliminary experiment in a straight machine showed some interesting results. A small, fat torus has been constructed whose major diameter is 50 cm and minor is 24 cm. The conducting shell also serves as the thetatron coil. The toroidal magnetic field is, 7 kG and the plasma current of 30 kA flows for about 10 ms. The Blumlein line charged to 100 kV is closed at the time of the plasma current maximum. The induced azimuthal electric field is estimated to be 0.4 kV/cm along the tube wall in the case of vacuum. The rise time of the magnetic field is 200 ns. (author)
Energy Technology Data Exchange (ETDEWEB)
Martin, R.; Manero, F.
1984-07-01
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.
Numerical computation of gravitational field for general axisymmetric objects
Fukushima, Toshio
2016-10-01
We developed a numerical method to compute the gravitational field of a general axisymmetric object. The method (i) numerically evaluates a double integral of the ring potential by the split quadrature method using the double exponential rules, and (ii) derives the acceleration vector by numerically differentiating the numerically integrated potential by Ridder's algorithm. Numerical comparison with the analytical solutions for a finite uniform spheroid and an infinitely extended object of the Miyamoto-Nagai density distribution confirmed the 13- and 11-digit accuracy of the potential and the acceleration vector computed by the method, respectively. By using the method, we present the gravitational potential contour map and/or the rotation curve of various axisymmetric objects: (i) finite uniform objects covering rhombic spindles and circular toroids, (ii) infinitely extended spheroids including Sérsic and Navarro-Frenk-White spheroids, and (iii) other axisymmetric objects such as an X/peanut-shaped object like NGC 128, a power-law disc with a central hole like the protoplanetary disc of TW Hya, and a tear-drop-shaped toroid like an axisymmetric equilibrium solution of plasma charge distribution in an International Thermonuclear Experimental Reactor-like tokamak. The method is directly applicable to the electrostatic field and will be easily extended for the magnetostatic field. The FORTRAN 90 programs of the new method and some test results are electronically available.
Study of the plasma SOL with fast reciprocating probe diagnostics on the SST-1 tokamak
International Nuclear Information System (INIS)
A reciprocating probe drive system has been designed, fabricated and successfully installed at the bottom port of Steady State Superconducting Tokamak (SST-1). The probe system has been designed to measure the local plasma parameters such as temperature (1 eV to 50 eV range), density (up to ∼ 1018 m-3) and floating potential (∼100V) near the lower X-point of the plasma column at the plasma current flat top. The probe head can move a total distance of 390 mm from its reference position during plasma shot with a combination of two pneumatic cylinders (slow and fast) and edge welded bellows. Slow movement is achieved from rest position to reference position (200mm) in 2sec. From the reference position, the fast movement over 190 mm of length is made in 300 ms. A programmable logic controlling (PLC) system records the number of scan and delay with reference to loop voltage. Timing between the scans is synchronized with the of SST-1 control system sequence. The density at 370 mm below the mid plane is measured to be 0.3-1 X 1011 cm-3 at a bias voltage of - 70 V. Interaction of the plasma with the probe tip and the probe movement during a plasma shot can be traced with the fast visible imaging in SST-1. The measured density and probe-plasma interaction will be correlated with the radiated power measured using bolometer diagnostics. Density fluctuations and radial electric field at the scrape off layer (SOL) and their implications on plasma performance will be reported. Further, these signals will also serve as an input for power balance studies. (author)
Nonlinear evolution of multi-helicity neo-classical tearing modes in rotating tokamak plasmas
Wei, Lai; Wang, Zheng-Xiong; Wang, Jialei; Yang, Xuefeng
2016-10-01
Plasma perturbations from the core and/or boundary regions of tokamaks can provide seed islands for the excitation of neo-classical tearing modes (NTMs) with negative {{ Δ }\\prime} , where {{ Δ }\\prime} is the linear instability parameter of the classical tearing mode. In this work, by means of reduced magnetohydrodynamic simulations, we numerically investigate the nonlinear evolution of multi-helicity NTMs in rotating tokamak plasmas with these two types of plasma perturbations with different boundary conditions. In the first case of initial plasma perturbations from the core region with a zero boundary condition, the meta-stable property of seed-island triggered NTM with negative {{ Δ }\\prime} is verified in the single helicity simulation. Nevertheless in the multiple helicity simulation, this seed-island triggered NTM with negative {{ Δ }\\prime} can be suppressed by a spontaneous NTM with positive {{ Δ }\\prime} through the competitive interaction between NTMs with different helicities. If a fixed poloidal rotation is taken into account in the first case, two different helicity NTMs could coexist in the saturation stage, which is different qualitatively from the process without plasma rotation. In the second case of initial plasma perturbations from the boundary region with a nonzero boundary condition, as the amplitude of plasma perturbations on the boundary increases, the mode with negative {{ Δ }\\prime} gradually changes from the driven-reconnection state to the NTM state, accompanied by an enhancement of magnetic island width in the single helicity simulation. Nevertheless in the multi-helicity simulation, the spontaneous NTM with positive {{ Δ }\\prime} can make the driven-reconnection triggered NTM with negative {{ Δ }\\prime} transfer from the NTM state back to the driven-reconnection state again. The underlying mechanism behind these transitions is analyzed step by step. Effects of fixed and unfixed poloidal rotations on the nonlinear
Characterization of Plasma Gun with TiH2/C60 Cartridge for Disruption Mitigation in Tokamaks
Bogatu, I. N.; Thompson, J. R.; Galkin, S. A.; Kim, J. S.; HyperV Technologies Corp. Team
2011-10-01
Impurity injection for disruption mitigation in tokamaks must be faster than growth time of plasma instabilities, requires sufficient mass to get critical electron density, high penetrability, and large assimilation fraction in the core plasma, with rapid impurity redistribution over the whole plasma. FAR-TECH, Inc. proposed the innovative idea to use hyper-velocity (>30 km/s), high-density (>1023 m-3) C60/C plasma jets with high ram pressure to deliver the impurity mass in 30 mg of C60 gas in gun (~35 cm length) prototype with TiH2/C60 cartridge for a small scale, proof-of-principle experiment on a tokamak. Work supported by the US DOE DE-FG02-08ER85196 grant.
Coherence imaging and tomography of fields and flows in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Howard, J.; Diallo, A.; Creese, M.; Blackwell, B.C. [Australian National Universityj, Canberra (Australia); Jaspers, R. [Eindhoven University of Technology, Eindhoven (Netherlands); Chung, J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Allen, S.L.; Meyer, W.; Fenstermacher, M.E.; Porter, G.D.; Ellis, R.M. [Lawrence Livermore National Laboratory at General Atomics, San Diego (United States); Van Zeeland, M.E.; Boivin, R.L.; Brooks, N. [General Atomics, San Diego (United States)
2011-07-01
In the last few years we have developed various spatial heterodyne polarization interferometers for spectrally-resolved optical imaging of edge and core parameters in high temperature magnetized plasmas. Applications include imaging motional Stark effect and Zeeman effect polarimetry for determination of the magnetic field pitch angle, Thomson scattering, and passive and active (charge exchange recombination spectroscopy - CXRS) Doppler imaging of plasma temperature and flow. In this paper we summarize recent innovations in imaging instrumentation and will present first results of motional Stark effect imaging of the internal magnetic field on the TEXTOR tokamak and Doppler flow imaging in the H-1 heliac and DIII-D divertor. The TEXTOR instrument uses a hybrid spatio-temporal multiplexing approach to capture 2 dimensional images of the projected beam velocity and magnetic field vector fields. While the Doppler projection agrees very well with modeling, there are some discrepancies in the polarimetric image which appear to be related to imperfections in the optical coupling prism. This issue will be addressed during a new set of measurements commencing in April 2010. During 2009 we installed instruments for imaging flows in the divertor and scrape-off-layer in the DIII-D tokamak. In these experiments, single snapshot interferometric images of the plasma in CII 514 nm, and CIII 465 nm emission have been demodulated to obtain flow and ion temperature projections. Tomographic reconstructions of the flow fields show encouraging agreement with UEDGE modeling, pointing the way towards experiments that address important divertor transport issues in future. This document is composed of an abstract followed by the slides of the presentation. (authors)
Porkolab, M.; Lloyd, B.; Takase, Y.; Bonoli, P.; Fiore, C.; Gandy, R.; Granetz, R.; Griffin, D.; Gwinn, D.; Lipschultz, B.; Marmar, E.; McCool, S.; Pachtman, A.; Pappas, D.; Parker, R.; Pribyl, P.; Rice, J.; Terry, J.; Texter, S.; Watterson, R.; Wolfe, S.
1984-09-01
The effectiveness of plasma heating by electron Landau interaction in the lower hybrid range of frequencies in tokamak plasmas is demonstrated. Upon injection of 850 kW of rf power at a density of n―e~=1.4×1014 cm-3, an electron temperature increase of 1.0 keV and an ion temperature increase of 0.8 keV was achieved. These results are compared with transport and ray-tracing code predictions.
Development of real-time plasma analysis and control algorithms for the TCV tokamak using SIMULINK
Energy Technology Data Exchange (ETDEWEB)
Felici, F., E-mail: f.felici@tue.nl [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Suisse, 1015 Lausanne (Switzerland); Eindhoven University of Technology, Department of Mechanical Engineering, Control Systems Technology Group, P.O. Box 513, 5600MB Eindhoven (Netherlands); Le, H.B.; Paley, J.I.; Duval, B.P.; Coda, S.; Moret, J.-M.; Bortolon, A.; Federspiel, L.; Goodman, T.P. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Suisse, 1015 Lausanne (Switzerland); Hommen, G. [FOM-Institute DIFFER, Association EURATOM-FOM, Nieuwegein (Netherlands); Eindhoven University of Technology, Department of Mechanical Engineering, Control Systems Technology Group, P.O. Box 513, 5600MB Eindhoven (Netherlands); Karpushov, A.; Piras, F.; Pitzschke, A. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Suisse, 1015 Lausanne (Switzerland); Romero, J. [National Laboratory of Fusion, EURATOM-CIEMAT, Madrid (Spain); Sevillano, G. [Department of Automatic Control and Systems Engineering, Bilbao University of the Basque Country, Bilbao (Spain); Sauter, O.; Vijvers, W. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Suisse, 1015 Lausanne (Switzerland)
2014-03-15
Highlights: • A new digital control system for the TCV tokamak has been commissioned. • The system is entirely programmable by SIMULINK, allowing rapid algorithm development. • Different control system nodes can run different algorithms at varying sampling times. • The previous control system functions have been emulated and improved. • New capabilities include MHD control, profile control, equilibrium reconstruction. - Abstract: One of the key features of the new digital plasma control system installed on the TCV tokamak is the possibility to rapidly design, test and deploy real-time algorithms. With this flexibility the new control system has been used for a large number of new experiments which exploit TCV's powerful actuators consisting of 16 individually controllable poloidal field coils and 7 real-time steerable electron cyclotron (EC) launchers. The system has been used for various applications, ranging from event-based real-time MHD control to real-time current diffusion simulations. These advances have propelled real-time control to one of the cornerstones of the TCV experimental program. Use of the SIMULINK graphical programming language to directly program the control system has greatly facilitated algorithm development and allowed a multitude of different algorithms to be deployed in a short time. This paper will give an overview of the developed algorithms and their application in physics experiments.
Theoretical and numerical studies of wave-packet propagation in tokamak plasmas
Lu, Z X; Cardinali, A
2011-01-01
Theoretical and numerical studies of wave-packet propagation are presented to analyze the time varying 2D mode structures of electrostatic fluctuations in tokamak plasmas, using general flux coordinates. Instead of solving the 2D wave equations directly, the solution of the initial value problem is used to obtain the 2D mode structure, following the propagation of wave-packets generated by a source and reconstructing the time varying field. As application, the 2D WKB method is applied to investigate the shaping effects (elongation and triangularity) of tokamak geometry on the lower hybrid wave propagation and absorbtion. Meanwhile, the Mode Structure Decomposition (MSD) method is used to handle the boundary conditions and simplify the 2D problem to two nested 1D problems. The MSD method is related to that discussed earlier by Zonca and Chen [Phys. Fluids B 5, 3668 (1993)], and reduces to the well-known "ballooning formalism" [J. W. Connor, R. J. Hastie, and J. B. Taylor, Phys. Rev. Lett. 40, 396 (1978)], when...
Data Acquisition and Automation for Plasma Rotation Diagnostic in the TCABR Tokamak
Ronchi, G.; Severo, J. H. F.; de Sá, W. P.; Galvão, R. M. O.
2015-03-01
In this work we describe the implementation of a full modular system of data acquisition and processing for the plasma rotation diagnostic in the TCABR tokamak. The experimental setup uses a single monochromator and six photomultipliers (PMT), in which pair of PMTs measures the light at slightly different wavelengths. Thus, it can measure the time evolution of the Doppler shift of the impurities emission lines coming from three spatial positions (one for toroidal rotation and two for poloidal rotation). The data acquisition and preanalysis program were written with LabVIEW software and is capable of controlling the spectrometer wavelength, PMTs power supplies, data acquisition, and storage. All data are recorded in MDSplus trees that easily allow data visualization and post-processing analysis (both locally and remotely) via MATLAB, Python, Java and others programming languages. This system can run independently from other diagnostics and machine systems and can be integrated with the main tokamak control system by means of TCP/IP messages.
Energy Technology Data Exchange (ETDEWEB)
B.C. Lyons, S.C. Jardin, and J.J. Ramos
2012-06-28
A new code, the Neoclassical Ion-Electron Solver (NIES), has been written to solve for stationary, axisymmetric distribution functions (f ) in the conventional banana regime for both ions and elec trons using a set of drift-kinetic equations (DKEs) with linearized Fokker-Planck-Landau collision operators. Solvability conditions on the DKEs determine the relevant non-adiabatic pieces of f (called h ). We work in a 4D phase space in which Ψ defines a flux surface, θ is the poloidal angle, v is the total velocity referenced to the mean flow velocity, and λ is the dimensionless magnetic moment parameter. We expand h in finite elements in both v and λ . The Rosenbluth potentials, φ and ψ, which define the integral part of the collision operator, are expanded in Legendre series in cos χ , where χ is the pitch angle, Fourier series in cos θ , and finite elements in v . At each ψ , we solve a block tridiagonal system for hi (independent of fe ), then solve another block tridiagonal system for he (dependent on fi ). We demonstrate that such a formulation can be accurately and efficiently solved. NIES is coupled to the MHD equilibrium code JSOLVER [J. DeLucia, et al., J. Comput. Phys. 37 , pp 183-204 (1980).] allowing us to work with realistic magnetic geometries. The bootstrap current is calculated as a simple moment of the distribution function. Results are benchmarked against the Sauter analytic formulas and can be used as a kinetic closure for an MHD code (e.g., M3D-C1 [S.C. Jardin, et al ., Computational Science & Discovery, 4 (2012).]).
Energy Technology Data Exchange (ETDEWEB)
NONE
1994-05-27
If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).
International Nuclear Information System (INIS)
Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks
International Nuclear Information System (INIS)
Over the last few years is remarkable, so increasingly evident the need for a new source of energy for mankind. One promising option is through nuclear fusion, where the plasma produced in the reactor can be converted into electrical energy. Therefore, knowing the characteristics of this plasma is very important to control it and understand it so desirable. One of the diagnostic options is called Thomson scattering . This is considered the most reliable method for the determination of important plasma parameters such as temperature and electron density, and may also help in the study and explanation of various internal mechanisms. The great advantage lies in the tact that they consist of a direct measurement and nonperturbative. But it is a diagnosis whose installation and execution is admittedly complex, limiting it only a few laboratories in the fíeld of fusion for the world. Among the main difficulties, wc can highlight the fact that the scattered signal is very small, thus requiring a large increase of the incident power. Moreover, the external physical conditions can cause mechanical vibrations that eliminate or minimize them as much as possible, is a great challenge, considering the optical micrometrically very sensitive and needs involved in the system. This work describes the entire process of installation and operation of Thomson scattering diagnostic in tokamak TCABR and through this diagnosis, we work on results of electron temperature, to finally be able to calculate the electron density of the plasma. (author)
Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape
Hommen, G.; de M. Baar,; Nuij, P.; McArdle, G.; Akers, R.; Steinbuch, M.
2010-01-01
A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma b
International Nuclear Information System (INIS)
Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality ν*e>4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera (Δt=20 μs) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density ne and temperature Te with high spatial (Δr=2 mm) and temporal resolution (Δt=20 μs). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke νRMP vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss along open magnetic field lines to the wall components. For high relative
Energy Technology Data Exchange (ETDEWEB)
Stoschus, Henning
2011-10-13
Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality {nu}{sup *}{sub e}>4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera ({delta}t=20 {mu}s) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density n{sub e} and temperature T{sub e} with high spatial ({delta}r=2 mm) and temporal resolution ({delta}t=20 {mu}s). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke {nu}{sub RMP} vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss
Impurities in a non-axisymmetric plasma: transport and effect on bootstrap current
Mollén, Albert; Smith, Håkan M; Braun, Stefanie; Helander, Per
2015-01-01
Impurities cause radiation losses and plasma dilution, and in stellarator plasmas the neoclassical ambipolar radial electric field is often unfavorable for avoiding strong impurity peaking. In this work we use a new continuum drift-kinetic solver, the SFINCS code (the Stellarator Fokker-Planck Iterative Neoclassical Conservative Solver) [M. Landreman et al., Phys. Plasmas 21 (2014) 042503] which employs the full linearized Fokker-Planck operator, to calculate neoclassical impurity transport coefficients for a Wendelstein 7-X (W7-X) magnetic configuration. We compare SFINCS calculations with theoretical asymptotes in the high collisionality limit. In intermediate and high collisionality regimes, a momentum conserving collision operator is critical to correctly determine the impurity transport coefficients, and a simple pitch-angle scattering approximation can lead to transport predictions in the wrong direction. In the low collisionality regime pitch-angle scattering is sufficient to accurately describe impuri...
Zhong, W. L.; Shen, Y.; Zou, X. L.; Gao, J. M.; Shi, Z. B.; Dong, J. Q.; Duan, X. R.; Xu, M.; Cui, Z. Y.; Li, Y. G.; Ji, X. Q.; Yu, D. L.; Cheng, J.; Xiao, G. L.; Jiang, M.; Yang, Z. C.; Zhang, B. Y.; Shi, P. W.; Liu, Z. T.; Song, X. M.; Ding, X. T.; Liu, Yong; HL-2A Team
2016-07-01
The impact of impurity ions on a pedestal has been investigated in the HL-2A Tokamak, at the Southwestern Institute of Physics, Chengdu, China. Experimental results have clearly shown that during the H -mode phase, an electromagnetic turbulence was excited in the edge plasma region, where the impurity ions exhibited a peaked profile. It has been found that double impurity critical gradients are responsible for triggering the turbulence. Strong stiffness of the impurity profile has been observed during cyclic transitions between the I -phase and H -mode regime. The results suggest that the underlying physics of the self-regulated edge impurity profile offers the possibility for an active control of the pedestal dynamics via pedestal turbulence.
Alpha particle effects in burning tokamak plasmas: overview and specific examples
International Nuclear Information System (INIS)
Using the total power balance of an ignited tokamak plasma as a guideline, a range of alpha driven effects is surveyed regarding their impact on achieving and maintaining fusion burn. Specific examples of MHD and kinetic modes and multi species transport dynamics are discussed, including the possible interaction of these categories of effects. This power balance approach rather than a straightforward enumeration of possible effects serves to reveal their non-linear dependence and the ensuing fragility of our understanding of the approach to and maintenance of ignition. Specific examples are given of the interaction between α-power driven sawtoothing and ideal MHD stability, and direct α-effects on MHD modes including kinetic corrections. Anomalous ion heat transport and central impurity peaking mechanisms and anomalous and collisional α-transport including the ambipolar electric field are discussed
Zhong, W L; Shen, Y; Zou, X L; Gao, J M; Shi, Z B; Dong, J Q; Duan, X R; Xu, M; Cui, Z Y; Li, Y G; Ji, X Q; Yu, D L; Cheng, J; Xiao, G L; Jiang, M; Yang, Z C; Zhang, B Y; Shi, P W; Liu, Z T; Song, X M; Ding, X T; Liu, Yong
2016-07-22
The impact of impurity ions on a pedestal has been investigated in the HL-2A Tokamak, at the Southwestern Institute of Physics, Chengdu, China. Experimental results have clearly shown that during the H-mode phase, an electromagnetic turbulence was excited in the edge plasma region, where the impurity ions exhibited a peaked profile. It has been found that double impurity critical gradients are responsible for triggering the turbulence. Strong stiffness of the impurity profile has been observed during cyclic transitions between the I-phase and H-mode regime. The results suggest that the underlying physics of the self-regulated edge impurity profile offers the possibility for an active control of the pedestal dynamics via pedestal turbulence. PMID:27494476
High-frequency gyrotrons and their application to tokamak plasma heating
International Nuclear Information System (INIS)
A comprehensive analysis of high frequency (100 to 200 GHz) and high power (> 100 kW) gyrotrons has been conducted. It is shown that high frequencies will be required in order for electron cyclotron radiation to propagate to the center of a compact tokamak power reactor. High power levels will be needed in order to ignite the plasma with a reasonable number of gyrotron units. In the first part of this research, a set of analytic expressions, valid for all TE cavity modes and all harmonics, is derived for the starting current and frequency detuning using the Vlasov-Maxwell equations in the weakly relativistic limit. The use of an optical cavity is also investigated
Fully Implicit Iterative Solving Method for the Fokker-Planck Equation in Tokamak Plasmas
Institute of Scientific and Technical Information of China (English)
ZHENG Pingwei; GONG Xueyu; YU Jun; DU Dan
2014-01-01
A three dimensional bounce-averaged Fokker-Planck (FP) numerical code has been newly developed based on fully implicit iterative solving method,and relativistic effect is also included in the code.The code has been tested against various benchmark cases:Ohmic conductivity in the presence of weak Ohmic electric field,runaway losses of electrons in the presence of strong Ohmic electric field,lower hybrid current drive and electron cyclotron current drive via two-or three-dimensional simulation.All the test cases run fast and correctly during calculations.As a result,the code provides a set of powerful tools for studying radio frequency wave heating and current drive in tokamak plasmas.
Control of the Resistive Wall Mode in Advanced Tokamak Plasmas on DIII-D
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Resistive wall mode (RWM) instabilities are found to be a limiting factor in advanced tokamak (AT) regimes with low internal inductance. Even small amplitude modes can affect the rotation profile and the performance of these ELMing H-mode discharges. Although complete stabilization of the RWM by plasma rotation has not yet been observed, several discharges with increased beam momentum and power injection sustained good steady-state performance for record time extents. The first investigation of active feedback control of the RWM has shown promising results: the leakage of the radial magnetic flux through the resistive wall can be successfully controlled, and the duration of the high beta phase can be prolonged. The results provide a comparative test of several approaches to active feedback control, and are being used to benchmark the analysis and computational models of active control
Impurities in a non-axisymmetric plasma: Transport and effect on bootstrap current
Energy Technology Data Exchange (ETDEWEB)
Mollén, A., E-mail: albertm@chalmers.se [Department of Applied Physics, Chalmers University of Technology, Göteborg (Sweden); Max-Planck-Institut für Plasmaphysik, 17491 Greifswald (Germany); Landreman, M. [Institute for Research in Electronics and Applied Physics, University of Maryland, College Park, Maryland 20742 (United States); Smith, H. M.; Helander, P. [Max-Planck-Institut für Plasmaphysik, 17491 Greifswald (Germany); Braun, S. [Max-Planck-Institut für Plasmaphysik, 17491 Greifswald (Germany); German Aerospace Center, Institute of Engineering Thermodynamics, Pfaffenwaldring 38-40, D-70569 Stuttgart (Germany)
2015-11-15
Impurities cause radiation losses and plasma dilution, and in stellarator plasmas the neoclassical ambipolar radial electric field is often unfavorable for avoiding strong impurity peaking. In this work we use a new continuum drift-kinetic solver, the SFINCS code (the Stellarator Fokker-Planck Iterative Neoclassical Conservative Solver) [M. Landreman et al., Phys. Plasmas 21, 042503 (2014)] which employs the full linearized Fokker-Planck-Landau operator, to calculate neoclassical impurity transport coefficients for a Wendelstein 7-X (W7-X) magnetic configuration. We compare SFINCS calculations with theoretical asymptotes in the high collisionality limit. We observe and explain a 1/ν-scaling of the inter-species radial transport coefficient at low collisionality, arising due to the field term in the inter-species collision operator, and which is not found with simplified collision models even when momentum correction is applied. However, this type of scaling disappears if a radial electric field is present. We also use SFINCS to analyze how the impurity content affects the neoclassical impurity dynamics and the bootstrap current. We show that a change in plasma effective charge Z{sub eff} of order unity can affect the bootstrap current enough to cause a deviation in the divertor strike point locations.
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Abe, Mitsushi; Takeuchi, Kazuhiro; Fukumoto, Hideshi; Shimizu, Masashi; Otsuka, Michio (Hitachi Ltd., Ibaraki (Japan). Energy Research Lab.)
1990-02-01
Electromagnetic interactions between plasmas and a vacuum vessel during disruptions are examined experimentally in the Hitachi tokamak HT-2. Eddy currents which flow in the toroidal direction and poloidal coil currents are determined from the measured magnetic data. The currents enable calculation of the electromagnetic force on the vacuum vessel and resistively dissipated magnetic energy. Eddy currents and electromagnetic forces are mainly due to the plasma displacement (shell effect), not decay of the plasma current. Large plasma current quench rate -dI{sub p}/dt is associated with scraping of the plasma by the inner limiter through the rapid plasma radial movement, and the decay rate in circular plasma is twice as large as that in elongated plasma. The magnetic energy dissipation is mainly due to the eddy current of the net toroidal current mode which is induced by large current quench rate. (author).
Abe, Mitsushi; Takeuchi, Kazuhiro; Fukumoto, Hideshi; Shimizu, Masashi; Otsuka, Michio
1990-02-01
Electromagnetic interactions between plasmas and a vacuum vessel during disruptions are examined experimentally in the Hitachi tokamak HT-2. Eddy currents which flow in the toroidal direction and poloidal coil currents are determined from the measured magnetic data. The currents enable calculation of the electromagnetic force on the vacuum vessel and resistively dissipated magnetic energy. Eddy currents and electromagnetic forces are mainly due to the plasma displacement (shell effect), not decay of the plasma current. Large plasma current quench rate -dIP/dt is associated with scraping of the plasma by the inner limiter through the rapid plasma radial movement, and the decay rate in circular plasma is twice as large as that in elongated plasma. The magnetic energy dissipation is mainly due to the eddy current of the net toroidal current mode which is induced by large current quench rate.
The effect of plasma minor-radius expansion in the current build-up phase of a large tokamak
International Nuclear Information System (INIS)
A plasma simulation code has been developed to study the plasma current build-up process in JT-60. Plasma simulation is made with a model which represents well overall plasma behavior of the present-day tokamaks. The external electric circuit is taken into consideration in simulation calculation. An emphasis is placed on the simulation of minor-radius expansion of the plasma and behavior of neutral particles in the plasma during current build-up. A calculation with typical parameters of JT-60 shows a week skin distribution in the current density and the electron temperature, if the minor radius of the plasma expands with build-up of the plasma current. (auth.)
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Among the R and D missions for possible new European plasma fusion devices, the FAST project will address the issue of 'First wall materials and compatibility with ITER /DEMO relevant plasmas'. FAST can operate with ITER relevant values of P/R (up to 22 MW/m, against the ITER 24 MW/m, inclusive of the alpha particles power), thanks to its compactness; thus it can investigate the physics of large heat loads on divertor plates. The FAST divertor will be made of bulk W tiles, for basic operations, but also fully toroidal divertor targets made of liquid lithium (L-Li) are foreseen. Viability tests of such a solution for DEMO divertor will be carried out as final step of an extended program started on FTU tokamak by using a liquid lithium limITER. To have reliable predictions of the thermal loads on the divertor plates and of the core plasma purity a number of numerical self-consistent simulations have been made for the H-mode and steady-state scenario by using the code COREDIV. This code, already validated in the past on experimental data (namely JET, FTU, Textor), is able to describe self-consistently the core and edge plasma in a tokamak device by imposing the continuity of energy and particle fluxes and of particle densities and temperatures at the separatrix. In the present work the results of such calculations will be illustrated, including heat loads on the divertor. The overall picture shows that at the low plasma densities typical of steady state regimes W is effective in dissipating input power by radiative losses, while Li needs additional impurities (Ar, Ne). In the intermediate and, mainly, in the high density H-mode scenarios impurity seeding is needed with either Li or W as target material, but a small (0.08% atomic concentration) amount of Ar, not affecting the core purity, is sufficient to maintain the divertor peak loads below 18 MW/m2 that represents the safety limit for the W monoblock technology, presently accepted for the ITER divertor tiles. The
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Alladio, F.; Micozzi, P. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia
1995-12-01
A correlation has been established between the improvement of the energy confinement time observed in some plasma regimes on the DIII-D tokamak (VH modes and shear reversed discharges) and a geometrical characteristic of the plasma column: the Pfirsch-Schluter-like factor, which multiplies the moment of inertia of the magnetic configuration. Such a quantity is generated by the compression that the flux tubes suffer going from the external to the internal part of the torus. Therefore the configurations in which the module of the total magnetic field is more constant upon the magnetic surfaces (near omnigeneous configurations) show a lower value of the moment of inertia. The geometric parameter of Pfirsch-Schluter determines the transient and steady state behaviour of the plasma rotation under the assumption that the anomalous parallel viscosity is greater that the neoclassical one. In this way, also the profile of the part of the radial electric field (and his absolute value) is influenced by the magnetic configuration. The radial electric field, or, at least, his radial derivative, is invoked by many authors as a principal factor in reducing the turbulence (and so the anomalous transport) in magnetically confined plasmas. In particular, DIII-D machine, the highly elongated and triangular plasma discharges that evolve toward the VH-mode show a lower value of the Pfirsch-Schluter quantity and a higher level of the radial electric field; also the shear reversed profiles tend to lower 1+2q{sup 2} in the central region of the plasma column, driving towards very high values of the electric field within the reversal region.
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Eich, T. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany)]. E-mail: thomas.eich@ipp.mpg.de; Herrmann, A.; Pautasso, G.; Fuchs, J.C.; Gruber, O. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany); Andrew, P. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Asakura, N. [Naka Fusion Research Establishment, JAERI (Japan); Boedo, J.A. [University of California, San Diego, La Jolla, CA 92093 (United States); Corre, Y. [Association EURATOM CEA, Cadarache, 13108 St. Paul-lez-Durance (France); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Fundamenski, W. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Federici, G. [ITER JWS, Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Gauthier, E. [Association EURATOM CEA, Cadarache, 13108 St. Paul-lez-Durance (France); Goncalves, B. [Associacao EURATOM/IST, Instituto Superior Tecnico (Portugal); Kirk, A. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Loarte, A. [CSU-EFDA, Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Matthews, G.F. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Neuhauser, J. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Boltzmann str. 2, D-85748 Garching (Germany); Pitts, R.A. [Association EURATOM, CRPP-EPFL, 1015 Lausanne (Switzerland); Riccardo, V. [EURATOM-UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon OX 14 3DB (United Kingdom); Silva, C. [Associacao EURATOM/IST, Instituto Superior Tecnico (Portugal)
2005-03-01
A comparative analysis of the spatial and temporal characteristics of transient energy loads (ELMs and disruptions) on plasma facing components (PFCs) in present tokamak devices and their extrapolation to next step devices is presented. Type I ELMs lead to the expulsion of energy by the plasma in helical structures with ballooning-like features and toroidal numbers in the range n = 10-15. The plasma energy is transported towards the divertor and the main chamber PFCs leading to significant transient energy loads at these two locations on small wetted area. The largest transient energy fluxes onto PFCs in tokamaks are measured during the thermal quench of disruptions. These fluxes do not exceed greatly those of large Type I ELMs, due to the much larger wetted area for energy flux during the thermal quench compared to Type I ELMs. The implications of these findings for the next step devices are discussed.
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Confinement by noninductive currents was investigated in experiments on MHD instability in a tokamak plasma. The change of dependence of plasma current density and resistivity as a function of electron temperature (parameters which govern the evolution of MHD instability) is stressed. Experiments on the PETULA-B tokamak allowed the determination of the nature of the instabilities (characterization of the numbers m and n for resistance tearing modes; characterization of sawtooth instabilities). Instabilities are analyzed as a function of discharges in plasma current generation by hybrid waves. On PETULA-B, the stabilisation takes two forms: stabilization of sawtooths in correlation with mode excitation (m=2, n=1); and stabilization of sawtooths by mode saturation (m=1, n=1)
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In the future tokamak ITER the damage to the wall after the disruptions can be mitigated using preventive massive gas injection (MGI) of noble gases into confined plasma during the thermal quench. The gas gets ionized in the plasma, and then the ions dump into the scrape-off layer (SOL) and impact on the target. The contamination of core plasma results in fast loss of plasma energy by radiation. The radiation distributes rather homogeneously over the wall. However, enhanced radiation load in e.g. vicinity of gas jet entry is an issue for ITER design that can be addressed numerically. For the modelling the tokamak code TOKES is applied, after upgrading it with toroidally symmetric 2D plasma model. This allowed detailed radiation fluxes and the expansion of noble ions both across and along the magnetic surfaces. In the work one- and two-dimensional (2D) MGI models are evaluated. 2D model is preliminary compared with the tokamak DIII-D. Substantial discrepancies were explained, and then predictive simulations for ITER performed, with the conclusion that after the radiation flush in front of jet entry the wall temperature can exceed the beryllium melting point.
Kinetically Stabilized Axisymmetric Tandem Mirrors: Summary of Studies
Energy Technology Data Exchange (ETDEWEB)
Post, R F
2005-02-08
The path to practical fusion power through plasma confinement in magnetic fields, if it is solely based on the present front-runner, the tokamak, is clearly long, expensive, and arduous. The root causes for this situation lie in the effects of endemic plasma turbulence and in the complexity the tokamak's ''closed'' field geometry. The studies carried out in the investigations described in the attached reports are aimed at finding an approach that does not suffer from these problems. This goal is to be achieved by employing an axisymmetric ''open'' magnetic field geometry, i.e. one generated by a linear array of circular magnet coils, and employing the magnetic mirror effect in accomplishing the plugging of end leakage. More specifically, the studies were aimed at utilizing the tandem-mirror concept in an axisymmetric configuration to achieve performance superior to the tokamak, and in a far simpler system, one for which the cost and development time could be much lower than that for the tokamak, as exemplified by ITER and its follow-ons. An important stimulus for investigating axisymmetric versions of the tandem mirror is the fact that, beginning from early days in fusion research there have been examples of axisymmetric mirror experiments where the plasma exhibited crossfield transport far below the turbulence-enhanced rates characteristic of tokamaks, in specific cases approaching the ''classical'' rate. From the standpoint of theory, axisymmetric mirror-based systems have special characteristics that help explain the low levels of turbulence that have been observed. Among these are the facts that there are no parallel currents in the equilibrium state, and that the drift surfaces of all of the trapped particles are closed surfaces, as shown early on by Teller and Northrop. In addition, in such systems it is possible to arrange that the radial boundary of the confined plasma terminates without
Tokamak plasma power balance calculation code (TPC code) outline and operation manual
International Nuclear Information System (INIS)
This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)
Plasma performance with multi-shot pellet fueling in the HL-1M tokamak
International Nuclear Information System (INIS)
Three to eight hydrogen pellets have been injected into the HL-1M tokamak under ohmically heated plasma condition. The essential features of pellet-fuelled plasma, including the electron density profile and its evolution and the perturbation in the plasma core, have been investigated. The relations of electron density profile with the pellet size, the launching interval and the recycling of the first wall, have been studied. The recycling of the first wall plays an important role in achieving high density discharge. Discharge parameters of ne(0) = 5.3 x 1013cm-3, Wp = 6.0 kJ, τE = 26 ms are obtained under high recycling condition with injection of three smaller pellets (φ 1.0 mm). The pellet ablation cloud image frames were obtained with CCD camera. Brief analyses of the pellet ablation process show that asymmetry ablation and track bending of pellet clouds exist and are the results of stronger ablation in the electron side than in the ion side. The important effect of pellet size and integration on the density turbulence is observed also
Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak
Labate, C.; Di Gironimo, G.; Renno, F.
2015-09-01
Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.
An experimental study of plasma fluctuations in the TCV and TEXTOR Tokamaks
International Nuclear Information System (INIS)
The main body of this thesis reports on the commissioning and first measurements with a novel tangential phase-contrast imaging (TPCI) diagnostic, which had previously been installed in the TCV tokamak. The instrument measures fluctuations in line-integrated electron density along 9 parallel chords within a 6 cm diameter CO2 laser beam. TPCI measurements reveal the first evidence in TCV of the geodesic acoustic mode (GAM), which is an oscillating zonal flow. Frequency, radial wavelength, radial extent and propagation are all in qualitative agreement with a gyro-kinetic simulation and recent theoretical work. The mode is found to have a modest, but measurable magnetic component, whose spatial structure is characterised for the first time in a toroidal plasma. For some experiments, clear evidence is found of the theoretically expected m/n = 2/0 mode structure, although in others the structure appears to be more complex. Electron energy confinement in X2 heated TCV L-mode plasmas had previously been observed to increase on changing the triangularity (δ) of the poloidal plasma cross-section from δ = +0.4 to δ = −0.4. Measurements with the TPCI diagnostic reveal that this change coincides with a clear decrease in both the absolute level and the decorrelation time of broadband electron density fluctuations. This is in agreement with the conjecture that the increased confinement time is caused by a change in the turbulent state. The second part of the thesis reports on a fluctuation study in the TEXTOR tokamak. At sufficiently weak toroidal magnetic field, NBI heated, limited TEXTOR plasmas exhibit bursts of beam-ion driven ‘fishbone’ and Alfvén modes, which are characterised using the multi-antenna reflectometer and Mirnov coils. In H-mode the fishbone triggers ELMs and in L-mode it triggers previously unobserved bursts of particle recycling, resembling the ELMs. The reflectometer phase shows statistically significant bispectral coherence between the fishbone
Experimental study of the interaction between RF antennas and the edge plasma of a tokamak
International Nuclear Information System (INIS)
Antennas operating in the ion cyclotron range of frequency (ICRF) provide a useful tool for plasma heating in many tokamaks and are foreseen to play an important role in ITER. However, in addition to the desired heating in the core plasma, spurious interactions with the plasma edge and material boundary are known to occur. Many of these deleterious effects are caused by the formation of radio-frequency (RF) sheaths. The aim of this thesis is to study, mainly experimentally, scrape-off layer (SOL) modifications caused by RF sheaths effects by means of Langmuir probes that are magnetically connected to a powered ICRH antenna. Effects of the two types of Faraday screens' operation on RF-induced SOL modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, injected power and SOL density. In addition to experimental work, the influence of RF sheaths on retarding field analyzer (RFA) measurements of sheath potential is investigated with one-dimensional particle-in-cell code. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the sheath potential only for ion plasma frequencies ωπ similar to RF cyclotron frequency ωrf, while for the real SOL conditions (ωπ ≥ ωrf), when the RFA is magnetically connected to RF region, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed by using broadening of the ion distribution function as an evidence of the RF electric fields in the sheath. RFA measurements in Tore Supra indicate that RF potentials do indeed propagate from the antenna 12 m along magnetic field lines. (author)
The two-dimensional structure of radiative divertor plasmas in the DIII-D tokamak
International Nuclear Information System (INIS)
Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne≤1021m-3, 0.5eV≤Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion endash neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3 endash 5, consistent with the measurements. Absolutely calibrated vacuum ultraviolet spectroscopy and 2-D images of impurity emission show that carbon radiation near the X-point, and deuterium radiation near the target plates contribute to the reduction in Te. Uniformity of radiated power (Prad) (within a factor of 2) along the outer divertor leg, with peak heat flux on the divertor target reduced fourfold, was obtained. A comparison with 2-D fluid simulations shows good agreement when physical sputtering and an ad hoc chemical sputtering source (0.5%) from the private flux region surface are used. copyright 1997 American Institute of Physics
International Nuclear Information System (INIS)
The behaviour of the FTU tokamak plasma has been analyzed by using two reconstructive MHD equilibrium codes: the first code works by using the magnetic data alone and the second one by including as well the shape of the kinetic pressure profile, as obtained from the measured profiles of electron temperature Te and density ne. The code that analyzes the magnetic data alone provides a good evaluation of the macroscopic quantities such as the poloidal beta Bp and the internal inductance li, if the plasma elongation is greater than 1. 04. No detailed information about the toroidal current density profile Jφ and the safety factor profile q can be obtained from the magnetic data alone. On the other hand, the coupling of magnetic and kinetic data is able to provide a reasonable estimate of the toroidal current density profile and of its behaviour during the plasma discharge. The reliability of the Jφ and q profiles reconstruction has been explored and validated by a detailed comparison with the observed MHD behaviour of the FTU plasma discharges. A good agreement between the appearance of the sawtooth activity and the drop of the safety factor on the magnetic axis qo to unity is observed. Also, at least for edge safety factors qψ less than four, the sawtooth inversion radius is found to be very close to the q=1 surface. A remarkable correspondence between Jφ and Te3/2 is found in sawtoothing discharges. The structure of the snake oscillation in pellet injected discharges is found to be strictly correlated to the position of the q=1 surface. A cylindrical linear tearing mode stability calculation applied to the reconstructed Jφ profile has shown qualitative agreement with the appearance of the Mirnov oscillations. Finally the magnetic reconnections between double resonant surfaces during the rise of the plasma current or after the pellet injection have provided an interesting validation of the Jφ profile reconstruction. 56 tabs
Tokamak advanced pump limiter experiments and analysis
International Nuclear Information System (INIS)
Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment
Energy Technology Data Exchange (ETDEWEB)
Schmitz, D.
2006-07-15
Within the scope of this work itermittent events in the plasma edge of the tokamak TEXTOR were characterized. For the data of measurements of the density and the poloidal electrical field were analysed. The data was collected by a reciprocating and a fixed probe as well as by a lithium beam. The intermittent behaviour was quantified by the statistical moments of the data. If intermittency is high, coherent structures (also called blobs) can be detected. The detected blobs were described using the statistical method of conditional averaging. The main results can be summarised as follows: Intermittent behavoiur has been detected in the scrap off layer of the tokamak TEXTOR and it is increasing with the radius from the last closed flux surface (LCFS) on. On the midplane the blobs in the limiter geometry have a radial size of up to 8 cm and move onto the wall with velocities as high as (1-7)% of the ion sound speed. It was found that intermittent transport causes 40% of the total perpendicular transport in the investigated discharges. In the upper part of the tokamak there is less intermittency. This is reasonable if intermittency is caused by interchange instabilities which mainly occur on the low field side of the tokamak. With the Dynamic Ergodic Divertor (DED) and the associated formation of tearing modes intermittency is increasing. This can also be due to the steeper gradient of density in the scrap off layer close to the LCFS which is caused by gas puffing used for the regulation of the density. Outside the LCFS the ergodic field does not have any influence on the characteristics of blobs. Within the LCFS density holes have been found which propagate towards the centre of the plasma. The radial transport due to blobs is still the same. In general the velocity of the detected blobs is proportional to the square root of their poloidal size. That confirms the prediction of the blob model in which the nonlinear development of interchange instabilities causes the
Favorable effects of turbulent plasma mixing on the performance of innovative tokamak divertors
Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.
2013-10-01
The problem of reducing the heat load on plasma-facing components is one of the most demanding issues for MFE devices. The general approach to the solution of this problem is the use of a specially configured poloidal magnetic field, so called magnetic divertors. In recent years, novel divertors possessing the 2-nd and 3-rd order nulls of the poloidal field (PF) have been proposed. They are called a ``snowflake'' (SF) and a ``cloverleaf'' (CL) divertor, respectively, due to characteristic shape of the magnetic separatrix. Among several beneficial features of such divertors is an effect of strong turbulent plasma mixing that is intrinsic to the zone of weak PF near the null-point. The turbulence spreads the heat flux between multiple divertor exhaust channels and increases the heat flux width within each channel. Among physical processes affecting the onset of convection the curvature-driven mode of axisymmetric rolls is most prominent. The effect is quite significant for the SF and is even stronger for the CL divertor. Projections to future ITER-scale facilities are discussed. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.
Saha, S. K.; Kumar, R.; Hui, A. K.
2001-11-01
Plasma diamagnetism has been measured in the SINP tokamak by a toroidal flux loop placed inside the vacuum vessel. The flux due to the strong toroidal field has been compensated for by a coplaner annular loop which encircles but does not contain the plasma column. The influence of the eddy currents in the vacuum vessel and the conducting shell in these loops has been calculated analytically by a circuit model using the theory of linear networks and compensated accordingly. This method has been shown to yield an almost exact compensation for toroidal flux (˜0.01%) as well as pickups from other fields. Typical results with plasma shots have been presented.
High-pressure duo-multichannel soft x-ray spectrometer for tokamak plasma diagnostics
International Nuclear Information System (INIS)
A high-resolution, time-resolving soft X-ray multichannel spectrometer (SOXMOS) that permits the simultaneous measurement of emission in two different spectral ranges has been developed and tested extensively for tokamak plasma diagnostics. The basic instrument is a high-resolution, interferometrically adjusted, extreme grazing incidence Schwob-Fraenkel duochromator. The instrument is equipped with two multichannel detectors that are adjusted interferometrically and scan along the Rowland circle. Each consists of an MgF2 coated, funneled microchannel plate, associated with a phosphor screen image intensifier that is coupled to a 1024-element photodiode array by a flexible fibrer optic conduit. The total wavelength coverage of the instrument is 5 to 3400 A with a measured resolution (FWHM) of about 0.2 A when equipped with a 600 g/mm grating, and 5 to 85 A with a resolution of about 0.06 A using a 2400 g/mm grating. The simultaneous spectral coverage of each detector varies from 15 A at the short wavelength limit to 70 A at the long wavelength limit with the lower dispersion grating. The minimum read-out time for a full spectral portion is 17 ms, but several individual lines can be measured with 1 ms time resolution by selected pixel readout. Higher time resolution can be achieved by replacing one multichannel detector with a single channel electron multiplier detector. Examples of data from the PLT and TFTR tokamaks are presented to illustrate the instrument's versatility, high spectral resolution, and high signal-to-noise ratio even in the 10 A region. 44 refs., 20 figs
Plasma-wall interaction study in the open divertor of Globus-M spherical tokamak
International Nuclear Information System (INIS)
Globus-M is the first Russian spherical tokamak, which was built at the A.F. Ioffe Institute in 1999. Currently about 90% of in vessel area, which is faced to plasma, is covered by protection tiles. Tiles are constructed from RGTi graphite, doped by 2 at.% Ti and 0.3-0.7 at.% Si. Gradual increase in plasma parameters was recorded as the wall area protected by tiles constantly grew up. The plasma-facing surface including graphite armor was covered by a-B/C:H layers periodically. Globus-M is usually operating in the density range n ∼ (3-10) x 1019 m-3 with plasma currents 0.2-0.25 MA and toroidal magnetic field of 0.4 T. The main working gas is deuterium. NB injection and IC resonance heating at hydrogen minority fundamental frequency are used as auxiliary heating methods. Specific power deposition is high up to value of several MW/m. Power density deposited at the first wall is also high. This is due to 'spherical' geometry of plasma column in which the first wall area is small with respect to plasma volume and plasma is close to the first wall position. The 'focusing' of power fluxes along separatrix strike points could increase power density up to 10 MW/m2 at the divertor target. RGTi diverter tiles analysis was performed after irradiation by plasma during big number of shots (10000 shots in average). Composition and morphology of the surface layers were examined by different diagnostic tools (electron probe microanalysis, scanning electron microscope, Rutherford backscattering, nuclear resonance reactions, thermal desorption spectroscopy). The most of tiles were covered with deposited mixed layers. Deposits exist even in high flux regions (separatrix strike points). The mixed layers are composed from the elements used in the vessel construction and in conditioning technology processes. The concentration and absorbed deuterium depth profiles in tiles are analyzed and conditions of deuterium desorption are studied. Important result is that deuterium is absorbed
International Nuclear Information System (INIS)
A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic close-quote s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. copyright 1997 American Institute of Physics
Energy Technology Data Exchange (ETDEWEB)
Meo, F.; Stansfield, B.L.; Chartre, M.; de Villers, P.; Marchand, R.; Ratel, G. [Centre Canadien de Fusion Magnetique, 1804 Boulevard Lionel-Boulet, Varennes, Quebec, J3X 1S1 (CANADA)
1997-09-01
A toroidally viewing spectroscopic imaging system has been developed for the Tokamak de Varennes providing measurements of the poloidal distribution of the absolute radiated power of deuterium and impurity species in the upper divertor region. Real time digitization is achieved using a low cost PC based digital imaging system. This system is used to obtain measurements of the divertor strike point as well as the shape of the flux surfaces in the divertor. The diagnostic{close_quote}s excellent spatial resolution and toroidal view provides an opportunity to quantitatively compare the measured two dimensional (2D) radiated power distribution to that calculated from 2D Monte Carlo transport codes. These 2D images provide unique and valuable information on the physics of local plasma interactions with divertor components and particle transport in a closed divertor. Additionally, by using two cameras simultaneously, the line ratio technique can be applied to the images to estimate plasma parameters in the divertor. {copyright} {ital 1997 American Institute of Physics. }
Energy Technology Data Exchange (ETDEWEB)
Semerok, A., E-mail: alexandre.semerok@cea.fr [CEA, DEN, DPC/SEARS/LISL, F-91191 Gif-sur-Yvette (France); Grisolia, C. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)
2013-08-21
Feasibility of in situ LIBS remote measurements with the plasma facing components (PFCs) from the European tokamaks (TORE SUPRA, CEA Cadarache, France and TEXTOR, Julich, Germany) has been studied in laboratory using Q-switched nanosecond Nd–YAG lasers. LIBS particular properties and optimal parameters were determined for in-depth PFCs characterisation. The LIBS method was in situ tested on the Joint European Torus (JET) in the UK with the EDGE LIDAR Laser System (Ruby laser, 3 J, 690 nm wavelength, 300 ps pulse duration, intensity up to 70 GW/cm{sup 2}). Several analytical spectral lines of H, CII, CrI, and BeII in plasma were observed and identified in 400–600 nm spectral range with the optimised LIBS and detection system. The LIBS in-depth cartography is in agreement with the surface properties of the tile under analysis, thus confirming feasibility of in situ LIBS. Further LIBS technique improvements required to provide tritium concentration measurements more accurately are discussed.
Microwave start-up of tokamak plasmas near electron cyclotron and upper hybrid resonances
International Nuclear Information System (INIS)
The scenario of toroidal plasma start-up with microwave initiation and heating near the electron cyclotron frequency is suggested and examined in this paper. Microwave irradiation from the high-field side and an anomalously large absorption of the extraordinary waves near the upper hybrid resonance are assumed. The dominant electron energy losses are assumed to be due to magnetic-field curvature and parallel drifts, ionization of neutrals, cooling by ions, and radiation by low-Z impurities. It is shown by particle and energy balance considerations that electron temperatures around 250eV and densities of 1012-1013cm-3 can be maintained, at least in a narrow region near the upper hybrid resonance, with modest microwave powers in the Impurity Study Experiment (ISX) (120kW at 28GHz) and The Next Step (TNS) (0.57MW at 120GHz). The loop voltages required for start-up from these initial plasmas are also estimated. It is shown that the loop voltage can be reduced by a factor of five to ten from that for unassisted start-up without an increase in the resistive loss in volt-seconds. If this reduction in loop voltage is verified in the ISX experiments, substantial savings in the cost of power supplies for the Ohmic heating (OH) and equilibrium field (EF) coils can be realized in future large tokamaks. (author)
Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak
Directory of Open Access Journals (Sweden)
Casali L.
2014-01-01
Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.
Development of burning plasma and advanced scenarios in the DIII-D tokamak
International Nuclear Information System (INIS)
Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q ∼ 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque. (author)
Ren, Jing; Liu, Yueqiang; Liu, Yue; Medvedev, S. Yu; Wang, Zhirui; Xia, Guoliang
2016-11-01
The effects of an ideal/resistive conducting wall, the drift kinetic resonances, as well as the toroidal plasma flow, on the stability of the ideal external kink mode are numerically investigated for a reactor-relevant tokamak plasma with strongly negative triangularity (NTR) shaping. Comparison is made for a similar plasma equilibrium, but with positive triangularity (PTR). It is found that the ideal wall stabilization is less efficient for the kink stabilization in the NTR plasma due to a less ‘external’ eigenmode structure compared to the PTR plasma. The associated plasma displacement in the NTR plasma does not ‘balloon’ near the outboard mid-plane, as is normally the case for the pressure-driven kink-ballooning instability in PTR plasmas, but being more pronounced near the X-points. The toroidal flow plays a similar role for the kink stability for both NTR and PTR plasmas. The drift kinetic damping is less efficient for the ideal external kink mode in the NTR plasma, despite a somewhat larger fraction of the particle trapping near the plasma edge compared to the PTR equilibrium. However, the drift kinetic damping of the resistive wall mode (RWM) in the NTR plasma is generally as efficient as that of the PTR plasma, although the RWM window, in terms of the normalized pressure, is narrower for the NTR plasma.
I.R. and F.I.R. laser polarimetry as a diagnostic tool in high-β and Tokamak plasmas
International Nuclear Information System (INIS)
The change of the polarization state of an electromagnetic wave (E.M.W.) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the E.M.W. allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-Pinches plasmas (lambda approx. 10μm). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-Pinch/UNICAMP, using lasers developed at UNICAMP. (Author)
Influence of plasma rotation on tearing mode stability on the ASDEX upgrade tokamak
International Nuclear Information System (INIS)
Neoclassical tearing modes (NTM) are one of the most serious performance limiting instabilities in next-step fusion devices like ITER. NTMs are destabilised as a consequence of a seed perturbation (trigger) and are driven by a loss of helical bootstrap current inside the island. The appearance of these instabilities is accompanied with a loss of confined plasma energy. Additionally, these modes can stop the plasma rotation, lock to the vessel wall, flush out all plasma energy and terminate a discharge via a disruption. In ITER the confinement reduction will limit the achievable fusion power, whereas a disruption is likely to damage the vessel wall. In order to mitigate and control NTMs in ITER, extrapolations based on the present understanding and observations must be made. One key issue is the rotation dependence of NTMs, especially at the NTM onset. ITER will be operated at low plasma rotation, which is different from most present day experiments. No theory is currently available to describe this dependence. Experiments are therefore required to provide a basis for the theory to describe the physics. Additionally from the experiments scalings can be developed and extrapolated in order to predict the NTM behaviour in the parameter range relevant for ITER. Another important issue is the influence of externally applied magnetic perturbation (MP) fields on the NTM stability and frequency. These fields will be used in ITER primarily for the mitigation of edge instabilities. As a side effect they can slow down an NTM and the plasma rotation, which supports the appearance of locked modes. Additionally, they can also influence the stability of an NTM. This interaction has to be predicted for ITER, based on models validated at present day devices. In this work the influence of plasma rotation on the NTM onset at the ASDEX Upgrade tokamak (AUG) is investigated. An onset database has been created in which the different trigger mechanisms have been identified. Based on this
International Nuclear Information System (INIS)
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
Transport and confinement in the Mega Ampère Spherical Tokamak (MAST) plasma
Akers, R. J.; Ahn, J. W.; Antar, G. Y.; Appel, L. C.; Applegate, D.; Brickley, C.; Bunting, C.; Carolan, P. G.; Challis, C. D.; Conway, N. J.; Counsell, G. F.; Dendy, R. O.; Dudson, B.; Field, A. R.; Kirk, A.; Lloyd, B.; Meyer, H. F.; Morris, A. W.; Patel, A.; Roach, C. M.; Rohzansky, V.; Sykes, A.; Taylor, D.; Tournianski, M. R.; Valovi, M.; Wilson, H. R.; Axon, K. B.; Buttery, R. J.; Ciric, D.; Cunningham, G.; Dowling, J.; Dunstan, M. R.; Gee, S. J.; Gryaznevich, M. P.; Helander, P.; Keeling, D. L.; Knight, P. J.; Lott, F.; Loughlin, M. J.; Manhood, S. J.; Martin, R.; McArdle, G. J.; Price, M. N.; Stammers, K.; Storrs, J.; Walsh, M. J.; MAST, the; NBI Team
2003-12-01
A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampère Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement HH factor (w.r.t. scaling law IPB98[y, 2]) around ~1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L H power threshold scaling proportional to plasma surface area (rather than PLH ~ R2). In addition, MAST favours an inverse aspect ratio scaling PLH ~ egr0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling Wped ~ egr-2.13 and modifies the exponents on R, BT and kgr. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect ratio. Electron and ion energy diffusivities
International Nuclear Information System (INIS)
Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in
Energy Technology Data Exchange (ETDEWEB)
Mantsinen, M. [Helsinki Univ. of Technology, Espoo (Finland). Dept. of Technical Physics
1999-06-01
Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in
Transport and confinement in the Mega Ampere Spherical Tokamak (MAST) plasma
International Nuclear Information System (INIS)
A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampere Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement HH factor (w.r.t. scaling law IPB98[y, 2]) around approx. 1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L-H power threshold scaling proportional to plasma surface area (rather than PLH approx. R2). In addition, MAST favours an inverse aspect ratio scaling PLH approx. epsilon 0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling Wped approx. epsilon -2.13 and modifies the exponents on R, BT and Kappa. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect ratio
Transport and confinement in the Mega Ampere Spherical Tokamak (MAST) plasma
Energy Technology Data Exchange (ETDEWEB)
Akers, R J; Ahn, J W; Appel, L C; Brickley, C; Bunting, C; Carolan, P G; Challis, C D; Conway, N J; Counsell, G F; Dendy, R O; Dudson, B; Field, A R; Kirk, A; Lloyd, B; Meyer, H F; Morris, A W; Patel, A; Roach, C M; Sykes, A; Taylor, D; Tournianski, M R; Valovic, M; Wilson, H R; Axon, K B; Buttery, R J; Ciric, D; Cunningham, G; Dowling J; Dunstan, M R; Gee, S J; Gryaznevich, M P; Helander, P; Keeling, D L; Knight, P J; Lott, F; Loughlin, M J; Manhood, S J; Martin, R; McArdle, G J; Price, M N; Stammers, K; Storrs, J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Antar, G Y [Fusion Energy Research Program, University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Applegate, D [Imperial College of Science, Technology and Medicine, University of London, London SW7 2BZ (United Kingdom); Rohzansky, V [St. Petersburg State Politechnical University, Polytechnicheskaya 29, 195251 St. Petersburg (Russian Federation); Walsh, M J [Walsh Scientific Ltd., Abingdon, Oxon OX14 3EB (United Kingdom)
2003-12-01
A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampere Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement H{sub H} factor (w.r.t. scaling law IPB98[y, 2]) around approx. 1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L-H power threshold scaling proportional to plasma surface area (rather than P{sub LH} approx. R{sup 2}). In addition, MAST favours an inverse aspect ratio scaling P{sub LH} approx. epsilon 0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling W{sub ped} approx. epsilon -2.13 and modifies the exponents on R, B{sub T} and Kappa. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using
Directory of Open Access Journals (Sweden)
D. Y. Klimushkin
Full Text Available The structure of monochromatic MHD-waves with large azimuthal wave number m≫1 in a two-dimensional model of the magnetosphere has been investigated. A joint action of the field line curvature, finite plasma pressure, and transversal equilibrium current leads to the phenomenon that waves, standing along the field lines, are travelling across the magnetic shells. The wave propagation region, the transparency region, is bounded by the poloidal magnetic surface on one side and by the resonance surface on the other. In their meaning these surfaces correspond to the usual and singular turning points in the WKB-approximation, respectively. The wave is excited near the poloidal surface and propagates toward the resonance surface where it is totally absorbed due to the ionospheric dissipation. There are two transparency regions in a finite-beta magnetosphere, one of them corresponds to the Alfvén mode and the other to the slow magnetosound mode.
Key words. Magnetosphere · Azimuthally small-scale waves · MHD waves
Axisymmetric Alfvén resonances in a multi-component plasma at finite ion gyrofrequency
Directory of Open Access Journals (Sweden)
D. Yu. Klimushkin
2006-05-01
Full Text Available This paper deals with the spatial structure of zero azimuthal wave number ULF oscillations in a 1-D inhomogeneous multi-component plasma when a finite ion gyrofrequency is taken into account. Such oscillations may occur in the terrestrial magnetosphere as Pc1-3 waves or in the magnetosphere of the planet Mercury. The wave field was found to have a sharp peak on some magnetic surfaces, an analogy of the Alfvén (field line resonance in one-fluid MHD theory. The resonance can only take place for waves with frequencies in the intervals ω<ω_{ch} or Ω_{0}<ω< ω_{cp}, where ω_{ch} and ω_{cp} are heavy and light ions gyrofrequencies, and Ω_{0} is a kind of hybrid frequency. Contrary to ordinary Alfvén resonance, the wave resonance under consideration takes place even at the zero azimuthal wave number. The radial component of the wave electric field has a pole-type singularity, while the azimuthal component is finite but has a branching point singularity on the resonance surface. The later singularity can disappear at some frequencies. In the region adjacent to the resonant surface the mode is standing across the magnetic shells.
Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2
Energy Technology Data Exchange (ETDEWEB)
Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio [Hitachi Ltd., Tokyo (Japan)
1997-11-01
A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)
Axisymmetric Toroidal Equilibrium with Sheared Toroidal Flows
Institute of Scientific and Technical Information of China (English)
石秉仁
2002-01-01
Problem of the axisymmetric toroidal equilibrium with pure sheared toroidal flow is involved. For standard tokamak equilibrium, general approximate solutions are analytically pursued for arbitrary current profile and non-circular cross-section. Equilibrium properties including the flow-induced density asymmetry are analyzed.
International Nuclear Information System (INIS)
It is shown that pressure-driven, ideal external modes in tokamaks can be fully stabilized by resistive walls when the plasma rotates at some fraction of the sound speed. For wall stabilized plasmas, there are two types of potentially unstable external modes: those which are nearly locked to the wall and those which rotate with the plasma. For the modes rotating with the plasma, the stabilizing effect of the wall increases when the wall is brought closer to the plasma, while, for the wall-locked modes, the stabilization improves with increasing wall distance. When the plasma rotates at some fraction of the sound speed, there is a window of stability to both the wall-locked and the rotating mode. This window closes when beta exceeds a new limit which can be significantly higher than the wall-at-infinity limit. The stabilization depends principally on the toroidal coupling to sound waves and is affected by ion Landau damping. Two dimensional stability calculations are presented to evaluate the gains in beta limit resulting from this wall stabilization for different equilibria and rotation speeds. In particular, results are shown for advanced tokamak configurations with bootstrap fractions of ≅ 100%. (author) 14 figs., 25 refs
Study of the linear and non-linear coupling of the LH wave to the tokamak plasmas
International Nuclear Information System (INIS)
In order to achieve long pulse operation with a tokamak, additional heating and current drive systems are necessary. High frequency antennas, which deliver several megawatts of power to the plasma, are currently used in several tokamaks. Moreover, a good control of the coupling of the wave launched by the antenna to the edge plasma is required to optimize the efficiency of heating and current drive LH systems. However, non-linear effects which depend on the level of injected power in the plasma strongly damage the coupling of the LH wave at particular edge parameters (density and temperature profiles). Results presented in the manuscript deal with the study of the linear and non-linear coupling of the LH wave to the plasma. In the framework of the commissioning of the Passive Active Multijunction antenna in 2009 on the Tore Supra tokamak aiming at validating the LH system suggested for ITER, the characterisation of its coupling properties was realized from low power experiments. The experimental results, which are compared with the linear coupling code ALOHA, have validated the theoretical predictions of good coupling at edge plasma density around the cut-off density. Besides, the ponderomotive effect is clearly identified as responsible for the deterioration in the coupling of the wave, which is measured under particular edge plasma conditions. A theoretical model combining the coupling of the LH wave with the ponderomotive force is suggested to explain the experimental observations. Thus, a new full wave code (named PICCOLO-2D) was developed and results from simulations validate the working hypothesis of the contribution of the ponderomotive effect in the non-linear observations of LHCD coupling on Tore Supra. (author)
Energy Technology Data Exchange (ETDEWEB)
Wijnands, T.J. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Sciences de la Matiere
1997-03-01
This work is focussed on an important and recent development in present day Controlled Nuclear Fusion Research and Tokamaks. The aim is to optimise the energy confinement for a certain magnetic configuration by adapting the radial distribution of the current. Of particular interest are feedback control scenarios with stationary modifications of the current profile using current, driven by Lower Hybrid waves. A new feedback control system has been developed for Tore Supra and has made a large number of new operation scenarios possible. In one of the experiments described here, there is no energy exchange between the poloidal field system and the plasma, the current is controlled by the power of the Lower Hybrid waves while the launched wave spectrum is used to optimise the current profile shape and the energy confinement. (author) 151 refs.
Blob/hole formation and zonal-flow generation in the edge plasma of the JET tokamak
DEFF Research Database (Denmark)
Xu, G.S.; Naulin, Volker; Fundamenski, W.;
2009-01-01
The first experimental evidence showing the connection between blob/hole formation and zonal-flow generation was obtained in the edge plasma of the JET tokamak. Holes as well as blobs are observed to be born in the edge shear layer, where zonal-flows shear off meso-scale coherent structures...... to the zonal flows via the turbulent Reynolds stress, resulting in nonlinear saturation of edge turbulence and suppression of meso-scale fluctuations. These findings carry significant implications for the mechanism of structure formation in magnetically confined plasma turbulence....
International Nuclear Information System (INIS)
The equipment (consisting of two soft-X-ray cameras and a set of diffused silicon detectors), the data processing system, and the computerized tomography for the study of sawtooth oscillations in the JET tokamak are described. Comparisons are given between the tomographic reconstructions and simulated (Kadomtsev and convection) model data. Also discussed are results on the development of ''snakes'' (i.e., thin regions of long-lived elevated plasma density) due to pellet injection. From the tomographically inverted profiles the impurity profiles and diffusion coefficients are obtainable in the case when one plasma impurity is dominant. 13 refs, 10 figs
Simple contour analysis of ignition conditions and plasma operating regimes in tokamaks
International Nuclear Information System (INIS)
Contour plots of ignition, auxiliary power requirements, heating and operating windows, optimal path to ignition, ignition margin, etc., are generated analytically in terms of a small number of parameters (aB02/q/sub */, R0/B0, , etc.) for classes of devices with equivalent performance. Numerical studies are carried out to map the physics design space. Considering both the Murakami density limit (approx.B0/R0) and the Troyon beta limit (approx.I/aB0), results from analytic calculations indicate that in a standard tokamak geometry (A approx. 2.5 to 3.5, kappa = b/a approx. 1.6 to 1.7, q/sub psi/ approx. 2.6) devices with aB02/q/sub */ approx. 20 should be ignitable provided confinement does not degrade with heating (ohmic + alpha + auxiliary, etc.) power; however, aB02/q/sub */ approx. 30 (25) may be required for minimal ignition for a typical L- (H-) mode confinement scaling. Increased plasma elongation (kappa approx. 2) may help to reduce these requirements
Energy Technology Data Exchange (ETDEWEB)
Martins, Caroline G.L.; Roberto, M. [Instituto Tecnologico de Aeronautica (ITA/CTA), Sao Jose dos Campos, SP (Brazil); Carvalho, R. Egydio de [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), SP (Brazil); Caldas, I.L. [Universidade de Sao Paulo (USP), SP (Brazil)
2012-07-01
Full text: We present a study that deals with meandering curves which arise after the reconnection process (or overlap) of resonances (1), that occurs only in non-twist discrete maps (2). Meandering curves formed by this kind of process play the role of barriers for chaotic transport in phase space, because inside the meandering region there is a special torus, called shearless torus, known as the strongest torus in a dynamical system (1). We introduce an extra perturbation in the Standard Non-twist Map (3), and we call this new map Labyrinthic Standard Non-twist Map (4). The labyrinthic map proposed in this work shows multiple reconnection processes of resonances, presenting multiple barriers for chaotic transport. Having applications in important areas such as the physics of thermonuclear plasmas confined in tokamaks for the extraction of clean energy. (1) D. del-Castillo-Negrete, J. M. Greene, P. J. Morrison, Physica D 91, 1 (1996) (2) A.J. Lichtenberg and M.A. Lieberman, Regular and Chaotic Dynamics (Springer, New York, 1992) (3) D. Del-Castillo-Negrete and P. J. Morrison, Phys. Fluids A 5, 948 (1993) (4) Caroline G. L. Martins; R. Egydio de Carvalho; I. L. Caldas; M. Roberto. Labyrinthic standard non-twist map. Journal of Physics A, Mathematical and Theoretical, v. 44, p. 045102 (2011). (author)
International Nuclear Information System (INIS)
Full text: We present a study that deals with meandering curves which arise after the reconnection process (or overlap) of resonances (1), that occurs only in non-twist discrete maps (2). Meandering curves formed by this kind of process play the role of barriers for chaotic transport in phase space, because inside the meandering region there is a special torus, called shearless torus, known as the strongest torus in a dynamical system (1). We introduce an extra perturbation in the Standard Non-twist Map (3), and we call this new map Labyrinthic Standard Non-twist Map (4). The labyrinthic map proposed in this work shows multiple reconnection processes of resonances, presenting multiple barriers for chaotic transport. Having applications in important areas such as the physics of thermonuclear plasmas confined in tokamaks for the extraction of clean energy. (1) D. del-Castillo-Negrete, J. M. Greene, P. J. Morrison, Physica D 91, 1 (1996) (2) A.J. Lichtenberg and M.A. Lieberman, Regular and Chaotic Dynamics (Springer, New York, 1992) (3) D. Del-Castillo-Negrete and P. J. Morrison, Phys. Fluids A 5, 948 (1993) (4) Caroline G. L. Martins; R. Egydio de Carvalho; I. L. Caldas; M. Roberto. Labyrinthic standard non-twist map. Journal of Physics A, Mathematical and Theoretical, v. 44, p. 045102 (2011). (author)
Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST
Xu, X. Q.
2008-07-01
We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field 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 backward 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 the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (ψ,θ,γ,μ) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.
Non-linear Heat Transport Modelling with Edge Localized Modes and Plasma Edge Control in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Becoulet, M.; Huysmans, G.; Thomas, P.; Ghendrih, P.; Grosman, A.; Monier-Garbet, P.; Garbet, X.; Zwingman, W.; Nardon, E. [Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Moyer, R. [California Univ., San Diego, La Jolla CA (United States); Evans, T.; Leonard, A. [General Atomics, San Diego, CA (United States)
2004-07-01
The paper presents a new approach for the modelling of the pedestal energy transport in the presence of Type I ELMs (edge localized mode) based on the linear ideal MHD code MISHKA coupled with the non-linear energy transport code TELM in a realistic tokamak geometry. The main mechanism of increased transport through the External Transport Barrier (ETB) in this model of ELMs is the increased convective flux due to the MHD velocity perturbation and an additional conductive flux due the radial perturbation of the magnetic field leading to a flattening of the pressure profile in the unstable zone. The typical Type I ELM time-cycle including the destabilization of the ballooning modes leading to the fast (200 {mu}s) collapse of the pedestal pressure followed by the edge pressure profile re-building on a diffusive time scale was reproduced numerically. The possible mechanism of Type I ELMs control using a stochastic plasma boundary created by external coils is modelled in the paper. In the stochastic layer the transverse transport is effectively increased by the magnetic field line diffusion. The modelling results for DIII-D experiment on Type I ELM suppression using the external perturbation from the I-coils demonstrated the possibility to decrease the edge pressure gradient just under the ideal ballooning limit, leading to the high confinement regime without Type I ELMs. (authors)
Science Court on ICRH [ion cyclotron resonance heating] modeling of tokamak plasmas
International Nuclear Information System (INIS)
The Applied Plasma Physics (APP) Theory program in the Office of Fusion Energy is charged with supporting the development of advanced physics models for fusion research. One such effort is ion cyclotron resonance heating (ICRH), which has seen substantial progress recently. However, due to serious questions about the adequacy of present models for CIT (Compact Ignition Tokamak), a Science Court was formed to assess ICRH models, including: validity of theoretical and computational approximations; underlying physics assumptions and corresponding limits on the results; self-consistency; any subsidiary issues needing resolution (e.g., new computer tools); adequacy of the models in simulating experiments (especially CIT); and new or improved experiments to validate and refine the models. The Court did not review work by specific individuals, institutions, or programs, thereby avoiding any biases along these lines. Rather, the Science Court was carefully structured as a technical review of ICRH theory and modeling in the US. This paper discusses the Science Court process, findings, and conclusions
Institute of Scientific and Technical Information of China (English)
Xu Xiao-Yuan; Wang Jun; Yu Yi; Wen Yi-Zhi; Yu Chang-Xuan; Liu Wan-Dong; Wan Bao-Nian; Gao Xiang; N. C. Luhmann; C. W. Domier; Jian Wang; Z. G. Xia; Zuowei Shen
2009-01-01
The fluctuation of the electron temperature has been measured by using the electron cyclotron emission imaging in the Hefei Tokamak-7 (HT-7) plasma. The electron temperature fluctuation with a broadband spectrum shows that it propagates in the electron diamagnetic drift direction, and the mean poloidal wave-number kg is calculated to be about 1.58 cm-1, or keps ≈0.34. It indicates that the fluctuation should come from the electron drift wave turbulence. The linear global scaling of the electron temperature fluctuation with the gradient of electron temperature is consistent with the mixing length scale qualitatively. Evolution of spectrum of the fluctuation during the sawtooth oscillation phases is investigated, and the fluctuation is found to increase with the gradient of electron temperature increasing during most phases of the sawtooth oscillation. The results indicate that the electron temperature gradient is probably the driver of the fluctuation enhancement. The steady heat flux driven by electron temperature fluctuation is estimated and compared with the results from power balance estimation.
International Nuclear Information System (INIS)
The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution
Indian Academy of Sciences (India)
Nizami Gasilov
2007-04-01
In designing tokamaks, the maintenance of vertical stability of plasma is one of the most important problems. Systems of the passive and active feedbacks are applied for this purpose. Role of the passive system consisting of a vacuum vessel and passive coils is to suppress fast MHD (magnetohydrodynamic) instabilities. The active feedback system is applied to control slow motions of plasma. The objective of the paper is to investigate two successive problems, solution of which allows to determine the possibility of controlling plasma motions. One of these is the problem of vertical stability under the assumption of ideal conductivity of plasma and passive stabilizing elements. The problem is solved analytically and on the basis of the obtained solution a criterion of MHD-stability is formulated. The other problem is connected with the control of plasma vertical position with active feedback system. Calculation of feedback control parameters is formulated as an optimization problem and an approximate method to solve the problem is suggested. Numerical simulations are performed with parameters of the T-15M tokamak in order to justify the suggested method.
International Nuclear Information System (INIS)
One of the most important issues for magnetic-confinement fusion research is the so-called anomalous transport across magnetic field lines, i.e. transport that is in excess of that caused by collisional processes. The need to reduce anomalous transport in order to increase the efficiency of a prospective fusion reactor must be addressed through an investigation of its fundamental underlying causes. This thesis is divided into two distinct components: one experimental and instrumental, and the other theoretical and based on numerical modeling. The experimental part consists of the design and installation of a new diagnostic for core turbulence fluctuations in the TCV tokamak. An extensive conceptual investigation of a number of possible solutions, including Beam Emission Spectroscopy, Reflectometry, Cross Polarization, Collective Scattering and different Imaging techniques, was carried out at first. A number of criteria, such as difficulties in data interpretation, costs, variety of physics issues that could be addressed and expected performance, were used to compare the different techniques for specific application to the TCV tokamak. The expected signal to noise ratio and the required sampling frequency for TCV were estimated on the basis of a large number of linear, local gyrokinetic simulations of plasma fluctuations. This work led to the choice of a Zernike phase contrast imaging system in a tangential launching configuration. The diagnostic was specifically designed to provide information on turbulence features up to now unknown. In particular, it is characterized by an outstanding spatial resolution and by the capability to measure a very broad range of fluctuations, from ion to electron Larmor radius scales, thus covering the major part of the instabilities expected to be at play in TCV. The spectrum accessible covers the wavenumber region from 0.9 cm-1 to 60 cm-1 at 24 radial positions with 3 MHz bandwidth. The diagnostic is an imaging technique and is
A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma
Energy Technology Data Exchange (ETDEWEB)
Ku, S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Hager, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Chang, C. S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kwon, J. M. [National Fusion Research Institute, Republic of Korea; Parker, S. E. [University of Colorado Boulder, USA
2016-06-01
In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles provide scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. The numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.
Energetic-particle-induced electromagnetic geodesic acoustic mode in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Wang, Lingfeng, E-mail: wanglf@swip.ac.cn; He, Zhixiong; He, Hongda; Shen, Y. [Southwestern Institute of Physics, Chengdu 610041 (China); Dong, J. Q. [Institute for Fusion Theory and Simulation, Zhejiang University, Hangzhou 310027 (China); Southwestern Institute of Physics, Chengdu 610041 (China)
2014-07-15
Energetic-particle-induced kinetic electromagnetic geodesic acoustic modes (EKEGAMs) are numerically studied in low β (=plasma pressure/magnetic pressure) tokamak plasmas. The parallel component of the perturbed vector potential is considered along with the electrostatic potential perturbation. The effects of finite Larmor radius and finite orbit width of the bulk and energetic ions as well as electron parallel dynamics are all taken into account in the dispersion relation. Systematic harmonic and ordering analysis are performed for frequency and growth rate spectra of the EKEGAMs, assuming (kρ{sub i})∼q{sup −3}∼β≪1, where q, k, and ρ{sub i} are the safety factor, radial component of the EKEGAMs wave vector, and the Larmor radius of the ions, respectively. It is found that there exist critical β{sub h}/β{sub i} values, which depend, in particular, on pitch angle of energetic ions and safety factor, for the mode to be driven unstable. The EKEGAMs may also be unstable for pitch angle λ{sub 0}B<0.4 in certain parameter regions. Finite β effect of the bulk ions is shown to have damping effect on the EKEGAMs. Modes with higher radial wave vectors have higher growth rates. The damping from electron dynamics is found decreasing with decrease of the temperature ratio T{sub e}/T{sub i}. The modes are easily to be driven unstable in low safety factor q region and high temperature ratio T{sub h}/T{sub i} region. The harmonic features of the EKEGAMs are discussed as well.
A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma
Ku, S.; Hager, R.; Chang, C. S.; Kwon, J. M.; Parker, S. E.
2016-06-01
In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles provide scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation - e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others - can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function - driven by ionization, charge exchange and wall loss - is allowed to be arbitrarily large. The numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.
Institute of Scientific and Technical Information of China (English)
无
2005-01-01
High-pressure gas injection has proved to be an effective disruption mitigation technique in DⅢ-D tokamak experiments. If the method can be applied in future tokamak reactors not only for disruption mitigation but also for plasma termination and fueling, it will have an attractive advantage over the pellet and liquid injection from the viewpoint of economy and engineering design. In order to investigate the feasibility of this option, a study has been carried out with relevant parameters for conveying tubes of different geometrical sizes and for different gases.These parameters include pressure drop, lagger time after the valve's opening, gas diffusion in an ultra-high vacuum condition, and particle number contour.
Modeling of electromagnetic fields during plasma startup in SST-1 tokamak
International Nuclear Information System (INIS)
The time varying currents in the Ohmic transformer in SST-1 tokamak induce large eddy currents in the passive structures like the vacuum vessel and cryostat. Especially since the vacuum vessel and the cryostat are toroidally continuous without breaks in SST-1, this leads to a shielding effect on the flux penetrating the vacuum vessel. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. Also the induced currents alter the null location of magnetic field. Studying the effective loop voltage and magnetic null location during the plasma breakdown and startup is important, as corrective measures may be required in case of an insufficient loop voltage or an improper null. The dynamics of the evolution of the loop voltage and the magnetic null due to the toroidal eddy currents in SST-1 passive structure has been studied in the breakdown phase of SST-1. At the time of the plasma initiation, the Ohmic transformer current is discharged by short-circuiting the central solenoid (CS) coil through a resistance. The flux stored in the CS coil is linked to the plasma region, as also the conductors surrounding the plasma region. The resulting eddy currents flowing in the passive conductors lead to Joule heating losses of the stored flux in the CS coil. The amount of this eddy current and the associated flux loss has to be accurately determined in order to estimate the external loop voltage seen by the plasma required for plasma breakdown and current ramp-up. We have studied the effect of the induced currents on the loop voltage and the magnetic null using a toroidal-filament model. As the vessel and cryostat are conductors with large poloidal cross-section, for the approximation to be valid and results to be accurate, they are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green functions and the
Runaway electrons and plasma turbulence in current ramping by lower hybrid waves in tokamak plasmas
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The role of runaway electrons in current ramping by lower hybrid waves is discussed. The back emf induced by current ramping using lower hybrid waves produces runaway electrons at such a rate that the rate of change of the current carried by runaways can completely compensate the rf current ramp rate if runaways are well-confined. However, runaway electrons can also destabilize the oblique plasma waves due to their anisotropy. The resulting turbulence can greatly reduce the run-away production rate, enabling the current ramping rate to exceed the above limit
Runaway electrons and plasma turbulence in current ramping by lower hybrid waves in tokamak plasmas
International Nuclear Information System (INIS)
The role of runaway electrons in current ramping by lower hybrid waves is discussed. The back emf induced by current ramping using lower hybrid waves produces runaway electrons at such a rate that the rate of change of the current carried by runaways can completely compensate the RF current ramp rate if runaways are well-confined. However, runaway electrons can also destabilize the oblique plasma waves due to their anisotropy. The resulting turbulence can greatly reduce the runaway production rate, enabling the current ramping rate to exceed the above limit
Chirkov, A. Yu.
2015-09-01
Low gain (Q ~ 1) fusion plasma systems are of interest for concepts of fusion-fission hybrid reactors. Operational regimes of large modern tokamaks are close to Q ≈ 1. Therefore, they can be considered as prototypes of neutron sources for fusion-fission hybrids. Powerful neutral beam injection (NBI) can support the essential population of fast particles compared with the Maxwellial population. In such two-component plasma, fusion reaction rate is higher than for Maxwellian plasma. Increased reaction rate allows the development of relatively small-size and relatively inexpensive neutron sources. Possible operating regimes of the NBI-heated tokamak neutron source are discussed. In a relatively compact device, the predictions of physics of two-component fusion plasma have some volatility that causes taking into account variations of the operational parameters. Consequent parameter ranges are studied. The feasibility of regimes with Q ≈ 1 is shown for the relatively small and low-power system. The effect of NBI fraction in total heating power is analyzed.
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This report contains the papers delivered at the AEB - Natal University summer school on plasma physics held in Durban during January 1979. The following topics were discussed: Tokamak devices; MHD stability; trapped particles in tori; Tokamak results and experiments; operating regime of the AEB Tokamak; Tokamak equilibrium; high beta Tokamak equilibria; ideal Tokamak stability; resistive MHD instabilities; Tokamak diagnostics; Tokamak control and data acquisition; feedback control of Tokamaks; heating and refuelling; neutral beam injection; radio frequency heating; nonlinear drift wave induced plasma transport; toroidal plasma boundary layers; microinstabilities and injected beams and quasilinear theory of the ion acoustic instability
Material migration in tokamak plasmas with a three-dimensional boundary
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In this work, the influence of a 3D boundary induced by resonant magnetic perturbations (RMPs) on the material migration, i.e. the erosion from wall material due to the plasma surface interaction and the transport of these impurities is investigated for the first time. With applied RMPs two new domains occur in the magnetic field structure. Three dimensional SOL flux tubes with predominantly transport parallel to short magnetic field lines and a region of longer stochastic field lines with diffusive gradient driven radial transport. The plasma wall interaction and the material transport in these domains were investigated. A globally higher radial electric field Er as well as local changes in the magnetic field structure such as pressure driven sonic flows or locally higher Er fields can potentially influence the material transport with applied RMPs. The experiments were performed at the tokamak TEXTOR, the RMPs were induced by the dynamic ergodic divertor (DED). The plasma discharges and DED application was chosen to have a spatially separated 3D structure to be able to investigate the underlying physics. Two spherical carbon test limiters were positioned in different poloidal and toroidal positions which allowed to analyse the material migration in a 3D SOL flux tube and a stochastic region at the same time. Methane doped with 13C was injected through the test limiters during three different plasma scenarios, without RMPs, with static RMPs and an RMP sweep. The test limiters and the injected methane were monitored in situ with different cameras and spectrometers. The deposition of the injected particles was measured post mortem by colourimetry, nuclear reaction analysis and secondary ion mass spectrometry. The most profound change from no RMP to the RMP cases is a 90 re-direction of the low ionised carbon C+ and C2+ into the Er x B-drift direction. From a comparison of the experiments and numerical field line tracing it was found that this is a global effect
First-principle description of collisional gyrokinetic turbulence in tokamak plasmas
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This dissertation starts in chapter 1 with a comprehensive introduction to nuclear fusion, its basic physics, goals and means. It especially defines the concept of a fusion plasma and some of its essential physical properties. The following chapter 2 discusses some fundamental concepts of statistical physics. It introduces the kinetic and the fluid frameworks, compares them and highlights their respective strengths and limitations. The end of the chapter is dedicated to the fluid theory. It presents two new sets of closure relations for fluid equations which retain important pieces of physics, relevant in the weakly collisional tokamak regimes: collective resonances which lead to Landau damping and entropy production. Nonetheless, since the evolution of the turbulence is intrinsically nonlinear and deeply influenced by velocity space effects, a kinetic collisional description is most relevant. First focusing on the kinetic aspect, chapter 3 introduces the so-called gyrokinetic framework along with the numerical solver - the GYSELA code - which will be used throughout this dissertation. Very generically, code solving is an initial value problem. The impact on turbulent nonlinear evolution of out of equilibrium initial conditions is discussed while studying transient flows, self-organizing dynamics and memory effects due to initial conditions. This dissertation introduces an operational definition, now of routine use in the GYSELA code, for the initial state and concludes on the special importance of the accurate calculation of the radial electric field. The GYSELA framework is further extended in chapter 4 to describe Coulomb collisions. The implementation of a collision operator acting on the full distribution function is presented. Its successful confrontation to collisional theory (neoclassical theory) is also shown. GYSELA is now part of the few gyrokinetic codes which can self-consistently address the interplay between turbulence and collisions. While
Plasma rotation and NTM onset driven by central EC deposition in TCV tokamak
Energy Technology Data Exchange (ETDEWEB)
Nowak, S.; Lazzaro, E. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Sauter, O.; Canal, G.; Duval, B.; Federspiel, L.; Karpushov, A. N.; Kim, D.; Reimerders, H.; Rossel, J.; Testa, D.; Wagner, D. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association EURATOM-Confederation Suisse, 1015 Lausanne (Switzerland); Raju, D. [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Collaboration: TCV Team
2014-02-12
The effects of the central electron cyclotron heating (ECH) and current drive (ECCD) on the spontaneous plasma rotation and on the presence of Tearing Modes (TM), observed in the TCV tokamak[1], were recently investigated as an interplay between the toroidal velocity and NTM onset in absence of sawteeth, ELMs and error fields [2–3]. In a set of reproducible TCV discharges (I{sub p}∼ −150 kA, B{sub t}∼ −1.4 T, ne,{sub av∼} 1.5 10{sup 19} m{sup −3}, T{sub e}∼ 3 keV and T{sub i}∼0.25 keV, q{sub 95}∼5.8) with both pure EC heating and current drive the cnt-Ip toroidal velocity was observed to be reduced with subsequent co-Ip appearance of 3/2 and 2/1 modes during the ramp up EC phases. The understanding of the capability of the on-axis EC power to modify the rotation profiles before and after the TM onset and of the sudden disappearance of 3/2 mode when 2/1 starts is the main purpose of this work. The velocity profile modifications are due to a direct effect of the EC absorbed power and also related to some variation of the perpendicular diffusion of the toroidal momentum and to magnetic braking effects of the kind of neoclassical toroidal viscosity (NTV) due to the NTM resonant field perturbations associated to the presence of TM. Numerical investigations are performed using a 1D toroidal momentum balance equation including contributions by external sources, as EC power, and NTV torques. Furthermore, the combined evolution of the 3/2 and 2/1 modes requires considering also coupling effects included in a generalized Rutherford equation for the modelling of the TM time growth.
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The development of nuclear fusion as an alternative energy source requires the research on magnetically confined, high temperature plasmas. In particular, the quantification of plasma flows in the domain near exposed material surfaces of the plasma container by computer simulations is of key importance, both for guiding interpretation of present fusion experiments and for aiding the ongoing design activities for large future devices such as ITER, W7-X or the DEMO reactor. There is a large number of computational issues related to the physics of hot, fully ionized and magnetized plasmas near surfaces of the vacuum chamber. This thesis is dedicated to one particular such challenge, namely the numerical quantification of self-consistent kinetic neutral gas and plasma fluid flows in very complex 3D (partially chaotic) magnetic fields, in the absence of any common symmetries for plasma and neutral gas dynamics. Such magnetic field configurations are e.g. generated by externally applied magnetic perturbations at the plasma edge, and are of great interest for the control of particle and energy exhausts. In the present thesis the 3D edge plasma and neutral particle transport code EMC3-EIRENE is applied to two distinct configurations of open chaotic magnetic system: at the TEXTOR and DIII-D tokamaks. Improvements of the edge transport model and extensions of the transport code are presented, which have allowed such simulations for the first time for 3D scenarios at DIII-D with ITER similar plasmas. A strong 3D effect of the chaotic magnetic field on the DIII-D edge plasma is found and analyzed in detail. It is found that a pronounced striation pattern of target particle and heat fluxes at DIII-D can only be obtained up to a certain upper limiting level of anomalous cross-field transport. Hence, in comparison to experimental data, these findings allow to narrow down the range of this model parameter. One particular interest at TEXTOR is the achievement of a regime with
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Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage