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Sample records for tokamak plasma transport

  1. 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).

  2. Mathematical modeling plasma transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1997-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 1020/m3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  3. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  4. Plasma transport in a compact ignition tokamak

    International Nuclear Information System (INIS)

    Singer, C.E.; Ku, L.P; Bateman, G.

    1987-02-01

    Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds 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 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

  5. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-09-01

    Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr

  6. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-06-01

    The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr

  7. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

    Hirshman, S.P.; Sigmar, D.J.

    1981-05-01

    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/e 2 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

  8. Transport Bifurcation in a Rotating Tokamak Plasma

    International Nuclear Information System (INIS)

    Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

    2010-01-01

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  9. TFTR/JET INTOR workshop on plasma transport tokamaks

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-01-01

    This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included

  10. Transport in the tokamak plasma edge

    International Nuclear Information System (INIS)

    Vold, E.L.

    1989-01-01

    Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

  11. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

    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) [fr

  12. Numerical studies of transport processes in Tokamak plasma

    International Nuclear Information System (INIS)

    Spineanu, F.; Vlad, M.

    1984-09-01

    The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included

  13. Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport

    International Nuclear Information System (INIS)

    Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.; Cowley, S. C.

    2011-01-01

    The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.

  14. Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)

    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.

  15. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

    The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

  16. Two-dimensional transport of tokamak plasmas

    International Nuclear Information System (INIS)

    Hirshman, S.P.; Jardin, S.C.

    1979-01-01

    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

  17. A quasi-linear gyrokinetic transport model for tokamak plasmas

    International Nuclear Information System (INIS)

    Casati, A.

    2009-10-01

    After a presentation of some basics around nuclear fusion, this research thesis introduces the framework of the tokamak strategy to deal with confinement, hence the main plasma instabilities which are responsible for turbulent transport of energy and matter in such a system. The author also briefly introduces the two principal plasma representations, the fluid and the kinetic ones. He explains why the gyro-kinetic approach has been preferred. A tokamak relevant case is presented in order to highlight the relevance of a correct accounting of the kinetic wave-particle resonance. He discusses the issue of the quasi-linear response. Firstly, the derivation of the model, called QuaLiKiz, and its underlying hypotheses to get the energy and the particle turbulent flux are presented. Secondly, the validity of the quasi-linear response is verified against the nonlinear gyro-kinetic simulations. The saturation model that is assumed in QuaLiKiz, is presented and discussed. Then, the author qualifies the global outcomes of QuaLiKiz. Both the quasi-linear energy and the particle flux are compared to the expectations from the nonlinear simulations, across a wide scan of tokamak relevant parameters. Therefore, the coupling of QuaLiKiz within the integrated transport solver CRONOS is presented: this procedure allows the time-dependent transport problem to be solved, hence the direct application of the model to the experiment. The first preliminary results regarding the experimental analysis are finally discussed

  18. Electromagnetic effects on trace impurity transport in tokamak plasmas

    Science.gov (United States)

    Hein, T.; Angioni, C.

    2010-01-01

    The impact of electromagnetic effects on the transport of light and heavy impurities in tokamak plasmas is investigated by means of an extensive set of linear gyrokinetic numerical calculations with the code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and of analytical derivations with a fluid model. The impurity transport is studied by appropriately separating diffusive and convective contributions, and conditions of background microturbulence dominated by both ion temperature gradient (ITG) and trapped electron modes (TEMs) are analyzed. The dominant contribution from magnetic flutter transport turns out to be of pure convective type. However it remains small, below 10% with respect to the E ×B transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter β is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high β plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing β in qualitative agreement with recent observations in tokamaks. In contrast, in these conditions, the ratio of the total off-diagonal convective velocity to the diagonal diffusivity is not strongly affected by an increase in β, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.

  19. The use of internal transport barriers in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Challis, C D [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

    2004-12-01

    Internal transport barriers (ITBs) can provide high tokamak confinement at modest plasma current. This is desirable for operation with most of the current driven non-inductively by the bootstrap mechanism, as currently envisaged for steady-state power plants. Maintaining such plasmas in steady conditions with high plasma purity is challenging, however, due to MHD instabilities and impurity transport effects. Significant progress has been made in the control of ITB plasmas: the pressure profile has been varied using the barrier location; q-profile modification has been achieved with non-inductive current drive, and means have been found to affect density peaking and impurity accumulation. All these features are, to some extent, interdependent and must be integrated self-consistently to demonstrate a sound basis for extrapolation to future devices.

  20. Tokamak transport phenomenology and plasma dynamic response

    International Nuclear Information System (INIS)

    Moret, J.M.; Association Euratom CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

    1991-07-01

    A system identification method is developed to estimate the transfer function of the system from the time evolution of its parameters to any excitation. The form of the identified transfer function is linked to a representation of the transport in terms of poles (eigenvalues) and eigenmodes. These eigenvalues and eigenvectors are thus directly deduced from the raw data with no restriction on the underlying processes and there is consequently no need to adjust any simplified transport model to the experimental data. This method is illustrated in this paper by analysing the injection of pellets on Tore Supra. The density and the temperature transfer functions were observed to share the same poles with the corresponding eigenmodes grouped in pairs with identical profiles. This implies the presence of a coupling between the particle and heat flow. A criterion is developed to select amongst the possible coupling mechanisms, based on compatibility with the observed transfer function. The selection suggests a model in which the particle diffusion coefficient depends on the density and on the temperature gradient

  1. Internal transport barrier in tokamak and helical plasmas

    Science.gov (United States)

    Ida, K.; Fujita, T.

    2018-03-01

    The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the

  2. Anomalous energy transport in hot plasmas: solar corona and Tokamak

    International Nuclear Information System (INIS)

    Beaufume, P.

    1992-04-01

    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

  3. Perturbative transport experiments in JET Advanced Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Mantica, P.; Gorini, G.; Sozzi, C. [Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan (Italy); Imbeaux, F.; Sarazin, Y.; Garbet, X. [Association Euratom-CEA, St. Paul-lez-Durance Cedex (France); Kinsey, J. [Lehigh Univ., Bethlehem, Pennsylvania (United States); Budny, R. [Princeton Plasma Physics Lab, New Jersey (United States); Coffey, I.; Parail, V.; Walden, A. [Euratom/UKAEA Fusion Association, Abingdon, Oxon (United Kingdom); Dux, R. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Garzotti, L. [Istituto Gas Ionizzati, Padova (Italy); Ingesson, C. [FOM-Instituut voor Plasmafysica, Nieuwegein (Netherlands); Kissick, M. [University of California, Los Angeles (United States)

    2003-07-01

    Perturbative transport experiments have been performed in JET Advanced Tokamak plasmas either in conditions of fully developed Internal Transport Barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced by either Laser Ablation or Shallow Pellet Injection and the ensuing travelling cold pulse was used to probe the plasma transport in the electron and, for the first time, also in the ion channel. Cold pulses travelling through ITBs are observed to erode the ITB outer part, but, if the inner ITB portion survives, it strongly damps the propagating wave. The result is discussed in the context of proposed possible pictures for ITB formation. In the absence of an ITB, the cold pulse shows a fast propagation in the outer plasma half, which is consistent with a region of stiff transport, while in the inner half it slows down but shows the peculiar feature of amplitude growing while propagating. The data are powerful tests for the validation of theoretical transport models. (author)

  4. Two dimensional neutral transport analysis in tokamak plasma

    International Nuclear Information System (INIS)

    Shimizu, Katsuhiro; Azumi, Masafumi

    1987-02-01

    Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)

  5. Nonlinear transport processes in tokamak plasmas. I. The collisional regimes

    International Nuclear Information System (INIS)

    Sonnino, Giorgio; Peeters, Philippe

    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 Joint European Torus (JET) plasmas are also reported. It is found that the nonlinear electron P-S transport coefficients exceed the values provided by neoclassical theory by a factor that may be of the order 10 2 . The nonlinear classical coefficients exceed the neoclassical ones by a factor that may be of order 2. For JET, the discrepancy between experimental and theoretical results for the electron losses is therefore significantly reduced by a factor 10 2 when the nonlinear contributions are duly taken into account but, there is still a factor of 10 2 to be explained. This is most likely due to turbulence. The expressions of the ion transport coefficients, determined by the neoclassical theory in these two regimes, remain unaltered. The low-collisional regimes, i.e., the plateau and the banana regimes, are analyzed in the second part of this work

  6. Plasma rotation and transport in MAST spherical tokamak

    Science.gov (United States)

    Field, A. R.; Michael, C.; Akers, R. J.; Candy, J.; Colyer, G.; Guttenfelder, W.; Ghim, Y.-c.; Roach, C. M.; Saarelma, S.; MAST Team

    2011-06-01

    The formation of internal transport barriers (ITBs) is investigated in MAST spherical tokamak plasmas. The relative importance of equilibrium flow shear and magnetic shear in their formation and evolution is investigated using data from high-resolution kinetic- and q-profile diagnostics. In L-mode plasmas, with co-current directed NBI heating, ITBs in the momentum and ion thermal channels form in the negative shear region just inside qmin. In the ITB region the anomalous ion thermal transport is suppressed, with ion thermal transport close to the neo-classical level, although the electron transport remains anomalous. Linear stability analysis with the gyro-kinetic code GS2 shows that all electrostatic micro-instabilities are stable in the negative magnetic shear region in the core, both with and without flow shear. Outside the ITB, in the region of positive magnetic shear and relatively weak flow shear, electrostatic micro-instabilities become unstable over a wide range of wave numbers. Flow shear reduces the linear growth rates of low-k modes but suppression of ITG modes is incomplete, which is consistent with the observed anomalous ion transport in this region; however, flow shear has little impact on growth rates of high-k, electron-scale modes. With counter-NBI ITBs of greater radial extent form outside qmin due to the broader profile of E × B flow shear produced by the greater prompt fast-ion loss torque.

  7. Orbit effects on impurity transport in a rotating tokamak plasma

    International Nuclear Information System (INIS)

    Wong, K.L.; Cheng, C.Z.

    1988-05-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs

  8. Double internal transport barrier triggering mechanism in tokamak plasmas

    International Nuclear Information System (INIS)

    Dong, Jiaqi; Mou, Zongze; Long, Yongxing; Mahajan, Swadesh M.

    2004-01-01

    Sheared flow layers created by energy released in magnetic reconnection processes are studied with the magneto hydrodynamics (MHD), aimed at internal transport barrier (ITB) dynamics. The double tearing mode induced by electron viscosity is investigated and proposed as a triggering mechanism for double internal transport barrier (DITB) observed in tokamak plasmas with non-monotonic safety factor profiles. The quasi-linear development of the mode is simulated and the emphasis is placed on the structure of sheared poloidal flow layers formed in the vicinity of the magnetic islands. For viscosity double tearing modes, it is shown that the sheared flows induced by the mode may reach the level required by the condition for ITB formation. Especially, the flow layers are found to form just outside the magnetic islands. The scaling of the generated velocity with plasma parameters is given. Possible explanation for the experimental observations that the preferential formation of transport barriers in the proximity of low order rational surface is discussed. (author)

  9. Electron cyclotron absorption in Tokamak plasmas in the presence of radial transport of particles

    International Nuclear Information System (INIS)

    Rosa, Paulo R. da S.; Ziebell, Luiz F.

    1998-01-01

    We use quasilinear theory to study effects of particle radial transport on the electron cyclotron absorption coefficient by a current carrying plasma, in a tokamak modelated as a plasma slab. Our numerical results indicate significant modification in the profile of the electron cyclotron absorption coefficient when transport is taken into account relative to the situation without transport. (author)

  10. Plasma flow and transport on the tokamak ISTTOK boundary plasma

    International Nuclear Information System (INIS)

    Figueiredo, H.; Silva, C.; Goncalves, B.; Duarte, P.; Fernandes, H.

    2011-01-01

    The ISTTOK boundary plasma velocity near the outer midplane is measured on the parallel and perpendicular directions in four different configurations by reversing independently the toroidal magnetic field and the plasma current directions. The parallel flow is found to not depend significantly on both the toroidal magnetic field and plasma current directions, being always directed towards the nearest limiter in the scrape-off layer. On the contrary, the perpendicular flow is found to follow the E r x B drift direction. The poloidal velocity has also been derived from the correlation of floating potential signals measured on poloidally separated probes and a good agreement with the value derived with the Gundestrup probe is found. Finally, the dynamical interplay between parallel momentum and turbulent particle flux has been investigated and a clear dynamical coupling between these quantities is found in the region inside the limiter.

  11. Dynamics of the edge transport barrier at plasma biasing on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Spolaore, M.; Peleman, P.; Brotánková, Jana; Horáček, Jan; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Kocan, M.; Martines, E.; Pánek, Radomír; Sharma, A.; Van Oost, G.

    2006-01-01

    Roč. 12, č. 6 (2006), s. 19-23 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * transport barrier * relaxations Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html

  12. Positron deposition in plasmas by positronium beam ionization and transport of positrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Murphy, T.J.

    1986-11-01

    In a recently proposed positron transport experiment, positrons would be deposited in a fusion plasma by forming a positronium (Ps) beam and passing it through the plasma. Positrons would be deposited as the beam is ionized by plasma ions and electrons. Radial transport of the positrons to the limiter could then be measured by detecting the gamma radiation produced by annihilation of positrons with electrons in the limiter. This would allow measurements of the transport of electron-mass particles and might shed some light on the mechanisms of electron transport in fusion plasmas. In this paper, the deposition and transport of positrons in a tokamak are simulated and the annihilation signal determined for several transport models. Calculations of the expected signals are necessary for the optimal design of a positron transport experiment. There are several mechanisms for the loss of positrons besides transport to the limiter. Annihilation with plasma electrons and reformation of positronium in positron-hydrogen collisions are two such processes. These processes can alter the signal and place restrictions ons on the plasma conditions in which positron transport experiments can be effectively performed

  13. Electron heat transport in current carrying and currentless thermonuclear plasmas. Tokamaks and stellarators compared

    International Nuclear Information System (INIS)

    Peters, M.

    1996-01-01

    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.)

  14. Electron heat transport in current carrying and currentless thermonuclear plasmas. Tokamaks and stellarators compared

    Energy Technology Data Exchange (ETDEWEB)

    Peters, M

    1996-01-16

    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 {chi} 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.).

  15. The critical temperature gradient model of plasma transport: applications to Jet and future tokamaks

    International Nuclear Information System (INIS)

    Rebut, P.H.; Lallia, P.P.; Watkins, M.L.

    1989-01-01

    The diversity and complexity of behaviour in tokamak plasmas place strong constraints on any model attempting a description in terms of a single underlying phenomenon. Assuming that turbulence in the magnetic topology is the underlying phenomenon, specific expressions for electron and ion heat flux are derived from heuristic and dimensional arguments. When used in plasma transport codes, rather satisfactory simulations of experimental results are achieved in different sized tokamaks in various regimes of operation. Predictions are given for the expected performance of JET at full planned power and implications for next step tokamaks are indicated

  16. Self-consistent treatment of transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Wilhelmsson, H.

    1993-01-01

    A theory is developed for the dynamics of tokamak plasmas considering the influence of combinations of simultaneous heating processes (alpha particle, auxiliary and ohmic), thermal conduction and particle diffusion, thermal and particle pinches, thermalization of alpha particles as well as the effects of boundary conditions. The analysis is based on a generalization of the central expansion technique which transforms the partial differential equations to a set of nonlinear coupled equations in time for the dynamic variables. Oscillatory solutions are found, but only in the presence of alpha particle heating. Examples of extensive computer simulations are included which support and complete the analytic results. (26 refs.)

  17. Effect of density control and impurity transport on internal transport barrier formation in tokamak plasma

    International Nuclear Information System (INIS)

    Yamakami, Tomoyuki; Fujita, Takaaki; Arimoto, Hideki; Yamazaki, Kozo

    2014-01-01

    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)

  18. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.

    2001-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  19. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Von Goeler, S.; Houlberg, W.A.

    1999-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  20. Finite beta effects on turbulent transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Hein, Tobias

    2011-01-01

    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.

  1. Extended fluid transport theory in the tokamak plasma edge

    Science.gov (United States)

    Stacey, W. M.

    2017-06-01

    Fluid theory expressions for the radial particle and energy fluxes and the radial distributions of pressure and temperature in the edge plasma are derived from fundamental conservation (particle, energy, momentum) relations, taking into account kinetic corrections arising from ion orbit loss, and integrated to illustrate the dependence of the observed edge pedestal profile structure on fueling, heating, and electromagnetic and thermodynamic forces. Solution procedures for the fluid plasma and associated neutral transport equations are discussed.

  2. The simulation of L-H transition in tokamak plasma using MMM95 transport model

    International Nuclear Information System (INIS)

    Intharat, P; Poolyarat, N; Chatthong, B; Onjun, T; Picha, R

    2015-01-01

    BALDUR integrative predictive modelling code together with a Multimode (MMM95) anomalous transport model is used to simulate the evolution profiles, including plasma current, temperature, density and energy in a tokamak reactor. It is found that a self - transition from low confinement mode (L-mode) to high confinement mode (H-mode) regimes can be achieved once a sufficient auxiliary heating applied to the plasma is reached. The result agrees with experimental observations from various tokamaks. A strong reduction of turbulent transport near the edge of plasma is also observed, which is related to the formation of steep radial electric field near the edge regime. From transport analysis, it appears that the resistive ballooning mode is the dominant term near the plasma edge regime, which is significantly reduced during the transition. (paper)

  3. Combined core/boundary layer plasma transport simulations in tokamaks

    International Nuclear Information System (INIS)

    Prinja, A.K.; Schafer, R.F. Jr.; Conn, R.W.; Howe, H.C.

    1987-01-01

    Significant new numerical results are presented from self-consistent core and boundary or scrape-off layer plasma simulations with 3-D neutral transport calculations. For a symmetric belt limiter it is shown that, for plasma conditions considered here, the pump limiter collection efficiency increases from 11% to 18% of the core efflux as a result of local reionization of blade deflected neutrals. This hitherto unobserved effect causes a significant amplification of upstream ion flux entering the pump limiter. Results from coupling of an earlier developed two-zone edge plasma model ODESSA to the PROCTR core plasma simulation code indicates that intense recycling divertor operation may not be possible because of stagnation of upstream flow velocity. This results in a self-consistent reduction of density gradient in an intermediate region between the central plasma and separatrix, and a concomitant reduction of core-efflux. There is also evidence of increased recycling at the first wall. (orig.)

  4. Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

    2018-02-01

    A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

  5. Effects of density asymmetries on heavy-impurity transport in a rotating tokamak-plasma

    International Nuclear Information System (INIS)

    Romanelli, M.; Ottaviani, M.

    1997-12-01

    The transport equations of heavy trace-impurities in a Tokamak plasma with strong toroidal rotation have been studied analytically in the collisional regime. It is found that the poloidal asymmetry of the impurity-density, which occurs because of the rotation, brings about a large enhancement of the diffusivity and indeed of the pinch velocity above the conventional Pfirsh-Schlueter values. (author)

  6. Electron thermal transport barrier and magnetohydrodynamic activity observed in Tokamak plasmas with negative central shear

    NARCIS (Netherlands)

    M.R. de Baar,; Hogeweij, G. M. D.; Cardozo, N. J. L.; Oomens, A. A. M.; Schüller, F. C.

    1997-01-01

    In the Rijnhuizen Tokamak Project, plasmas with steady-state negative central shear (NCS) are made with off-axis electron cyclotron heating. Shifting the power deposition by 2 mm results in a sharp transition of confinement. The good confinement branch features a transport barrier at the off-axis

  7. Influence of plasma shape on transport in the TCV tokamak

    International Nuclear Information System (INIS)

    Moret, J.M.; Franke, S.; Weisen, H.

    1996-11-01

    The energy confinement time of TCV ohmic, L-mode plasmas is observed to depend on the shape, improving slightly with elongation and degrading strongly with positive triangularity. This dependence can be explained by combination of geometrical effects on the temperature gradient and power degradation, without invoking a shape dependence of the transport coefficients. (author) 5 figs., 8 refs

  8. Transport and stability studies in negative central shear advanced tokamak plasmas

    International Nuclear Information System (INIS)

    Jayakumar, R.J.

    2003-01-01

    Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q min values. (orig.)

  9. Transport and stability studies in negative central shear advanced tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, R.J. [Lawrence Livermore National Laboratory (United States)

    2003-07-01

    Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q{sub min} values. (orig.)

  10. Fast ions and momentum transport in JET tokamak plasmas

    International Nuclear Information System (INIS)

    Salmi, A.

    2012-01-01

    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.)

  11. 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.)

  12. Impurity transport in a collision-dominated rotating tokamak plasma

    International Nuclear Information System (INIS)

    Eriksson, G.; Liljegren, A.

    1981-04-01

    The flux of heavy impurities is an axisymmetric, toroidal plasma with all particles in the collision-dominated regime is considered. Plasma rotation and charge-exchange with neutrals are taken into account. A hydrodynamic model employing Braginskii's transport equations is used. The theorry is extended to higher collision freqencies as compared to previous treatments. It is found that the Pfirsch-Schlueter flux is significantly reduced as compared to the value given by Rutherford and that it is of the same order of magnitude, or less, than the classical flux in all regimes considered. It is also shown that the impurity flux can be influenced by charge-exchange with neutrals. (author)

  13. Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

    2009-01-01

    Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

  14. Observation of internal transport barrier in ELMy H-mode plasmas on the EAST tokamak

    Science.gov (United States)

    Yang, Y.; Gao, X.; Liu, H. Q.; Li, G. Q.; Zhang, T.; Zeng, L.; Liu, Y. K.; Wu, M. Q.; Kong, D. F.; Ming, T. F.; Han, X.; Wang, Y. M.; Zang, Q.; Lyu, B.; Li, Y. Y.; Duan, Y. M.; Zhong, F. B.; Li, K.; Xu, L. Q.; Gong, X. Z.; Sun, Y. W.; Qian, J. P.; Ding, B. J.; Liu, Z. X.; Liu, F. K.; Hu, C. D.; Xiang, N.; Liang, Y. F.; Zhang, X. D.; Wan, B. N.; Li, J. G.; Wan, Y. X.; EAST Team

    2017-08-01

    The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ safety factor q(0) ˜ 1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.

  15. Relaxation oscillations and transport barrier dynamics in tokamak edge plasmas

    International Nuclear Information System (INIS)

    Benkadda, Sadruddin; Beyer, Peter; Fuhr-Chaudier, Guillaume; Garbet, Xavier; Ghendrih, Philippe; Sarazin, Yanick

    2004-01-01

    Oscillations of turbulent transport of particles and energy in magnetically confined plasmas can be easily observed in simulations of a variety of turbulence models. These oscillations typically involve a mechanism of energy exchange between fluctuations and a poloidal shear flow. This kind of ''predator-prey'' mechanism is found to be not relevant for transport barrier relaxations. In RBM simulations of resistive ballooning turbulence with transport barrier, relaxation oscillations of the latter are observed even in the case of frozen poloidal shear flow. These relaxations are due to a transitory growth of a mode localized at the barrier center. A one-dimensional model for the evolution of such a mode in the presence of a shear flow describes a transitory growth of an initial perturbation. Oscillations in the case of a finite steady-state shear flow are possible due to the coupling of the mode to the dynamics of the pressure profile. (author)

  16. Development of plasma diagnostics technologies - Measurement of transport= parameters in tokamak edge plasma by using electric transport probes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kyu Sun; Chang, Do Hee; Sim, Yeon Gun; Kim, Jin Hee [Hanyang University, Seoul (Korea, Republic of)

    1995-08-01

    Electric transport probe system is developed for the measurement of electron temperature, floating potential, plasma density and flow velocity of= edge plasmas in the KT-2 medium size tokamak. Experiments have been performed in KT-1 small size tokamak. Electric transport probe is composed of a single probe(SP) and a Mach probe (MP). SP is used for the measurements of electron density, floating potential, and plasma density and measured values are {approx} 3*10{sup 11}/cm{sup -3}, -20 volts, 15 {approx} 25 eV. For the most discharges, respectively. MP is for the measurements of toroidal(M{sub T}) and poloidal(M{sub P}) flow velocities, and density, which are M{sub T} {approx_equal} .0.85, M{sub P} {approx_equal}. 0.17, n. {approx_equal} 2.1*10{sup 11} cm{sup -3}, respectively. A triple probe is also developed for the direct reading of T{sub e} and n{sub e}, and is used for DC, RF, and RF+DC plasma in APL of Hanyang university. 38 refs., 36 figs. (author)

  17. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Langer, W.D.

    1982-08-01

    We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

  18. Plasmator. A numerical code for simulation of plasma transport in Tokamaks

    International Nuclear Information System (INIS)

    Guasp, J.

    1979-01-01

    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)

  19. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  20. Large plasma pressure perturbations and radial convective transport in a tokamak

    International Nuclear Information System (INIS)

    Krasheninnikov, Sergei; Yu, Guanghui; Ryutov, Dmitri

    2004-01-01

    Strongly localized plasma structures with large pressure inhomogeneities (such as plasma blobs in the scrape-off-layer (SOL)/shadow regions, pellet clouds, Edge localized Modes (ELMs)) observed in the tokamaks, stellarators and linear plasma devices. Experimental studies of these phenomena reveal striking similarities including more convective rather than diffusive radial plasma transport. We suggest that rather simple models can describe many essentials of blobs, ELMs, and pellet clouds dynamics. The main ingredient of these models is the effective plasma gravity caused by magnetic curvature, centrifugal or friction forces effects. As a result, the equations governing plasma transport in such localized structures appear to be rather similar to that used to describe nonlinear evolution of thermal convection in the Boussinesq approximation (directly related to the Rayleigh-Taylor (RT) instability). (author)

  1. Experimental investigation of turbulent transport at the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Fedorczak, N.

    2010-01-01

    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)

  2. A one-dimensional transport code for the simulation of D-T burning tokamak plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

    1980-11-01

    A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

  3. Transient heat transport studies in JET conventional and advanced tokamak plasmas

    International Nuclear Information System (INIS)

    Mantica, P.; Coffey, I.; Dux, R.

    2003-01-01

    Transient transport studies are a valuable complement to steady-state analysis for the understanding of transport mechanisms and the validation of physics-based transport models. This paper presents results from transient heat transport experiments in JET and their modelling. Edge cold pulses and modulation of ICRH (in mode conversion scheme) have been used to provide detectable electron and ion temperature perturbations. The experiments have been performed in conventional L-mode plasmas or in Advanced Tokamak regimes, in the presence of an Internal Transport Barrier (ITB). In conventional plasmas, the issues of stiffness and non-locality have been addressed. Cold pulse propagation in ITB plasmas has provided useful insight into the physics of ITB formation. The use of edge perturbations for ITB triggering has been explored. Modelling of the experimental results has been performed using both empirical models and physics-based models. Results of cold pulse experiments in ITBs have also been compared with turbulence simulations. (author)

  4. TEMPEST simulations of the plasma transport in a single-null tokamak geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Rognlien, T.D.; Bodi, K.; Krasheninnikov, S.

    2010-01-01

    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.

  5. MTX [Microwave Tokamak Experiment] diagnostic and auxiliary systems for confinement, transport, and plasma physics studies

    International Nuclear Information System (INIS)

    Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.

    1989-01-01

    This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively

  6. Fokker--Planck/transport analyses of fusion plasmas in contemporary beam-driven tokamaks

    International Nuclear Information System (INIS)

    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 approximately 0.5, [ 2 / 3 ] = 22 keV approximately 6 , and the D-D neutron intensity is 10 16 n/sec

  7. Transport of carbon ion test particles and hydrogen recycling in the plasma of the Columbia tokamak ''HBT'' [High Beta Tokamak

    International Nuclear Information System (INIS)

    Wang, Jian-Hua.

    1990-01-01

    Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n e ∼ 1 - 5 x 10 14 (cm -3 ), T e ∼ 4 - 10 (eV), B t ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of C II (4267A, emitted by the C + ions) and C III (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 H α 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 τ p is comparable with the plasma energy confinement time τ E ; 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

  8. Plasma Turbulence Suppression and Transport Barrier Formation by Externally Driven RF Waves in Spherical Tokamaks

    International Nuclear Information System (INIS)

    Bruma, C.; Cuperman, S.C.; Komoshvili, K.

    2002-01-01

    Turbulent transport of heat and particles is the principle obstacle confronting controlled fusion today. Thus, we investigate quantitatively the suppression of turbulence and formation of transport barriers in spherical tokamaks by sheared electric fields generated by externally driven radio-frequency (RF) waves, in the frequency range o)A n o] < o)ci (e)A and o)ci are the Alfven and ion cyclotron frequencies). This investigation consists of the solution of the full-wave equation for a spherical tokamak in the presence of externally driven fast waves and the evaluation of the power dissipation by the mode-converted Alfven waves. This in turn, provides a radial flow shear responsible for the suppression of plasma turbulence. Thus, a strongly non-linear equation for the radial sheared electric field is solved, the turbulent transport suppression rate is evaluated and compared with the ion temperature gradient (ITG) instability increment. For illustration, the case of START-like device (Sykes 2000) is treated. Thus, (i) the exact D-shape cross-section is considered; (ii) additional kinetic (including Landau damping) and particle trapping effects are added to the resistive two-fluid dielectric tensor operator; (iii) a finite extension antenna located on the low-field-side of the plasma is considered; (iv) a rigorous 2.5 finite elements numerical code (Sewell 1993) is used; and (v) the turbulence and transport barrier generated as a result of wave-plasma interaction is evaluated

  9. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    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

  10. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  11. Possible effects of drift wave turbulence on magnetic structure and plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.

    1977-07-01

    A new mechanism is proposed by which low level, drift wave type fluctuations, such as those observed in the ATC and TFR experiments, can cause anomalous radial electron heat transport in tokamaks. The model is based on the fact that since transport processes parallel to the magnetic field are many orders of magnitude more rapid than perpendicular ones, very small helically resonant magnetic perturbations that cause field lines to move radially allow the parallel transport process to contribute to radial electron heat transport. It is hypothesized that the small magnetic perturbations accompanying drift waves at any nonzero plasma β are large enough to produce significant effects in present tokamak experiments. The helical magnetic component of drift waves produces magnetic island structures whose spatial widths can easily exceed the ion gyroradius. In a drift wave oscillation period, electrons circumnavigate a magnetic island, whereas the slower moving ions see only a tilt of the magnetic field lines. Thus, electrons try to diffuse radially more rapidly than ions; however, a radialpotential builds up on a very short time scale to confine the electrons electrostatically and thereby keep the particle diffusion ambipolar. Nonetheless, this parallel electron diffusion process does cause net radial electron heat conduction through an ensemble of closely packed island structures. The heat conduction coefficient is estimated. Other effects that these magnetic flutters may have on plasma transport and runaway electron processes are also discussed

  12. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  13. A study on tokamak fusion reactor - Numerical analyses of MHD equilibrium= and edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)

  14. The role of drifts in the plasma transport at the tokamak core–SOL interface

    Energy Technology Data Exchange (ETDEWEB)

    Chankin, A.V., E-mail: Alex.Chankin@ipp.mpg.de [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany); Coster, D.P. [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2013-07-15

    The interface between the core (inside the magnetic separatrix in X-point configurations) and the scrape-off layer (SOL) of tokamaks is a delicate region of the magnetic topology transition from closed to open field lines where neither the standard neoclassical theory nor the SOL physics fully apply. Sharp gradients of plasma parameters in the outer core, caused by the proximity of divertor sinks in the near SOL, invalidate some ordering assumptions of the neoclassical theory. At the same time, the existence of closed flux surfaces in the core enforces ambipolarity of radial plasma flows, in difference to the situation in the SOL where the current loop may close through the divertor. Detailed analysis of the plasma transport and flows with the emphasis on the outer core region, just inside the separatrix, is carried out in the paper, based on EDGE2D modelling and analytical formulas.

  15. Extended numerical modeling of impurity neoclassical transport in tokamak edge plasmas

    International Nuclear Information System (INIS)

    Inoue, H.; Yamoto, S.; Hatayama, A.; Homma, Y.

    2016-01-01

    Understanding of impurity transport in tokamaks is an important issue in order to reduce the impurity contamination in fusion core plasmas. Recently, a new kinetic numerical scheme of impurity classical/neoclassical transport has been developed. This numerical scheme makes it possible to include classical self-diffusion (CL SD), classical inward pinch (CL IWP), and classical temperature screening effect (CL TSE) of impurity ions. However, impurity neoclassical transport has been modeled only in the case where background plasmas are in the Pfirsch-Schluter (PS) regime. The purpose of this study is to extend our previous model to wider range of collisionality regimes, i.e., not only the PS regime, but also the plateau regime. As in the previous study, a kinetic model with Binary Collision Monte-Carlo Model (BMC) has been adopted. We focus on the modeling of the neoclassical self-diffusion (NC SD) and the neoclassical inward pinch (NC IWP). In order to simulate the neoclassical transport with the BCM, velocity distribution of background plasma ions has been modeled as a deformed Maxwell distribution which includes plasma density gradient. Some test simulations have been done. As for NC SD of impurity ions, our scheme reproduces the dependence on the collisionality parameter in wide range of collisionality regime. As for NC IWP, in cases where test impurity ions and background ions are in the PS and plateau regimes, parameter dependences have been reproduced. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Extended numerical modeling of impurity neoclassical transport in tokamak edge plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, H.; Yamoto, S.; Hatayama, A. [Graduate School of Science and Technology, Keio University, Hiyoshi, Yokohama (Japan); Homma, Y. [Graduate School of Science and Technology, Keio University, Hiyoshi, Yokohama (Japan); Research Fellow of Japan Society for the Promotion of Science, Tokyo (Japan)

    2016-08-15

    Understanding of impurity transport in tokamaks is an important issue in order to reduce the impurity contamination in fusion core plasmas. Recently, a new kinetic numerical scheme of impurity classical/neoclassical transport has been developed. This numerical scheme makes it possible to include classical self-diffusion (CL SD), classical inward pinch (CL IWP), and classical temperature screening effect (CL TSE) of impurity ions. However, impurity neoclassical transport has been modeled only in the case where background plasmas are in the Pfirsch-Schluter (PS) regime. The purpose of this study is to extend our previous model to wider range of collisionality regimes, i.e., not only the PS regime, but also the plateau regime. As in the previous study, a kinetic model with Binary Collision Monte-Carlo Model (BMC) has been adopted. We focus on the modeling of the neoclassical self-diffusion (NC SD) and the neoclassical inward pinch (NC IWP). In order to simulate the neoclassical transport with the BCM, velocity distribution of background plasma ions has been modeled as a deformed Maxwell distribution which includes plasma density gradient. Some test simulations have been done. As for NC SD of impurity ions, our scheme reproduces the dependence on the collisionality parameter in wide range of collisionality regime. As for NC IWP, in cases where test impurity ions and background ions are in the PS and plateau regimes, parameter dependences have been reproduced. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  17. Modelling the turbulent transport of angular momentum in tokamak plasmas - A quasi-linear gyrokinetic approach

    International Nuclear Information System (INIS)

    Cottier, Pierre

    2013-01-01

    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) [fr

  18. Anomalous transport in tokamaks

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1989-01-01

    A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed

  19. Plasma features and alpha particle transport in low-aspect ratio tokamak reactor

    International Nuclear Information System (INIS)

    Xu Qiang; Wang Shaojie

    1997-06-01

    The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)

  20. Energy confinement and transport of H-mode plasmas in tokamak

    International Nuclear Information System (INIS)

    Urano, Hajime

    2005-02-01

    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

  1. Characterization of intermittency of impurity turbulent transport in tokamak edge plasmas

    International Nuclear Information System (INIS)

    Futatani, S.; Benkadda, S.; Nakamura, Y.; Kondo, K.

    2008-01-01

    The statistical properties of impurity transport of a tokamak edge plasma embedded in a dissipative drift-wave turbulence are investigated using structure function analysis. The impurities are considered as a passive scalar advected by the plasma flow. Two cases of impurity advection are studied and compared: A decaying impurities case (given by a diffusion-advection equation) and a driven case (forced by a mean scalar gradient). The use of extended self-similarity enables us to show that the relative scaling exponent of structure functions of impurity density and vorticity exhibit similar multifractal scaling in the decaying case and follows the She-Leveque model. However, this property is invalidated for the impurity driven advection case. For both cases, potential fluctuations are self-similar and exhibit a monofractal scaling in agreement with Kolmogorov-Kraichnan theory for two-dimensional turbulence. These results obtained with a passive scalar model agree also with test-particle simulations.

  2. Poloidal density variation of impurities in a rotating tokamak plasma - flux surface coordinates and effect on transport coefficients

    International Nuclear Information System (INIS)

    Romanelli, M.

    1999-09-01

    The poloidal variation of impurity densities over magnetic surfaces brings about an enhancement of neoclassical transport coefficients, as shown by Romanelli and Ottaviani for impurities in the Pfirsch Schlueter regime and by Helander for particles in the banana-plateau regime, both in a large aspect ratio tokamak. The same effect will occur in a finite aspect ratio tokamak and therefore it is considered to be relevant for inclusion in transport codes for comparison with the experimental measurements of impurity transport. Here an expression for the impurity-density poloidal-variation generated by the fast toroidal rotation of the plasma column is presented in general coordinates. (author)

  3. Impurity transport and plasma rotation in the ISX-B tokamak

    International Nuclear Information System (INIS)

    Isler, R.C.; Murray, L.E.; Crume, E.C.

    1983-01-01

    Recent calculations have shown that when external momentum sources and plasma rotation are included in the neoclassical theory, the standard results for impurity transport can be strongly altered. Under appropriate conditions, inward convection is reduced by co-injection and enhanced by counter-injection. In order to examine the theoretical predictions, several observations of impurity transport have been made in the ISX-B tokamak during neutral-beam injection for comparison with the transport seen with Ohmic heating alone. Both intrinsic contaminants and deliberately introduced test impurities display a behaviour that is in qualitative agreement with the predicted beam-driven effects. These correlations are particularly noticeable when the comparisons are made for deuterium when the impurity transport in the Ohmically heated discharges exhibits neoclassical-like characteristics, i.e. accumulation and long confinement times. Similar but smaller effects are observed in beam-heated hydrogen discharges; neoclassical-like behaviour is not seen in Ohmically heated hydrogen sequences. Emphasis has been placed on measuring toroidal plasma rotation, and semiquantitative comparisons with the theories of beam-induced impurity transport have been made. It is possible that radial electric fields other than those associated with momentum transfer and increased anomalous processes during injection could also play a role. (author)

  4. 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.

  5. Numerical simulation of edge plasma in tokamak

    International Nuclear Information System (INIS)

    Chen Yiping; Qiu Lijian

    1996-02-01

    The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

  6. Multiscale gyrokinetics for rotating tokamak plasmas: fluctuations, transport and energy flows.

    Science.gov (United States)

    Abel, I G; Plunk, G G; Wang, E; Barnes, M; Cowley, S C; Dorland, W; Schekochihin, A A

    2013-11-01

    This paper presents a complete theoretical framework for studying turbulence and transport in rapidly rotating tokamak plasmas. The fundamental scale separations present in plasma turbulence are codified as an asymptotic expansion in the ratio ε = ρi/α of the gyroradius to the equilibrium scale length. Proceeding order by order in this expansion, a set of coupled multiscale equations is developed. They describe an instantaneous equilibrium, the fluctuations driven by gradients in the equilibrium quantities, and the transport-timescale evolution of mean profiles of these quantities driven by the interplay between the equilibrium and the fluctuations. The equilibrium distribution functions are local Maxwellians with each flux surface rotating toroidally as a rigid body. The magnetic equilibrium is obtained from the generalized Grad-Shafranov equation for a rotating plasma, determining the magnetic flux function from the mean pressure and velocity profiles of the plasma. The slow (resistive-timescale) evolution of the magnetic field is given by an evolution equation for the safety factor q. Large-scale deviations of the distribution function from a Maxwellian are given by neoclassical theory. The fluctuations are determined by the 'high-flow' gyrokinetic equation, from which we derive the governing principle for gyrokinetic turbulence in tokamaks: the conservation and local (in space) cascade of the free energy of the fluctuations (i.e. there is no turbulence spreading). Transport equations for the evolution of the mean density, temperature and flow velocity profiles are derived. These transport equations show how the neoclassical and fluctuating corrections to the equilibrium Maxwellian act back upon the mean profiles through fluxes and heating. The energy and entropy conservation laws for the mean profiles are derived from the transport equations. Total energy, thermal, kinetic and magnetic, is conserved and there is no net turbulent heating. Entropy is produced

  7. Radio-frequency heating and neutral atom transport in a fluid-magnetohydrodynamic treatment of burning tokamak plasmas

    International Nuclear Information System (INIS)

    Conn, R.W.; Mau, T.K.; Prinja, A.K.

    1983-01-01

    A physical model for the space and time evolution of the primary parameters of ordinary and burning tokamak plasmas is described by employing a fluid plasma treatment coupled to a magnetohydrodynamic equilibrium description, the solution to the appropriate Maxwell equations, and the solution of the linear transport equation describing neutral atom transport in plasmas. The specific problems of plasma heating by ion cyclotron radiofrequency (ICRF) waves and neutral atom transport in the plasma edge and in complicated geometrical components such as divertor channels or pumped limiter structures are analyzed. A theoretical, onedimensional slab model of ICRF heating at ω = 2ω/SUB cD/ is developed and applied to determine the space-time response of tokamak plasmas. Generally, strong single-pass absorption is found for high-density, high (β) plasmas using a low k 11 spectrum (0.05 to 0.1 cm -1 ) although for (β > 1%, electron Landau damping becomes important. Deterministic and Monte Carlo methods to solve the neutral atom transport problem are described. Specific application to determine the spectrum of neutral atoms emerging from the duct of a pump limiter shows it to be hard (mean energy > 20 eV), indicating very incomplete energy thermalization. Uncertainties are identified in the overall problem of dynamic burning plasma analysis caused by the complexity of the problem itself and by uncertainties in fundamental areas such as plasma transport coefficients, stability, and plasma edge physics

  8. Investigation of the energy transport mechanism in the TCA tokamak by studying the plasma dynamical response

    International Nuclear Information System (INIS)

    Dudok de Wit, Th.; Duval, B.P.; Joye, B.; Lister, J.B.; Moret, J.M.

    1989-01-01

    The energy transport mechanisms that govern the electron temperature behaviour of a tokamak remain very badly understood and up to now no proper model has been proposed that can explain experimental observations such as profile consistency or the influence of the density profile. One approach to this problem, extensively used on TCA, is to study the dynamical response of the plasma due to externally imposed modifications of parameters which have an influence on the plasma energy content. The temporal evolution of the electron temperature will closely depend on the type and the characteristics of the implied mechanisms. Thus a detailed measurement of the dynamical response would reveal experimentally the dominant properties that would have to be taken into account in the elaboration of a model of the transport processes. Most of the results presented here were obtained by analysing the electron temperature response inferred from soft X-ray emissivity during modification of the plasma density due to either gas puffing, laser impurity ablation or alfven wave heating on TCA (a = 0.18 m, R = 0.61 m, B Φ = 1.52 T). 4 refs., 3 figs

  9. Plasma transport in stochastic magnetic field caused by vacuum resonant magnetic perturbations at diverted tokamak edge

    International Nuclear Information System (INIS)

    Park, G.; Chang, C. S.; Joseph, I.; Moyer, R. A.

    2010-01-01

    A kinetic transport simulation for the first 4 ms of the vacuum resonant magnetic perturbations (RMPs) application has been performed for the first time in realistic diverted DIII-D tokamak geometry [J. Luxon, Nucl. Fusion 42, 614 (2002)], with the self-consistent evaluation of the radial electric field and the plasma rotation. It is found that, due to the kinetic effects, the stochastic parallel thermal transport is significantly reduced when compared to the standard analytic model [A. B. Rechester and M. N. Rosenbluth, Phys. Rev. Lett. 40, 38 (1978)] and the nonaxisymmetric perpendicular radial particle transport is significantly enhanced from the axisymmetric level. These trends agree with recent experimental result trends [T. E. Evans, R. A. Moyer, K. H. Burrell et al., Nat. Phys. 2, 419 (2006)]. It is also found, as a side product, that an artificial local reduction of the vacuum RMP fields in the vicinity of the magnetic separatrix can bring the kinetic simulation results to a more detailed agreement with experimental plasma profiles.

  10. Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Stoschus, Henning

    2011-01-01

    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 n e and temperature T e 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 ν 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 along open magnetic field lines to the wall components. For high

  11. Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

    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

  12. Transport and confinement in the Mega Ampere Spherical Tokamak (MAST) plasma

    International Nuclear Information System (INIS)

    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; 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; Walsh, M J

    2003-01-01

    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 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 LH approx. R 2 ). In addition, MAST favours an inverse aspect ratio scaling P 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 ped approx. epsilon -2.13 and modifies the exponents on R, B 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 techniques developed at conventional aspect

  13. 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

  14. Effect of nonlinear energy transport on neoclassical tearing mode stability in tokamak plasmas

    Science.gov (United States)

    Fitzpatrick, Richard

    2017-05-01

    An investigation is made into the effect of the reduction in anomalous perpendicular electron heat transport inside the separatrix of a magnetic island chain associated with a neoclassical tearing mode in a tokamak plasma, due to the flattening of the electron temperature profile in this region, on the overall stability of the mode. The onset of the neoclassical tearing mode is governed by the ratio of the divergences of the parallel and perpendicular electron heat fluxes in the vicinity of the island chain. By increasing the degree of transport reduction, the onset of the mode, as the divergence ratio is gradually increased, can be made more and more abrupt. Eventually, when the degree of transport reduction passes a certain critical value, the onset of the neoclassical tearing mode becomes discontinuous. In other words, when some critical value of the divergence ratio is reached, there is a sudden bifurcation to a branch of neoclassical tearing mode solutions. Moreover, once this bifurcation has been triggered, the divergence ratio must be reduced by a substantial factor to trigger the inverse bifurcation.

  15. Computational images of internal-transport-barrier oscillations in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Bizarro, J.P S. [Inst Super Tecn, Ctr Fusao Nucl, EURATOM Assoc, P-1049001 Lisbon (Portugal); Litaudon, X.L. [CEA Cadarache, Dept Rech Fus Controlee, EURATOM Assoc, F-13108 St Paul Les Durance (France); Tala, T.J.J. [Assoc Euratom Tekes, FIN-02044 Espoo (Finland); JET EFDA Contributors [Culham Sci Ctr, Abingdon OX14 3DB, Oxon (United Kingdom)

    2008-07-01

    A well-known benchmarked code, where a Bohm-gyro-Bohm transport model is complemented with an empirical scaling for the dynamics of internal transport barriers (ITBs), is used to model the ITB oscillations that are often seen in advanced tokamak scenarios with a dominant fraction of bootstrap current. (authors)

  16. Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

    International Nuclear Information System (INIS)

    Peterson, R.R.; MacFarlane, J.J.; Wang, P.

    1994-01-01

    Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

  17. Turbulent transport reduction by E x B velocity shear during edge plasma biasing in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, G. [Dept. of Applied Physics, Ghent Univ., Ghent (Belgium); Adamek, J.; Antoni, V.; Balan, P.; Boedo, J.A.; Devynck, P.; Duran, I.; Eliseev, L.; Gunn, J.P.; Hron, M.; Ionita, C.; Jachmich, S.; Kirnev, G.S.; Martines, E.; Melnikov, A.; Peleman, P.; Schrittwieser, R.; Silva, C.; Stoeckel, J.; Tendler, M.; Varandas, C.; Van Schoor, M.; Vershkov, V.; Weynants, R.R.

    2004-07-01

    Experiments in the tokamaks TEXTOR, CASTOR, T-10 and ISTTOK have provided new and complementary evidence on the physics of the universal mechanism of E x B velocity shear stabilization of turbulence, concomitant transport barrier formation and radial conductivity by using various edge biasing techniques. (orig.)

  18. Understanding the SOL flow in L-mode plasma on divertor tokamaks, and its influence on the plasma transport

    International Nuclear Information System (INIS)

    Asakura, Nobuyuki

    2007-01-01

    Significant progress has been made in understanding the driving mechanisms in SOL mass transport along the magnetic field lines (SOL flow). SOL flow measurements by Mach probes and impurity plume have been performed in L-mode plasma at various poloidal locations in divertor tokamaks. All results showed common SOL flow patterns: subsonic flow with parallel Mach number (M parallel ) of 0.2-1 was generated from the Low-Field-Side (LFS) SOL to the High-Field-Side (HFS) divertor for the ion ∇B drift towards the divertor. The SOL flow pattern was formed mainly by LFS-enhanced asymmetry in diffusion and by classical drifts. In addition, divertor detachment and/or intense puffing-and-pump enhanced the HFS SOL flow. Most codes have incorporated drift effects, and asymmetric diffusion was modelled to simulate the fast SOL flow. Influences of the fast SOL flow on the impurity flow in the SOL, shielding from core plasma, and deposition profile, were directly observed in experiments

  19. Classical tokamak transport theory

    International Nuclear Information System (INIS)

    Nocentini, Aldo

    1982-01-01

    A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)

  20. Global gyrokinetic simulation of tokamak transport

    International Nuclear Information System (INIS)

    Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

    1998-10-01

    A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

  1. Turbulent transport and shear at the E x B velocity in wall plasma of the TF-2 tokamak

    International Nuclear Information System (INIS)

    Budaev, V.P.

    1999-01-01

    Turbulence of near-the-wall plasma and potentialities of affecting the turbulence and periphery transport of the TF-2 tokamak by inducing radial electric fields and ergodization of periphery magnetic structure have been investigated, the results are presented. Essential role of the E x B velocity shear in suppression of the turbulence and turbulent transport in periphery has been pointed out. Decrease in transport losses stemming from effect of radial electric fields is brought about suppression of turbulence amplitude, decrease in correlations and decrease in the width of the wave numbers spectrum. Profiles of plasma density, electron temperature, turbulence level, electric fields over entire periphery of discharge change as a result. Ergodization of magnetic structure also results in the change of properties of periphery turbulence and turbulent transport [ru

  2. Application of a two fluid theoretical plasma transport model on current tokamak reactor designs

    International Nuclear Information System (INIS)

    Ibrahim, E.; Fowler, T.K.

    1987-06-01

    In this work, the new theoretical transport models to TIBER II design calculations are described and the results are compared with recent experimental data in large tokamaks (TFTR, JET). Tang's method is extended to a two-fluid model treating ions and electrons separately. This allows for different ion and electron temperatures, as in recent low-density experiments in TFTR, and in the TIBER II design itself. The discussion is divided into two parts: (1) Development of the theoretical transport model and (2) calibration against experiments and application to TIBER II

  3. The possible role of Reynolds stress in the creation of a transport barrier in tokamak edge plasmas

    International Nuclear Information System (INIS)

    Vergote, M.; Van Schoor, M.; Xu, Y.; Jachmich, S.; Weynants, R.; Hron, M.; Stoeckel, J.

    2005-01-01

    To obtain a good confinement, mandatory in a fusion reactor, the understanding of the formation of transport barriers in the edge plasma of a tokamak is essential. Turbulence, the major candidate to explain anomalous transport, can be quenched by sheared flows in the edge which rip the convective cells apart, thus forming a barrier. Experimental evidence from the Chinese HT-6M tokamak [Y.H. Xu et al.: Phys. Rev. Lett. 84 (2000) 3867], points to the fact that momentum transfer from the turbulence can create these sheared flows via the Reynolds stresses. A new 1-d fluid model for the generation of the poloidal flow, has been developed taking into account the driving force of the Reynolds stress and the friction forces due to neutrals and parallel viscosity. Special attention has been dedicated to the computation of the flux-surface-averaging for the various terms. This model has been confronted with the experimental results obtained in the HT-6M tokamak, where Reynolds stresses were generated by application of a turbulent heating pulse. If the model is applied in cylindrical geometry, the calculated Reynolds stress-induced flow agrees well with the measured poloidal velocity in the plasma edge. However, when the full toroidal geometry is taken into account, it seems that the Reynolds stresses are too small to explain the observed rotation. This indicates that the role of the Reynolds stresses in inducing macroscopic flow in the torus is weakened. A combined system of probes allowing to measure the Reynolds stress and the rotation velocity simultaneously, has been developed and installed on the CASTOR tokamak. We report here on the first results obtained. (author)

  4. Three-dimensional plasma transport in open chaotic magnetic fields. A computational assessment for tokamak edge layers

    International Nuclear Information System (INIS)

    Frerichs, Heinke Gerd

    2010-04-01

    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

  5. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    Pitcher, C.S.

    1987-11-01

    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 D 2 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

  6. Developments in tokamak transport modeling

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Attenberger; Lao, L.L.

    1981-01-01

    A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements

  7. Effects of Collisionality on the Nonlinear Characteristics of Boundary Turbulence and Blob/hole Transport in Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Yasuaki, K., E-mail: lijq@energy.kyoto-u.ac.jp [Kyoto University, Kyoto (Japan); Cheng, J.; Longwen, Y.; Jiaqi, D. [Southwestern Institute of Physics, Chengdu (China)

    2012-09-15

    Full text: Blob/hole dynamics near tokamak separatrix is of striking importance in determining the boundary transport. Based on simulations using an extended 2-region (edge/SOL) fluid model, we found that blob/hole dynamics are sensitively influenced by the plasma collisionality, i.e., ion-electron and ion-neutral collisions. Namely, the holes are enhanced in highly collisional edge whereas the blobs are weakened at the SOL, causing larger particle convection. These blob/hole dynamics are closely correlated with potential dipoles. The trends are experimentally evidenced on the HL-2A tokamak. Moreover, as the neutral-ion collision increases, the blobs at the SOL tend to develop into streamers propagating outwards with reduced amplitude while the holes inwards are suppressed, showing a key role in nonlinear structure regulation and resultant transport suppression. Results suggest that adjusting the plasma collisionality by fueling, e.g., gas puffing, could serve as a method to nonlinearly select turbulent structures, i.e., blobs, holes or streamers, to access the control of boundary transport. (author)

  8. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  9. Turbulence induced radial transport of toroidal momentum in boundary plasma of EAST tokamak

    International Nuclear Information System (INIS)

    Zhao, N.; Yan, N.; Xu, G. S.; Wang, H. Q.; Wang, L.; Ding, S. Y.; Chen, R.; Chen, L.; Zhang, W.; Hu, G. H.; Shao, L. M.; Wang, Z. X.

    2016-01-01

    Turbulence induced toroidal momentum transport in boundary plasma is investigated in H-mode discharge using Langmuir-Mach probes on EAST. The Reynolds stress is found to drive an inward toroidal momentum transport, while the outflow of particles convects the toroidal momentum outwards in the edge plasma. The Reynolds stress driven momentum transport dominates over the passive momentum transport carried by particle flux, which potentially provides a momentum source for the edge plasma. The outflow of particles delivers a momentum flux into the scrape-off layer (SOL) region, contributing as a momentum source for the SOL flows. At the L-H transitions, the outward momentum transport suddenly decreases due to the suppression of edge turbulence and associated particle transport. The SOL flows start to decelerate as plasma entering into H-mode. The contributions from turbulent Reynolds stress and particle transport for the toroidal momentum transport are identified. These results shed lights on the understanding of edge plasma accelerating at L-H transitions.

  10. Fast transient transport phenomena measured by soft X-ray emission in TCV tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Furno, I. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-08-01

    Energy and particle transport during sawtooth activity in TCV plasmas has been studied in this thesis with high temporal resolution many chord diagnostics. We indicated the influence of sawteeth on plasma profiles in ohmic conditions and in the presence of auxiliary electron cyclotron resonance heating and current drive. A 2-dimensional model for heat transport, including localised heat source and a magnetic island, has been used to interpret the experimental observations. These results provided a new interpretation of a coupled heat and transport phenomenon which is potentially important for plasma confinement. The observations validate the applicability and show the possibility of improvement of a 2-dimensional theoretic a1 model for the study of heat transport in the presence of localised heat source and a magnetic island. Furthermore, the TCV results showed a new possibility for the interpretation of a coupled heat and particle transport phenomenon previously understood only in stellarators. (author)

  11. Cross-field blob transport in tokamak scrape-off-layer plasmas

    International Nuclear Information System (INIS)

    D'Ippolito, D.A.; Myra, J.R.; Krasheninnikov, S.I.

    2002-01-01

    Recent measurements show that nondiffusive, intermittent transport of particles can play a major role in the scrape-off-layer (SOL) of fusion experiments. A possible mechanism for fast convective plasma transport is related to the plasma filaments or 'blobs' observed in the SOL with fast cameras and probes. In this paper, physical arguments suggesting the importance of blob transport [S. I. Krasheninnikov, Phys. Lett. A 283, 368 (2001)] have been extended by calculations using a three-field fluid model, treating the blobs as coherent propagating structures. The properties of density, temperature and vorticity blobs, and methods of averaging over ensembles of blobs to get the average SOL profiles, are illustrated. The role of ionization of background neutrals in sustaining the density blob transport is also discussed. Many qualitative features of the experiments, such as relatively flat density profiles and transport coefficients increasing toward the wall, are shown to emerge naturally from the blob transport paradigm

  12. Comparative studies of stellarator and tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

    Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team

    1997-09-01

    Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.

  13. SAFETY FACTOR SCALING OF ENERGY TRANSPORT IN L-MODE PLASMAS ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PETTY, C.C.; KINSEY, J.E.; LUCE, T.C.

    2003-01-01

    OAK-B135 The scaling of energy transport with safety factor (q) at fixed magnetic shear has been measured on the DIII-D tokamak [Nucl. Fusion 42, 614 (2002)] for low confinement (L) mode discharges. At constant density, temperature, and toroidal magnetic field strength, such that the toroidal dimensionless parameters other than q are held fixed, the one-fluid thermal diffusivity is found to scale like χ ∝ q 0.84±0.15 , with the ion channel having a stronger q dependence than the electron channel in the outer half of the plasma. The measured q scaling is in good agreement with the predicted scaling by the GLF23 transport model for the ion temperature gradient and trapped electron modes, but it is significantly weaker than the inferred scaling from empirically-derived confinement scaling relations

  14. Multigroup calculations of low-energy neutral transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Gralnick, S.L.; Price, W.G. Jr.; Kammash, T.

    1978-01-01

    Multigroup discrete ordinates methods avoid many of the approximations that have been used in previous neutral transport analyses. Of particular interest are the neutral profiles generated as an integral part of larger plasma system simulation codes. To determine the appropriateness of utilizing a particular multigroup code, ANISN, for this purpose, results are compared with the neutral transport module of the Duechs code. For a typical TFTR plasma, predicted neutral densities differ by a maximum factor of three on axis and outfluxes at the plasma boundary by approximately 40%. This is found to be significant for a neutral transport module. Possible sources of the observed discrepancies are indicated from an analysis of the approximations used in the Duechs model. Recommendations are made concerning the future application of the multigroup method. (author)

  15. A study on the fusion reactor - Numerical analyses of MHD equilibrium and= edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Deok Kyu; Chung, TaeKyun; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-D MHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we computed the MHD equilibria of single and double null configurations and determined the external coil currents and the plasma parameters used for operation and control data. Also we numerically acquired the distributions of edge plasma parameters in poloidal and radial directions= and the design-related values according to the various operating conditions using the developed plasma transport code. Furthermore, a neutral particle transport code for the edge region is developed and them used for the analysis of the neutral particle behavior yielding the source terms in the fluid transport equations, and expected to supply the input parameters for the edge plasma transport code. 53 refs., 12 tabs., 44 figs. (author)

  16. Local transport analysis of L-mode plasmas in JT-60 tokamak

    International Nuclear Information System (INIS)

    Hirayama, Toshio; Kikuchi, Mitsuru; Shirai, Hiroshi; Shimizu, Katsuhiro; Yagi, Masatoshi; Koide, Yoshihiko; Ishida, Shinichi; Azumi, Masafumi.

    1991-03-01

    Local heat transport has been studied in auxiliary heated JT-60 plasmas with emphasis on understanding the deteriorated confinement observed in L-mode plasmas. The systematic experiment and analysis have been carried out in L-mode phase of divertor (single null, lower X-point), and limiter discharges with hydrogen neutral beam heating into hydrogen plasmas, based on sets of consistent experimental data including ion temperature profiles from CXR measurements. The deterioration in the energy confinement time with increasing the auxiliary heating power, so-called the power scaling, is mainly due to the degradation in ion energy transport. The confinement improvement as the plasma current increases is followed by both improvement in ion and electron transport properties. It is found that the ion thermal diffusivity has an approval dependence on the density. High ion temperature (T i (0) ≤ 12 keV) L-mode plasmas are attained at high β p up to 3.5. The centrally peaked ion temperature is significantly due to the improvement in ion transport property, which is reduced to the level of the electron thermal diffusivities. (author)

  17. Simulations of phenomena related to edge transport in tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Konzett, S.

    2011-01-01

    This thesis investigates turbulence in a tokamak fusion plasma using numerical simulations. The fluid turbulence code ATTEMPT, which computes the drift dynamics of ions and electrons in electromagnetic fields, is applied to investigate three physical effects which are motivated by recent experimental findings. The first part shows that the statistics of drift fluid turbulence are largely unaffected by the presence of rational magnetic surfaces for typical edge parameter regimes. The second part contains an analysis of the dependence of correlation lengths on various physical parameters. A systematic approach reveals the impact of plasma parameters - which change in the transition from L to H-mode - on parallel, radial and perpendicular correlation lengths. In the last part of the thesis a new flux surface geometry is implemented in the ATTEMPT code. The modified geometry models the onset of the change in magnetic topology near a magnetic X-point. Computations show that turbulent fluctuations are reduced in an X-point distorted flux surface geometry, and the spectral structure of turbulence is altered substantially. (author) [de

  18. Comparison of transient electron heat transport in LHD helical and JT-60U tokamak plasmas

    International Nuclear Information System (INIS)

    Inagaki, S.; Ida, K.; Tamura, N.; Shimozuma, T.; Kubo, S.; Nagayama, Y.; Kawahata, K.; Sudo, S.; Ohkubo, K.; Takenaga, H.; Isayama, A.; Takizuka, T.; Kamada, Y.; Miura, Y.

    2005-01-01

    Transient transport experiments are performed in plasmas with and without Internal Transport Barrier (ITB) on LHD and JT-60U. The dependence of χ e on electron temperature, T e , and electron temperature gradient, ∇T e , is analyzed by an empirical non-linear heat transport model. In plasmas without ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation. The χ e depends on T e and ∇T e in JT-60U, while the ∇T e dependence is weak in LHD. Inside the ITB region, there is no or weak ∇T e dependence both in LHD and JT-60U. A cold pulse growing driven by the negative T e dependence of χ e is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U). (author)

  19. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  20. Design of multi-range tomographic system for transport studies in tokamak plasmas

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Naydenkova, Diana; Šesták, David; Vlček, Jiří; Mlynář, Jan; Melich, Radek; Jareš, Daniel; Malot, J.; Sarychev, D.; Igochine, V.

    2010-01-01

    Roč. 623, č. 2 (2010), s. 806-808 ISSN 0168-9002. [International Conference on Frontiers in Diagnostic Technologies/1st./. Frascati, 25.11.2009-27.11.2009] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : SXR * Bolometry * Visible light * Tokamak * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.142, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6TJM-4YTFBF2-C&_user=6542793&_coverDate=11%2F11%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_searchStrId=1546173561&_rerunOrigin= google &_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=1642ebdad07f4009a3a7b81bb4d11c8f&searchtype=a

  1. Plasma rotation evolution near the peripheral transport barrier in the presence of low-frequency MHD bursts in TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Bulanin, V V; Askinazi, L G; Lebedev, S V; Gorohov, M V; Kornev, V A; Petrov, A V; Tukachinsky, A S; Vildjunas, M I

    2006-01-01

    The experiments described in the paper are aimed at investigating the possible influence of the low frequency magnetohydrodynamic (MHD) activity burst on the Ohmic H-mode in the TUMAN-3M tokamak. During the MHD burst a transient deterioration of improved confinement was observed. The study has been focused on the measurements of plasma fluctuation poloidal velocity performed by microwave Doppler reflectometry. The plasma fluctuation rotation observed before the MHD burst in the vicinity of the edge transport barrier was in the direction of plasma drift in the negative radial electric field. During the MHD activity the measured poloidal velocity was drastically decreased and even changed its sign. Radial profiles of the poloidal velocity measured in a set of reproducible tokamak shots exhibited the plasma fluctuation rotation in the ion diamagnetic drift direction at the location of the peripheral transport barrier. The possible reasons for this phenomenon are discussed

  2. Turbulent transport modeling in the edge plasma of tokamaks: verification, validation, simulation and synthetic diagnostics

    International Nuclear Information System (INIS)

    Colin-Bellot, Clothilde

    2015-01-01

    The possibility to produce power by using magnetically confined fusion is a scientific and technological challenge. The perspective of ITER conveys strong signals to intensify modeling effort on magnetized fusion plasmas. The success of the fusion operation is conditioned by the quality of plasma confinement in the core of the reactor and by the control of plasma exhaust on the wall. Both phenomena are related to turbulent cross-field transport that is at the heart of the notion of magnetic confinement studies, particle and heat losses. The study of edge phenomena is therefore complicated by a particularly complex magnetic geometry.This calls for an improvement of our capacity to develop numerical tools able to reproduce turbulent transport properties reliable to predict particle and energy fluxes on the plasma facing components. This thesis introduces the TOKAM3X fluid model to simulate edge plasma turbulence. A special focus is made on the code Verification and the Validation. It is a necessary step before using a code as a predictive tool. Then new insights on physical properties of the edge plasma turbulence are explored. In particular, the poloidal asymmetries induced by turbulence and observed experimentally in the Low-Field-Side of the devices are investigated in details. Great care is dedicated to the reproduction of the MISTRAL base case which consists in changing the magnetic configuration and observing the impact on parallel flows in the poloidal plane. The simulations recover experimental measurements and provide new insights on the effect of the plasma-wall contact position location on the turbulent features, which were not accessible in experiments. (author) [fr

  3. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    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 (n e and T e ) 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.)

  4. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  5. Properties of the tokamak edge plasma

    International Nuclear Information System (INIS)

    Wolff, H.

    1988-01-01

    A short review of some features of the edge plasma in limiter tokamaks is given. The limits of the simple one-dimensional scrape-off layer (SOL) model and the relation between the core plasma are discussed. Multifaceted asymmetric radiation from the edge (MARFE) phenomena and detached plasma are closely connected with the particle and energy balance of the SOL. Their occurrence is based on the relation of plasma parameters of the edge plasma to those of the core. Important problems of plasma wall interactions are the detection of the impurity sources and sinks and the study of the impurity transport and shielding. The non-uniform character of plasma wall interactions and their dependence on the discharge performance still renders difficult any theoretical forecast of impurity distribution and transport and calls for better diagnostics. (author)

  6. Tokamak plasma transport simulation in the presence of neoclassical tearing modes

    International Nuclear Information System (INIS)

    Takahashi, Y.; Yamazaki, K.; Arimoto, H.; Shoji, T.; Garcia, J.

    2008-01-01

    For the prediction of the ITER plasmas, the effect of the neoclassical tearing mode (NTM) on the plasma confinement has been calculated using the 1.5-dimensional equilibrium and transport simulation code TOTAL. The time evolution of the NTM magnetic island has been analyzed using the modified Rutherford equation for a ITER normal shear plasma. The anomalous transport model used here is GLF23. The saturated magnetic island widths are w/a - 0.048 at 3/2 mode and w/a - 0.21 at 2/1 mode, and the reduction in fusion power output by NTM is 27% at the 3/2 mode, 82% at the 2/1 mode, and 89% at the 3/2 + 2/1 double mode. The stabilization effect of the electron cyclotron current drive (ECCD) with EC is also clarified. The threshold of ECCD power for the full stabilization is ∼10[MW] against the 3/2 mode, and ∼23[MW] against the 2/1 mode. (author)

  7. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

    LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

    1988-04-01

    The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

  8. Topics in stability and transport in tokamaks: Dynamic transition to second stability with auxiliary heating; stability of global Alfven waves in an ignited plasma

    International Nuclear Information System (INIS)

    Fu, G.

    1988-01-01

    The problem of access to the high-beta ballooning second-stability regime by means of auxiliary heating and the problem of the stability of global-shear Alfven waves in an ignited tokamak plasma are theoretically investigated. These two problems are related to the confinement of both the bulk plasma as well as the fusion-product alpha particles an dare fundamentally important to the operation of ignited tokamak plasma. First, a model that incorporates both transport and ballooning mode stability was developed in order to estimate the auxiliary heating power required for tokamak plasma to evolve in time self-consistently into a high-beta, globally self-stabilized equilibrium. The critical heating power needed for access to second stability is found to be proportional to the square root of the anomalous diffusivity induced by the ballooning instability. Next, the full effects of toroidicity are retained in a theoretical description of global-type-shear Alfven modes whose stability can be modified by the fusion-product alpha particles that will present in an ignited tokamak plasma. Toroidicity is found to induce mode coupling and to stabilize the so-called Global Alfven Eigenmodes (GAE)

  9. Non-local transport in a tokamak plasma divertor with recycling

    International Nuclear Information System (INIS)

    Abou-Assaleh, Z.; Petravic, M.; Vesey, R.

    1993-01-01

    The plasma transport, particle and energy fluxes, near the diverter plate with high recycling has been modeled by using an electron kinetic code (Fokker-Planck International) in conjunction with a two-fluid ambipolar code. We include the effects of ionization and excitation of the hydrogen atoms. The electron energy distribution calculated from the kinetic code shows a large deviation from Maxwellian especially near the plate. This deviation from Maxwellian is due to the non-local transport of the suprathermal electrons from the SOL, and due also to the absorption of the fast electrons by the target plate. The heat flux near the plate is shown to be nonlocal, in that it is not determined uniquely by the local plasma parameters. Therefore the classical transport coefficients in the fluid model must be modified by including a nonlocal effect to produce the kinetic results. The kinetic calculation is compared with those of the fluid code with different values of the electron heat flux limiter factor (f). To reduce the computer load, the initial condition we used corresponds to the equilibrium solution already found with the fluid code with f=0.2. The fluid and Fokker-Planck codes are relaxed until all transients associated with electron dynamics have disappeared. In section 2, we present the kinetic code. The fluid code is presented in section 3. The boundary conditions used in these simulations are given in section 4. Finally the results and conclusion of these simulations are presented in section 5

  10. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

    Koniges, A.E.; Craddock, G.G.

    1992-08-01

    Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

  11. Radiation, impurity effects, instability characteristics and transport in Ohmically heated plasmas in the PLT tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Arunasalam, V.; Bitter, M.

    1979-01-01

    Titanium-gettered deuterium plasmas, with graphite or steel limiters to define the plasma minor radius, have Zsub(eff) approximately 1 for 3x10 13 14 cm -3 . In ungettered discharges the density limit set by disruptions is about half the value in gettered discharges. The bolometrically measured energy flux from the whole plasma volume is 80-100% of the Ohmic input power for ungettered discharges and 50-70% for gettered ones. The strucutre of MHD modes continues to be intensively studied by means of soft X-ray detector arrays; however, the connection with the disruptive instability remains unclear. Microinstabilities, studied by means of a 2-mm homodyne scattering system, appear to be of sufficient magnitude to influence energy and particle transport. Ion energy confinement times in the central region of the plasma have been estimated to be 50-100ms. Gross electron energy confinement time increases linearly with density at constant temperature. The longest electron energy confinement times observed are approximately >40ms in dense gettered discharges, giving total energy confinement times approximately 80ms. (author)

  12. Edge plasma physical investigations of tokamak plasmas in CRIP

    International Nuclear Information System (INIS)

    Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

    1988-01-01

    The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

  13. An axially averaged-radial transport model of tokamak edge plasmas

    International Nuclear Information System (INIS)

    Prinja, A.K.; Conn, R.W.

    1984-01-01

    A two-zone axially averaged-radial transport model for edge plasmas is described that incorporates parallel electron and ion conduction, localized recycling, parallel electron pressure gradient effects and sheath losses. Results for high recycling show that the radial electron temperature profile is determined by parallel electron conduction over short radial distances (proportional 3 cm). At larger radius where Tsub(e) has fallen appreciably, convective transport becomes equally important. The downstream density and ion temperature profiles are very flat over the region where electron conduction dominates. This is seen to result from a sharply decaying velocity profile that follows the radial electron temperature. A one-dimensional analytical recycling model shows that at high neutral pumping rates, the plasma density at the plate, nsub(ia), scales linearly with the unperturbed background density, nsub(io). When ionization dominates nsub(ia)/nsub(io) proportional exp(nsub(io)) while in the intermediate regime nsub(ia)/nsub(io) proportional exp(proportional nsub(io)). Such behavior is qualitatively in accord with experimental observations. (orig.)

  14. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  15. Theory of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

    1989-01-01

    The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

  16. Low Z impurity transport in tokamaks

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

    1978-10-01

    Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed

  17. Local transport in Joint European Tokamak edge-localized, high-confinement mode plasmas with H, D, DT, and T isotopes

    International Nuclear Information System (INIS)

    Budny, R. V.; Ernst, D. R.; Hahm, T. S.; McCune, D. C.; Christiansen, J. P.; Cordey, J. G.; Gowers, C. G.; Guenther, K.; Hawkes, N.; Jarvis, O. N.

    2000-01-01

    The edge-localized, high-confinement mode regime is of interest for future Tokamak reactors since high performance has been sustained for long durations. Experiments in the Joint European Tokamak [M. Keilhacker , Nuclear Fusion 39, 209 (1999)] have studied this regime using scans with the toroidal field and plasma current varied together in H, D, DT, and T isotopes. The local energy transport in more than fifty of these plasmas is analyzed, and empirical scaling relations are derived for energy transport coefficients during quasi-steady state conditions using dimensionless parameters. Neither the Bohm nor gyro-Bohm expressions give the shapes of the profiles. The scalings with β and ν * are in qualitative agreement with Ion Temperature Gradient theory

  18. Local transport barrier formation and relaxation in reverse-shear plasmas on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.

    1997-02-01

    The roles of turbulence stabilization by sheared E x B flow and Shafranov-shift gradients are examined for TFTR. Enhanced Reverse-Shear plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E x B and Shafranov shift effects was performed by varying the E x B shear through changes in the toroidal velocity with nearly-steady-state pressure profiles. Transport and fluctuation levels increase only when E x B shearing rates are driven below a critical value that is comparable to the fastest linear growth rates of the dominant instabilities. While a turbulence suppression criterion that involves the ratio of shearing to linear growth rates is in accord with many of these results, the existence of hidden dependencies of the criterion is suggested in experiments where the toroidal field was varied. The forward transition into the ERS regime has also been examined in strongly rotating plasmas. The power threshold is higher with unidirectional injection than with balanced injection

  19. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

    The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

  20. A study on the fusion reactor - Study on the plasma transport in advanced= tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Bae; Hong, Kyu Seon; Choi, Kang Oak [Soongsil University, Seoul (Korea, Republic of)

    1996-09-01

    Upper bound on the heat transport is considered with varying {eta}{sub i} and the curvature of the q profile. Nonlinear variational principle is adopted to= maximize the heat transport under the conditions of energy balance and the parallel energy balance. As a model, fluid plasma is assumed in the slab geometry. Resulting Lagrange equations are solved on the computer. Critical value of {eta}{sub i} energy stability is obtained and eigen states for higher {eta}{sub i}`s are found upon considering only the critical mode. Upper bound is found to be smaller when the curvature of q profile id of the same sign as the magnetic shear than the opposite case. As future studies, study on the bifurcation of the modes and, thus, inclusion of more modes are suggested. For more sophisticated model, there seems to be no untoward problems in applying the same method and, after such calculation is done, useful bounds with direct implications to real experiments will be predicted. Lower bound may be interesting, as well, and the work is in progress. 23 refs., 4 figs. (author)

  1. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

    A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

  2. Scrape-off layer tokamak plasma turbulence

    Science.gov (United States)

    Bisai, N.; Singh, R.; Kaw, P. K.

    2012-05-01

    Two-dimensional (2D) interchange turbulence in the scrape-off layer of tokamak plasmas and their subsequent contribution to anomalous plasma transport has been studied in recent years using electron continuity, current balance, and electron energy equations. In this paper, numerically it is demonstrated that the inclusion of ion energy equation in the simulation changes the nature of plasma turbulence. Finite ion temperature reduces floating potential by about 15% compared with the cold ion temperature approximation and also reduces the radial electric field. Rotation of plasma blobs at an angular velocity about 1.5×105 rad/s has been observed. It is found that blob rotation keeps plasma blob charge separation at an angular position with respect to the vertical direction that gives a generation of radial electric field. Plasma blobs with high electron temperature gradients can align the charge separation almost in the radial direction. Influence of high ion temperature and its gradient has been presented.

  3. Tokamak plasma boundary layer model

    International Nuclear Information System (INIS)

    Volkov, T.F.; Kirillov, V.D.

    1983-01-01

    A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics

  4. Transport barriers in plasmas

    International Nuclear Information System (INIS)

    Caldas, I L; Szezech, J D Jr; Kroetz, T; Marcus, F A; Roberto, M; Viana, R L; Lopes, S R

    2012-01-01

    We discuss the creation of transport barriers in magnetically confined plasmas with non monotonic equilibrium radial profiles. These barriers reduce the transport in the shearless region (i.e., where the twist condition does not hold). For the chaotic motion of particles in an equilibrium electric field with a nonmonotonic radial profile, perturbed by electrostatic waves, we show that a nontwist transport barrier can be created in the plasma by modifying the electric field radial profile. We also show non twist barriers in chaotic magnetic field line transport in the plasma near to the tokamak wall with resonant modes due to electric currents in external coils.

  5. Two-zone model for the transport of wall released impurities in the edge plasma of a limiter tokamak

    International Nuclear Information System (INIS)

    Claassen, H.A.; Gerhauser, H.

    1987-02-01

    The transmission of a Gaussian metal impurity pulse uniformly injected from the torus wall is studied within a two-zone plasma model, which separates the plasma into scrape-off layer and plasma core and allows for discontinuous changes of the plasma parameters at the separatrix. The plasma parameters are supposed to ensure a collision dominated scrape-off plasma, in which case we may restrict the solution of the transport equations to its zero order approximation. (orig./GG)

  6. Computational studies in tokamak equilibrium and transport

    International Nuclear Information System (INIS)

    Braams, B.J.

    1986-01-01

    This thesis is concerned with some problems arising in the magnetic confinement approach to controlled thermonuclear fusion. The work address the numerical modelling of equilibrium and transport properties of a confined plasma and the interpretation of experimental data. The thesis is divided in two parts. Part 1 is devoted to some aspects of the MHD equilibrium problem, both in the 'direct' formulation (given an equation for the plasma current, the corresponding equilibrium is to be determined) and in the 'inverse' formulation (the interpretation of measurements at the plasma edge). Part 2 is devoted to numerical studies of the edge plasma. The appropriate Navier-Stokes system of fluid equations is solved in a two-dimensional geometry. The main interest of this work is to develop an understanding of particle and energy transport in the scrape-off layer and onto material boundaries, and also to contribute to the conceptual design of the NET/INTOR tokamak reactor experiment. (Auth.)

  7. Energetics of turbulent transport processes in tokamaks

    International Nuclear Information System (INIS)

    Haas, F.A.; Thyagaraja, A.

    1987-01-01

    The effect of electromagnetic turbulence on electrons and ions under Tokamak conditions is considered using a kinetic description. Taking the magnetic fluctuation spectrum as given, the density fluctuation spectrum is self-consistently calculated taking account of quasi-neutrality. The calculation is valid for arbitrary collisionality and appropriate to low frequencies typical of experiment. In addition to the usual enhancement of the radial electron energy transport, it is found that the turbulent fluctuations can heat the plasma at rates comparable to ordinary ohmic heating under well-defined conditions. Interestingly, electromagnetic turbulence appears to imply only an insignificant correction to the toroidal resistance of the plasma as estimated from Spitzer resistivity. The scalings of anomalous transport, fluctuations and heating with temperature and plasma volume are investigated. The assumption that the magnetic fluctuation spectrum of the turbulence is invariant under a wide range of conditions is shown to result in interesting consequences for JET-like plasmas. (author)

  8. Plasma confinement modification and convective transport suppression in the scrape-off layer using additional gas puffing in the STOR-M tokamak

    International Nuclear Information System (INIS)

    Dreval, M; Hubeny, M; Ding, Y; Onchi, T; Liu, Y; Hthu, K; Elgriw, S; Xiao, C; Hirose, A

    2013-01-01

    The influence of short gas puffing (GP) pulses on the scrape-off layer (SOL) transport is studied. Similar responses of ion saturation current and floating potential measured near the GP injection valve and in the 90° toroidally separated cross-section suggest that the GP influence on the SOL region should be global. A drop in plasma temperature and a decrease in the rotational velocity of the plasma are observed in the SOL region immediately after the GP pulse; however, an unexpected increase in electron and ion temperatures is observed in the second stage of the plasma response. The decrease in floating potential fluctuations indicates that the turbulent transport is dumped immediately after the GP pulse. The GP-induced modification of turbulence properties in the SOL points to a convective transport suppression in the STOR-M tokamak. A substantial decrease in the skewness and kurtosis of ion saturation current fluctuations is observed in the SOL region resulting in the probability distribution function (PDF) getting closer to the Gaussian distribution. The plasma potential reduction, the change in plasma rotation and the suppression of turbulent transport in the SOL region indicate that the plasma confinement is modified after the GP injection. Some features of the H-mode-like confinement in the plasma bulk also accompany the SOL observations after application of the additional sharp GP pulse. (paper)

  9. Low Z impurity transport in tokamaks. [Neoclassical transport theory

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

    1978-10-01

    Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed.

  10. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Jaspers, R.E.

    1995-01-01

    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)

  11. The effects of magnetic perturbations on plasma transport or is magnetic turbulence important in tokamaks?

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-01-01

    A test particle model is verified and then used to interpret estimates of fast electron diffusivities in terms of magnetic fluctuation levels. The implied fluctuation levels are themselves interpreted with another verified model to predict electron thermal diffusivities. If the fast electron diffusivities represent local values, then the implied associated thermal transport is too small to explain experimental values. That is, magnetic fluctuations are not important. However, if the fast electron diffusivities represent effective values across mixed good (i.e. with no magnetic fluctuations) and bad (with magnetic fluctuations) surfaces then the implied magnetic fluctuation levels can influence electron thermal transport. (author)

  12. Convective transport in tokamaks

    International Nuclear Information System (INIS)

    D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.; Krasheninnikov, S.I.; Pigarov, A.Yu.; Yu, G.Q.; Xu, X.Q.; Nevins, W.M.

    2005-01-01

    Scrape-off-layer (SOL) convection in fusion experiments appears to be a universal phenomenon that can 'short-circuit' the divertor in some cases. The theory of 'blob' transport provides a simple and robust physical paradigm for studying convective transport. This paper summarizes recent advances in the theory of blob transport and its comparison with 2D and 3D computer simulations. We also discuss the common physical basis relating radial transport of blobs, pellets, and ELMs and a new blob regime that may lead to a connection between blob transport and the density limit. (author)

  13. Viscosity in the edge of tokamak plasmas

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1993-05-01

    A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ''short-radial-gradient-scale-length'' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates

  14. Simplified models for radiational losses calculating a tokamak plasma

    International Nuclear Information System (INIS)

    Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.

    1990-01-01

    To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs

  15. Application of optical fluctuation diagnostics to transport studies in high temperature tokamak plasmas

    International Nuclear Information System (INIS)

    Fonck, R.J.

    1991-01-01

    First data was obtained from the TFTR BES system in May 1990, after the prototype 4-channel system with photomultiplier detectors from PBX-M was transferred to the TFTR Hot Cell. The system was expanded to 16 spatial channels during the Summer 1990 run period, and data of interest was obtained near the end of the run in October 1990. The most attention has been given to a radial profile data set obtained for an L-mode plasma condition in TFTR, where we characterized the nature of the long-wavelength turbulence seen by the BES diagnostic. We summarize here highlights of the initial run with BES or TFTR. We also give a progress report on support activities at UW

  16. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  17. Gyrokinetic particle simulation of neoclassical transport in the pedestal/scrape-off region of a tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Ku, S [Courant Institute of Mathematical Sciences, New York University (United States); Chang, C-S [Courant Institute of Mathematical Sciences, New York University (United States); Adams, M [Columbia University (United States); Cummings, J [California Institute of Technology (United States); Hinton, F [Hinton Associates (United States); Keyes, D [Columbia University (United States); Klasky, S [Oak Ridge National Laboratory (United States); Lee, W [Princeton Plasma Physics Laboratory (United States); Lin, Z [University of California at Irvine (United States); Parker, S [University of Colorado at Boulder (United States)

    2006-09-15

    A gyrokinetic neoclassical solution for a diverted tokamak edge plasma has been obtained for the first time using the massively parallel Jaguar XT3 computer at Oak Ridge National Laboratory. The solutions show similar characteristics to the experimental observations: electric potential is positive in the scrape-off layer and negative in the H-mode layer, and the parallel rotation is positive in the scrape-off layer and at the inside boundary of the H-mode layer. However, the solution also makes a new physical discovery that there is a strong ExB convective flow in the scrape-off plasma. A general introduction to the edge simulation problem is also presented.

  18. ASTRA - an automatic system for transport analysis in a tokamak

    International Nuclear Information System (INIS)

    Pereverzev, G.V.; Yushmanov, P.N.; Dnestrovskii, A.Yu.; Polevoi, A.R.; Tarasjan, K.N.; Zakharov, L.E.

    1991-08-01

    The set of codes described here - ASTRA (Automatic System of Transport Analysis) - is a flexible and effective tool for the study of transport mechanisms in reactor-oriented facilities of the tokamak type. Flexibility is provided within the ASTRA system by a wide choice of standard relationships, functions and subroutines representing various transport coefficients, methods of auxiliary heating and other physical processes in the tokamak plasma, as well as by the possibility of pre-setting transport equations and variables for data output in a simple and conseptually transparent form. The transport code produced by the ASTRA system provides an adequate representation of the discharges for present experimental conditions. (orig.)

  19. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

    1998-01-01

    The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

  20. Simulation models for tokamak plasmas

    International Nuclear Information System (INIS)

    Dimits, A.M.; Cohen, B.I.

    1992-01-01

    Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth

  1. SCORCH - a zero dimensional plasma evolution and transport code for use in small and large tokamak systems

    International Nuclear Information System (INIS)

    Clancy, B.E.; Cook, J.L.

    1984-12-01

    The zero-dimensional code SCORCH determines number density and temperature evolution in plasmas using concepts derived from the Hinton and Hazeltine transport theory. The code uses the previously reported ADL-1 data library

  2. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    International Nuclear Information System (INIS)

    Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

    2016-01-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  3. Transport Barriers in Bootstrap Driven Tokamaks

    Science.gov (United States)

    Staebler, Gary

    2017-10-01

    Maximizing the bootstrap current in a tokamak, so that it drives a high fraction of the total current, reduces the external power required to drive current by other means. Improved energy confinement, relative to empirical scaling laws, enables a reactor to more fully take advantage of the bootstrap driven tokamak. Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is due to the suppression of turbulence primarily due to the large Shafranov shift. ExB velocity shear does not play a significant role in the transport barrier due to the high safety factor. It will be shown, that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift driven barrier formation. The ion energy transport is reduced to neoclassical and electron energy and particle transport is reduced, but still turbulent, within the barrier. Deeper into the plasma, very large levels of electron transport are observed. The observed electron temperature profile is shown to be close to the threshold for the electron temperature gradient (ETG) mode. A large ETG driven energy transport is qualitatively consistent with recent multi-scale gyrokinetic simulations showing that reducing the ion scale turbulence can lead to large increase in the electron scale transport. A new saturation model for the quasilinear TGLF transport code, that fits these multi-scale gyrokinetic simulations, can match the data if the impact of zonal flow mixing on the ETG modes is reduced at high safety factor. This work was supported by the U.S. Department of Energy under DE-FG02-95ER54309 and DE-FC02

  4. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    2000-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge MHD instabilities and plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. The article examines these phenomena and their interaction. (author)

  5. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    1999-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  6. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    2001-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  7. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  8. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  9. Transport, chaos and plasma physics

    International Nuclear Information System (INIS)

    Benkadda, S.; Doveil, F.; Elskens, Y.

    1993-01-01

    This workshop made it possible to gather for the first time plasma physicists, dynamical systems physicists and mathematicians, around a general theme focusing on the characterisation of chaotic transport. The participations have been divided into 4 topics: - dynamical systems and microscopic models of chaotic transport, - magnetic fluctuations and transport in tokamaks, - drift wave turbulence, self-organisation and intermittency, and - Wave-particle interactions

  10. Effects of the location of a biased limiter on turbulent transport in the IR-T1 tokamak plasma

    International Nuclear Information System (INIS)

    Alipour, R.; Ghoranneviss, M.; Salar Elahi, A.; Meshkani, S.

    2017-01-01

    Plasma confinement plays an important role in fusion study. Applying an external voltage using limiter biasing system is proved to be an efficient approach for plasma confinement. In this study, the position of the limiter biasing system was changed to investigate the effect of applying external voltages at different places to the plasma. The external voltages of ±200 V were applied at the different positions of edge, 5 mm and 10 mm inside the plasma. Then, the main plasma parameters were measured. The results show that the poloidal turbulent transport and radial electric field increased about 25-35% and 35-45%, respectively (specially when the limiter biasing system was placed 5 mm inside the plasma). Also, the Reynolds stress has experienced its maximum reduction about 5-10% when the limiter biasing system was at 5 mm inside the plasma and the voltage of +200 V was applied to the plasma. When the limiter biasing system move 10 mm inside the plasma, the main plasma parameters experienced more instabilities and fluctuations than other positions. (authors)

  11. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1985-01-01

    The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

  12. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

  13. Tokamak nonmaxwellian plasma dynamics in thermonuclear regime

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1987-01-01

    To reach ignition in a Tokamak plasma, large additional power P aux has to be injected in the device on top of the Joule heating P OH =VI r , V the plasma loop voltage, I r the resistive port of plasma current. Typi-cally JH ∼ 1 KeV, whereas ignition would requi- re IG ∼ 7-10 KeV. To gain this factor 7, one at least should inject additional power P aux ∼ 7P OH , supposing that nothing, especially the heat transport, is modified. This is by far not the case, with the so-called energy lifetime degradation, largely observed in oil experiments (but less dramatic with divertors), where energy lifetime tau E behaves like P tot -b with b∼1/2. In large machines where ignition temperature is the target to be imperiously reached, this implies to inject a very large power, typically P aux ∼ 50 to 100 MW, depending on size and parameters and on actual transport. So it is of importance with such figures, or even larger ones owing to uncertain ties, to optimize at best injected power by increasing its efficiency, both with respect to possible transport laws, and to physical phenomena governing heat flow in the system from the sources. This leads to the concept of scenarios, as time sequences of power input, where physical properties of the plasma system are used to build up ion temperature so that ignition is reached with minimum P tot = P OH + P aux and with fixed Q = Q o > 1. Elements for this study are given. The method is outlined. The resulting system of equations describing the evolution of a thermonuclear plasma is given

  14. Neoclassical MHD descriptions of tokamak plasmas

    International Nuclear Information System (INIS)

    Callen, J.D.; Kim, Y.B.; Sundaram, A.K.

    1988-01-01

    Considerable progress has been made in extending neoclassical MHD theory and in exploring the linear instabilities, nonlinear behavior and turbulence models it implies for tokamak plasmas. The areas highlighted in this paper include: extension of the neoclassical MHD equations to include temperature-gradient and heat flow effects; the free energy and entropy evolution implied by this more complete description; a proper ballooning mode formalism analysis of the linear instabilities; a new rippling mode type instability; numerical simulation of the linear instabilities which exhibit a smooth transition from resistive ballooning modes at high collisionality to neoclassical MHD modes at low collisionality; numerical simulation of the nonlinear growth of a single helicity tearing mode; and a Direct-Interaction-Approximation model of neoclassical MHD turbulence and the anomalous transport it induces which substantially improves upon previous mixing length model estimates. 34 refs., 2 figs

  15. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

    Vayakis, G.

    1991-04-01

    The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

  16. Multiscale coherent structures in tokamak plasma turbulence

    International Nuclear Information System (INIS)

    Xu, G. S.; Wan, B. N.; Zhang, W.; Yang, Q. W.; Wang, L.; Wen, Y. Z.

    2006-01-01

    A 12-tip poloidal probe array is used on the HT-7 superconducting tokamak [Li, Wan, and Mao, Plasma Phys. Controlled Fusion 42, 135 (2000)] to measure plasma turbulence in the edge region. Some statistical analysis techniques are used to characterize the turbulence structures. It is found that the plasma turbulence is composed of multiscale coherent structures, i.e., turbulent eddies and there is self-similarity in a relative short scale range. The presence of the self-similarity is found due to the structural similarity of these eddies between different scales. These turbulent eddies constitute the basic convection cells, so the self-similar range is just the dominant scale range relevant to transport. The experimental results also indicate that the plasma turbulence is dominated by low-frequency and long-wavelength fluctuation components and its dispersion relation shows typical electron-drift-wave characteristics. Some large-scale coherent structures intermittently burst out and exhibit a very long poloidal extent, even longer than 6 cm. It is found that these large-scale coherent structures are mainly contributed by the low-frequency and long-wavelength fluctuating components and their presence is responsible for the observations of long-range correlations, i.e., the correlation in the scale range much longer than the turbulence decorrelation scale. These experimental observations suggest that the coexistence of multiscale coherent structures results in the self-similar turbulent state

  17. Zeff measurements and low-Z impurity transport for NBI and ICRF heated plasma in JIPP T-IIU tokamak

    International Nuclear Information System (INIS)

    Ida, K.; Amano, T.; Kawahata, K.; Kaneko, O.

    1988-12-01

    A visible bremsstrahlung detector array system for Z eff measurements and a charge exchange recombination spectroscopy (CXRS) system for fully ionized impurity profile measurements were installed on JIPP TII-U to study impurity transport for NBI and ICRF heated plasma. More impurities are sputtered by ICRF heating than by NBI and/or ohmic heatings. The carbon contribution to Z eff is 80-90 % for NBI heated plasmas, and 60 % for NBI + ICRF heated plasmas. With a carbon coating of vacuum vessel, the Z eff value decreases 2.4 to 1.7 and the carbon contribution to Z eff increases up to 80-90 %. We obtain the diffusion coefficient D a = 1.0 m 2 /s and the convective velocity V a (a) = 13 m/s at the plasma edge for carbon impurity from the radial profile and time evolution of fully ionized carbon after the ICRF pulse is turned on. (author)

  18. Plasma diagnostics using synchrotron radiation in tokamaks

    International Nuclear Information System (INIS)

    Fidone, I.; Giruzzi, G.; Granata, G.

    1995-09-01

    This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs

  19. Increase in beta limit in tokamak plasmas

    International Nuclear Information System (INIS)

    Kamada, Yutaka

    2003-01-01

    This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

  20. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    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

  1. Problems with the concept of plasma equilibrium in tokamaks

    International Nuclear Information System (INIS)

    Carreras, B.A.

    1992-01-01

    The equilibrium condition for a magnetically confined plasma in normally formulated in terms of macroscopic equations. In these equations, the plasma pressure is assumed to be a function of the magnetic flux with continuous derivatives. However, in three- dimensional systems this is not necessarily the case. Here, we look at the case of an intrinsically three-dimensional realistic tokamak, and we discuss the possible interconnection between the equilibrium and anomalous transport

  2. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

    From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

  3. On the HL-1M tokamak plasma confinement time

    International Nuclear Information System (INIS)

    Qin Yunwen

    2001-01-01

    Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed

  4. Theory of incremental turbulent transport in tokamaks

    International Nuclear Information System (INIS)

    Similon, P.L.

    1991-01-01

    The goal of this research is to understand how the various aspect of turbulent transport operate in tokamaks, in the presence of low frequency fluctuations such as drift waves or trapped electron modes

  5. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.; Bell, R.E.

    2001-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  6. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.; Bell, R.E.

    1999-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  7. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

    Xu, X.; Umansky, M.; Dudson, B.; Snyder, P.

    2008-05-01

    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 e ; T 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

  8. Physics of transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Garbet, X [Assoc. EURATOM-CEA, CEA/DSM/DRFC CEA-Cadarache, 13108 Saint Paul lez Durance (France); Mantica, P [Istituto di Fisica del Plasma EURATOM-ENEA/CNR, via Cozzi 53, 20125 Milan (Italy); Angioni, C [MPI fuer Plasmaphysik, EURATOM-Assoz., D-8046 Garching (Germany)] [and others

    2004-12-01

    This paper is an overview of recent results relating to turbulent particle and heat transport, and to the triggering of internal transport barriers (ITBs). The dependence of the turbulent particle pinch velocity on plasma parameters has been clarified and compared with experiment. Magnetic shear and collisionality are found to play a central role. Analysis of heat transport has made progress along two directions: dimensionless scaling laws, which are found to agree with the prediction for electrostatic turbulence, and analysis of modulation experiments, which provide a stringent test of transport models. Finally the formation of ITBs has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear, combined with the Shafranov shift, is a robust stabilizing mechanism. However, some well established features of internal barriers are not explained by theory.

  9. Multi-field plasma sandpile model in tokamaks and applications

    Science.gov (United States)

    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.

  10. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  11. A complex probe for tokamak plasma edge conditions

    International Nuclear Information System (INIS)

    Castro, R.M. de; Silva, R.P. da; Heller, M.V.A.P.; Caldas, I.L.; Nascimento, I.C.; Degasperi, F.T.

    1995-01-01

    The study of the physical processes that occur in the plasma edge of tokamak machines has recently grown due to the evidence that these processes influence those that occur in the center of the plasma column. Experimental studies show the existence of a strong level of fluctuations in the plasma edge. The results of these studies indicate that these fluctuations enhance particle and energy transport and degrade the confinement. In order to investigate these processes in the plasma edge of the TBR-1 Tokamak, a Langmuir probe array, a triple and a set of magnetic probes have been designed and constructed. With this set probes the mean and fluctuation values of the magnetic field were detected and correlated with the fluctuating parameters obtained with the electrostatic probes. (author). 7 refs., 5 figs

  12. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

    2001-01-01

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  13. Plasma edge physics in an actively cooled tokamak

    International Nuclear Information System (INIS)

    Gunn, J.P.; Adamek, A.; Boucher, C.

    2005-01-01

    Tore Supra is a large tokamak with a plasma of circular cross section (major radius 2.4 m and minor radius 0.72 m) lying on a toroidal limiter. Tore Supra's main mission is the development of technology to inject up to 25 MW of microwave heating power and extract it continuously for up to 1000 s in steady state without uncontrolled overheating of, or outgassing from, plasma-facing components. The entire first wall of the tokamak is actively cooled by a high pressure water loop and special carbon fiber composite materials have been designed to handle power fluxes up to 10 MW/m 2 . The edge plasma on open magnetic flux surfaces that intersect solid objects plays an important role in the overall behaviour of the plasma. The transport of sputtered impurity ions and the fueling of the core plasma are largely governed by edge plasma density, temperature, and flow profiles. Measurements of these quantities are becoming more reliable and frequent in many tokamaks, and it has become clear that we do not understand them very well. Classical two-dimensional fluid modelling fails to reproduce many aspects of the experimental observations such as the significant thickness of the edge plasma, and the near-sonic flows that occur where none should be expected. It is suspected that plasma turbulence is responsible for these anomalies. In the Tore Supra tokamak, various kinds of Langmuir probes are used to characterize the edge plasma. We will present original measurements that demonstrate the universality of many phenomena that have been observed in X-point divertor tokamaks, especially concerning the ion flows. As in the JET tokamak, surprisingly large values of parallel Mach number are measured midway between the two strike zones, where one would expect to find nearly stagnant plasma if the particle source were poloidally uniform. We will present results of a novel experiment that provides evidence for a poloidally localized particle and energy source on the outboard midplane of

  14. Physics of transport in Tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Asp, E.; Bourdelle, C.; Hoang, T.; Imbeaux, F.; Joffrin, E.; Litaudon, X.; Angioni, C.; Manini, A.; Peeter, A.; Ryter, F.; Baranov, Y.; Cordey, G.; McDonald, D.C.; Parail, V.; Valovic, M.; Thyagaraja, A.; Voitsekhovitch, I.; Budny, R.; Kirneva, N.; Hogeweij, D.; Nordman, H.; Weilland, J.; Tala, T.; Weisen, H.; Zabolotsky, A.

    2004-01-01

    This paper is an overview of recent results related to turbulent particle and heat transport, and triggering of Internal Transport Barriers. Particle transport is characterised by a pinch velocity that is found to be larger than the value predicted by theory of collisional transport (Ware pinch) in L-mode plasmas. Also it increases with magnetic shear. In H-mode, density peaking decreases with collisionality. Pinch velocity reaches the Ware value for large collision frequencies. Heat transport has made progress along two directions: dimensionless scaling laws and analysis of modulation experiments. Dimensionless scaling law of thermal confinement agrees with the prediction for electrostatic turbulence. Dependence with collisionality remains to be understood. Heat modulation experiments have been investigated in several devices using a critical gradient model, micro-stability analysis and predictive modelling. Thresholds and stiffness are correctly reproduced by stability analysis and modelling with Weiland model. Analysis with a critical gradient model leads to a large variability of stiffness. Finally the question of triggering Internal Transport Barriers has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear combined with Shafranov shift is a robust stabilizing mechanism. However, some well established features of internal barriers are hardly explained by theory, in particular the role of low order rational values of the minimum safety factor, and the existence of multiple barriers. (authors)

  15. Empirical particle transport model for tokamaks

    International Nuclear Information System (INIS)

    Petravic, M.; Kuo-Petravic, G.

    1986-08-01

    A simple empirical particle transport model has been constructed with the purpose of gaining insight into the L- to H-mode transition in tokamaks. The aim was to construct the simplest possible model which would reproduce the measured density profiles in the L-regime, and also produce a qualitatively correct transition to the H-regime without having to assume a completely different transport mode for the bulk of the plasma. Rather than using completely ad hoc constructions for the particle diffusion coefficient, we assume D = 1/5 chi/sub total/, where chi/sub total/ ≅ chi/sub e/ is the thermal diffusivity, and then use the κ/sub e/ = n/sub e/chi/sub e/ values derived from experiments. The observed temperature profiles are then automatically reproduced, but nontrivially, the correct density profiles are also obtained, for realistic fueling rates and profiles. Our conclusion is that it is sufficient to reduce the transport coefficients within a few centimeters of the surface to produce the H-mode behavior. An additional simple assumption, concerning the particle mean-free path, leads to a convective transport term which reverses sign a few centimeters inside the surface, as required by the H-mode density profiles

  16. Internal transport barrier physics for steady state operation in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Wakatani, Masahiro [Kyoto Univ., Graduate School of Engineering, Uji, Kyoto (Japan); Fukuda, Takeshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Connor, Jack W. [Culham Science Centre, EURATOM/UKAEA Association (United Kingdom); Garbet, Xavier [Culham Science Centre, EFDA-JET CSU (United Kingdom); Gormezano, Claude [Associazone EURATOM-ENEA sulla Fusione C.R. Frascati (Italy); Mukhovatov, Vladimir [ITER Naka Joint Work Site, ITER Physics Unit, Naka, Ibaraki (Japan)

    2003-07-01

    Experimental results for the ITB (Internal Transport Barrier) formation and sustainment are compiled in a unified manner to find common features of ITBs in tokamaks. Global scaling laws for threshold power to obtain the ITBs are discussed. Theoretical models for plasmas with ITBs are summarized from stability and transport point of view. Finally possibility to obtain steady-state ITBs will be discussed in addition to extrapolation to ITER. (author)

  17. Investigation of the impurity transport in the ASDEX tokamak by spectroscopical methods

    International Nuclear Information System (INIS)

    Krieger, K.W.

    1990-12-01

    Plasma impurities: a central problem of controlled thermonuclear fusion; magnetic plasma confinement in a Tokamak; methods to the determination of plasma impurity transport coefficients - by temporally modulated gas admission; the transport equation for impurities; neoclassical and anomalous transport; harmonic analysis of time-dependent signals; solutions of the transport equation; experimental equipment and measurements; measuring results - consistency of simple transport models with radial phase measurements; linearity of the transport processes; plasma disturbance by impurity injection; determination of the diffusion coefficient by simplified transport models; comparison of transport models for impurities and background plasma; measurements of the impurity transport at the plasma edge by high modulation frequencies. (AH)

  18. Computer simulation of transport driven current in tokamaks

    International Nuclear Information System (INIS)

    Nunan, W.J.; Dawson, J.M.

    1993-01-01

    Plasma transport phenomena can drive large currents parallel to an externally applied magnetic field. The Bootstrap Current Theory accounts for the effect of Banana diffusion on toroidal current, but the effect is not confined to that transport regime. The authors' 2 1/2-D, electromagnetic, particle simulations have demonstrated that Maxwellian plasmas in static toroidal and vertical fields spontaneously develop significant toroidal current, even in the absence of the open-quotes seed currentclose quotes which the Bootstrap Theory requires. Other simulations, in both toroidal and straight cylindrical geometries, and without any externally imposed electric field, show that if the plasma column is centrally fueled, and if the particle diffusion coefficient exceeds the magnetic diffusion coefficient (as is true in most tokamaks) then the toroidal current grows steadily. The simulations indicate that such fueling, coupled with central heating due to fusion reactions may drive all of the tokamak's toroidal current. The Bootstrap and dynamo mechanisms do not drive toroidal current where the poloidal magnetic field is zero. The simulations, as well as initial theoretical work, indicate that in tokamak plasmas, various processes naturally transport current from the outer regions of the plasma to the magnetic axis. The mechanisms which cause this effective electron viscosity include conventional binary collisions, wave emission and reabsorption, and also convection associated with rvec E x rvec B vortex motion. The simulations also exhibit preferential loss of particles carrying current opposing the bulk plasma current. This preferential loss generates current even at the magnetic axis. If these self-seeding mechanisms function in experiments as they do in the simulations, then transport driven current would eliminate the need for any external current drive in tokamaks, except simple ohmic heating for initial generation of the plasma

  19. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  20. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    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.

    2004-03-01

    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

  1. Plasma internal inductance dynamics in a tokamak

    International Nuclear Information System (INIS)

    Romero, J.A.

    2010-01-01

    A lumped parameter model for tokamak plasma current and inductance time evolution as a function of plasma resistance, non-inductive current drive sources and boundary voltage or poloidal field coil current drive is presented. The model includes a novel formulation leading to exact equations for internal inductance and plasma current dynamics. Having in mind its application in a tokamak inductive control system, the model is expressed in state space form, the preferred choice for the design of control systems using modern control systems theory. The choice of system states allows many interesting physical quantities such as plasma current, inductance, magnetic energy, and resistive and inductive fluxes be made available as output equations. The model is derived from energy conservation theorem, and flux balance theorems, together with a first order approximation for flux diffusion dynamics. The validity of this approximation has been checked using experimental data from JET showing an excellent agreement.

  2. Effects of radial electrical field on neoclassical transport in tokamaks

    International Nuclear Information System (INIS)

    Wang Zhongtian; Le Clair, G.

    1996-07-01

    Neoclassical transport theory for tokamaks in presence of a radial electrical field with shear is developed using Hamiltonian formalism. Diffusion coefficients are derived in both plateau regime including a large electric field and banana regime including the squeezing factor which can greatly affect diffusion at the plasma edge. The scaling on squeezing factor is different from the one given by Shaing and Hazeltine. Rotation speeds are calculated in the scrape-off region. They are in good agreement with measurements on TdeV Tokamak. (2 figs.)

  3. Electron cyclotron heating (ECH) of tokamak plasmas

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1990-01-01

    Electron cyclotron heating (ECH) is one of the intense methods of plasma heating, and which utilizes the collisionless electron-cyclotron-resonance-interaction between the launched electromagnetic waves (called electron cyclotron waves) and electrons which are one of the constituents of the high temperature plasmas. Another constituent, namely the ions which are subject to nuclear fusion, are heated indirectly but strongly and instantly (in about 0.1 s) by the collisions with the ECH-heated electrons in the fusion plasmas. The recent progress on the development of high-power and high-frequency millimeter-wave-source enabled the ECH experiments in the middle size tokamaks such as JFT-2M (Japan), Doublet III (USA), T-10 (USSR) etc., and ECH has been demonstrated to be the sure and intense plasma heating method. The ECH attracts much attention for its remarkable capabilities; to produce plasmas (pre-ionization), to heat plasmas, to drive plasma current for the plasma confinement, and recently especially by the localization and the spatial controllability of its heating zone, which is beneficial for the fine controls of the profiles of plasma parameters (temperature, current density etc.), for the control of the magnetohydrodynamic instabilities, or for the optimization/improvement of the plasma confinement characteristics. Here, the present status of the ECH studies on tokamak plasmas are reviewed. (author)

  4. Analysis of tokamak plasma confinement modes using the fast

    Indian Academy of Sciences (India)

    The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

  5. Heating of plasmas in tokamaks by current-driven turbulence

    International Nuclear Information System (INIS)

    Kluiver, H. de.

    1985-10-01

    Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued

  6. Studies of the disruption prevention by ECRH at plasma current rise stage in limiter discharges/Possibility of an internal transport barrier producing under dominating electron transport in the T-10 tokamak

    International Nuclear Information System (INIS)

    Alikaev, V.V.; Borshegovskij, A.A.; Chistyakov, V.V.

    2001-01-01

    'Studies of the Disruption Prevention by ECRH at Plasma Current Rise Stage in Limiter Discharges' - Studies of disruption prevention by means of ECRH in T-10 at the plasma current rise phase in limiter discharges with circular plasma cross-section were performed. Reliable disruption prevention by ECRH at HF power (P HF ) min level equal to 20% of ohmic heating power P OH was demonstrated. m/n=2/1 mode MHD-activity developed before disruption (with characteristic time ∼ 120 ms) can be considered as disruption precursor and can be used in a feedback system. 'Possibility of an Internal Transport Barrier Producing under Dominating Electron Transport in the T-10 Tokamak' - The reversed shear experiments were carried out on T-10 at the HF power up to 1MW. The reversed shear in the core was produced by on-axis ECCD in direction opposite to the plasma current. There are no obvious signs of Internal Transport Barriers formation under condition when high-k turbulence determines the electron transport. (author)

  7. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    International Nuclear Information System (INIS)

    Castracane, J.

    2001-01-01

    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

  8. 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.

  9. Application studies of spherical tokamak plasma merging

    International Nuclear Information System (INIS)

    Ono, Yasushi; Inomoto, Michiaki

    2012-01-01

    The experiment of plasma merging and heating has long history in compact torus studies since Wells. The study of spherical tokamak (ST), starting from TS-3 plasma merging experiment of Tokyo University in the late 1980s, is followed by START of Culham laboratory in the 1900s, TS-4 and UTST of Tokyo University and MAST of Culham laboratory in the 2000s, and last year by VEST of Soul University. ST has the following advantages: 1) plasma heating by magnetic reconnection at a MW-GW level, 2) rapid start-up of high beta plasma, 3) current drive/flux multiplication and distribution control of ST plasma, 4) fueling and helium-ash exhaust. In the present article, we emphasize that magnetic reconnection and plasma merging phenomena are important in ST plasma study as well as in plasma physics. (author)

  10. [High beta tokamak research and plasma theory

    International Nuclear Information System (INIS)

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report

  11. Carbon impurity transport around limiters in the DITE tokamak

    International Nuclear Information System (INIS)

    Pitcher, C.S.; Stangeby, P.C.; Goodall, D.H.J.; Matthews, G.F.; McCracken, G.M.

    1989-01-01

    The transport of impurity ions originating at the limiter in a tokamak is critically dependent on the location of the ion in the boundary plasma. In the confined plasma, just inboard of the limiter, impurity ions will disperse freely into the discharge whilst in the scrape-off layer the pre-sheath plasma flow and the associated ambipolar electric field may tend to sweep impurities back to the limiter surface. In this paper we have studied, both by experiment and by theory, the transport of carbon impurity ions in the vicinity of the limiter. By comparing experimental measurements of the spatial distributions of impurities around the limiter with that predicted from a Monte Carlo computer code it appears that the parallel dispersal on closed field lines in the confined plasma is consistent with classical transport processes and that in the scrape-off layer the dispersal is indeed impeded by the pre-sheath plasma flow. (orig.)

  12. Temperature and impurity transport studies of heated tokamak plasmas by means of a collisional-radiative model of x-ray emission from Mo30+ to Mo39+

    International Nuclear Information System (INIS)

    Pacella, D.; Fournier, K. B.; Zerbini, M.; Finkenthal, M.; Mattioli, M.; May, M. J.; Goldstein, W. H.

    2000-01-01

    This work presents and interprets, by means of detailed atomic calculations, observations of L-shell (n=3→n=2) transitions in highly ionized molybdenum, the main intrinsic heavy impurity in the Frascati tokamak upgrade plasmas. These hot plasmas were obtained by additional electron cyclotron resonance heating (ECRH), at the frequency of 140 Ghz, during the current ramp-up phase of the discharge. Injecting 400 kW on axis and 800 kW slightly off axis, the peak central electron temperature reached 8.0 and 7.0 keV, respectively, for a time much longer than the ionization equilibrium time of the molybdenum ions. X-ray emissions from rarely observed high charge states, Mo 30+ to Mo 39+ , have been studied with moderate spectral resolution (λ/Δλ∼150) and a time resolution of 5 ms. A sophisticated collisional-radiative model for the study of molybdenum ions in plasmas with electron temperature in the range 4-20 keV is presented. The sensitivity of the x-ray emission to the temperature and to impurity transport processes is discussed. This model has been then used to investigate two different plasma scenarios. In the first regime the ECRH heating occurs on axis during the current ramp up phase, when the magnetic shear is evolving from negative to zero up to the half radius. The spectrum is well reproduced with the molybdenum ions in coronal equilibrium and with a central impurity peaking. In the second regime, at the beginning of the current flat top when magnetic shear is monotonic and sawtoothing activity is appearing, the lowest charge states (Mo 33+ to Mo 30+ ), populated off axis, are affected by anomalous transport and the total molybdenum profile is found to be flat up to the half radius. We conclude with the presentation of ''synthetic spectra'' computed for even higher temperature plasmas that are expected in future experiments with higher ECRH power input. (c) 2000 The American Physical Society

  13. Plasma transport in the stochastic fields at the tokamak edge. Final report, February 15, 1993--February 14, 1994

    International Nuclear Information System (INIS)

    Punjabi, A.

    1994-01-01

    The purpose of this project is to calculate the contribution of field line diffusion to particle diffusion in the stochastic magnetic field at the tokamak edge. The author uses the approach of quasi magnetic surfaces. If the magnetic field line makes sufficiently large number of toroidal transits before suffering large radial excursion, then the method of quasi magnetic surface is valid for this problem. This method has three components: determination of particle drift trajectories, a model for magnetic field configuration, and determination of field line diffusion

  14. Magnetohydrodynamic stability of tokamak edge plasmas

    International Nuclear Information System (INIS)

    Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.

    1998-01-01

    A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement open-quotes Hclose quotes-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates ∼n 1/3 rational surfaces into the plasma (rather than ∼n 1/2 , expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle

  15. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Buchenauer, D.; Chance, M.

    1986-04-01

    Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit

  16. An overview on plasma disruption mitigation and avoidance in tokamak

    International Nuclear Information System (INIS)

    He Kaihui; Pan Chuanhong; Feng Kaiming

    2002-01-01

    Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

  17. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  18. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  19. Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel

    2008-01-01

    Roč. 15, č. 8 (2008), 082505-1-082505-8 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146

  20. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

  1. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

    Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

    2010-11-01

    The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

  2. Neutron measurement techniques for tokamak plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1994-01-01

    The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)

  3. Semi-analytic modeling of tokamak particle transport

    International Nuclear Information System (INIS)

    Shi Bingren; Long Yongxing; Li Jiquan

    2000-01-01

    The linear particle transport equation of tokamak plasma is analyzed. Particle flow consists of an outward diffusion and an inward convection. General solution is expressed in terms of a Green function constituted by eigen-functions of corresponding Sturm-Liouville problem. For a particle source near the plasma edge (shadow fueling), a well-behaved solution in terms of Fourier series can be constituted by using the complementarity relation. It can be seen from the lowest eigen-function that the particle density becomes peaked when the wall recycling reduced. For a transient point source in the inner region, a well-behaved solution can be obtained by the complementarity as well

  4. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.

    1990-01-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)

  5. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.

    1990-06-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab

  6. Diffusive heat transport across magnetic islands and stochastic layers in tokamaks

    International Nuclear Information System (INIS)

    Hoelzl, Matthias

    2010-01-01

    Heat transport in tokamak plasmas with magnetic islands and ergodic field lines was simulated at realistic plasma parameters in realistic tokamak geometries. This requires the treatment of anisotropic heat diffusion, which is more efficient along magnetic field lines by up to ten orders of magnitude than perpendicular to them. Comparisons with analytical predictions and experimental measurements allow to determine the stability properties of neoclassical tearing modes as well as the experimental heat diffusion anisotropy.

  7. Effect of lower hybrid waves on turbulence and transport of particles and energy in the FTU tokamak scrape-off layer plasma

    Energy Technology Data Exchange (ETDEWEB)

    Ridolfini, V Pericoli [ENEA-CR Frascati, Via Enrico Fermi 45-00044 Frascati, Roma (Italy)

    2011-11-15

    All the main features of the scrape-off layer turbulence, magnitude, frequency spectrum and perpendicular wave vector, {xi}{sub t}, are strongly affected by the injection of lower hybrid (LH) power into the FTU tokamak. The governing parameters are the local last closed magnetic surface values of density, n{sub e,LCMS}, and temperature, T{sub e,LCMS}. n{sub e,LCMS} determines the perpendicular wave vector of the LH waves, which is a key parameter for the multiple scattering processes, and together with T{sub e,LCMS} the collisionality that exerts a stabilizing effect on the fluctuations. This effect, still to be examined in the light of theoretical models, leads to an asymptotic value for the fluctuation relative amplitude in the ohmic phase close to 25%, and {approx}10% in the LH phase, or even less, since the saturation level is not yet attained. The LH waves also can strongly raise {xi}{sub t}, about 3 times, and double the root mean square frequency. The transfer of momentum and energy in the mutual scattering of LH and turbulence 'waves' drives these changes. An increase also of the cross-correlation between temperature and electric potential fluctuations should occur in order to explain the magnitude of the fluctuation amplitude drop and the large increment of the temperature e-folding decay, by more than a factor of 2.5. Particle transport, however, does not appear to be affected to a large extent-the density e-folding decay length is almost unchanged but the power flow typical length rises by about a factor of 1.5, which is a relevant figure in view of the problem of mitigating the power loads on divertor targets in future reactors. These changes are confined mainly within the flux tube connected with the LH waves launching antenna, but start to spread significantly out of it at high plasma densities.

  8. Effect of lower hybrid waves on turbulence and transport of particles and energy in the FTU tokamak scrape-off layer plasma

    International Nuclear Information System (INIS)

    Ridolfini, V Pericoli

    2011-01-01

    All the main features of the scrape-off layer turbulence, magnitude, frequency spectrum and perpendicular wave vector, ξ t , are strongly affected by the injection of lower hybrid (LH) power into the FTU tokamak. The governing parameters are the local last closed magnetic surface values of density, n e,LCMS , and temperature, T e,LCMS . n e,LCMS determines the perpendicular wave vector of the LH waves, which is a key parameter for the multiple scattering processes, and together with T e,LCMS the collisionality that exerts a stabilizing effect on the fluctuations. This effect, still to be examined in the light of theoretical models, leads to an asymptotic value for the fluctuation relative amplitude in the ohmic phase close to 25%, and ∼10% in the LH phase, or even less, since the saturation level is not yet attained. The LH waves also can strongly raise ξ t , about 3 times, and double the root mean square frequency. The transfer of momentum and energy in the mutual scattering of LH and turbulence 'waves' drives these changes. An increase also of the cross-correlation between temperature and electric potential fluctuations should occur in order to explain the magnitude of the fluctuation amplitude drop and the large increment of the temperature e-folding decay, by more than a factor of 2.5. Particle transport, however, does not appear to be affected to a large extent-the density e-folding decay length is almost unchanged but the power flow typical length rises by about a factor of 1.5, which is a relevant figure in view of the problem of mitigating the power loads on divertor targets in future reactors. These changes are confined mainly within the flux tube connected with the LH waves launching antenna, but start to spread significantly out of it at high plasma densities.

  9. Impurity flux collection at the plasma edge of the tokamak MT-1

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Bakos, J.S.; Petravich, G.

    1989-09-01

    Fluxes of intrinsic and injected impurities and background plasma ions were collected using a bidirectional probe at the plasma edge of the tokamak MT-1. The directional and radial dependences of injected impurities and plasma ions were very similar indicating a strong coupling of the impurity transport to the dynamics of the background plasma. The measured intrinsic concentration of about 10 -4 for Mo at the plasma edge is derived. (author) 17 refs.; 5 figs

  10. Plasma startup patterns in tokamak reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Tone, Tatsuzo.

    1983-01-01

    Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)

  11. Comments on experimental results of energy confinement of tokamak plasmas

    International Nuclear Information System (INIS)

    Chu, T.K.

    1989-04-01

    The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ΔW/ΔP is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ν/sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab

  12. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

    htmlabstractOne 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

  13. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

    Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

    1999-01-01

    This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

  14. Turbulent ion heating in TCV Tokamak plasmas

    International Nuclear Information System (INIS)

    Schlatter, Ch.

    2009-08-01

    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

  15. Arbitrary poloidal gyroradius effects in tokamak pedestals and transport barriers

    International Nuclear Information System (INIS)

    Kagan, Grigory; Catto, Peter J

    2008-01-01

    A technique is developed and applied for analyzing pedestal and internal transport barrier (ITB) regions in a tokamak by formulating a special version of gyrokinetics. In contrast to typical gyrokinetic treatments, canonical angular momentum is taken as the gyrokinetic radial variable rather than the radial guiding center location. Such an approach allows strong radial plasma gradients to be treated, while retaining zonal flow and neoclassical (including orbit squeezing) behavior and the effects of turbulence. The new, nonlinear gyrokinetic variables are constructed to higher order than is typically the case. The nonlinear gyrokinetic equation obtained is capable of handling such problems as collisional zonal flow damping with radial wavelengths comparable to the ion poloidal gyroradius, as well as zonal flow and neoclassical transport in the pedestal or ITB. This choice of gyrokinetic variables allows the toroidally rotating Maxwellian solution of the isothermal tokamak limit to be recovered. More importantly, we prove that a physically acceptable solution for the lowest order ion distribution function in the banana regime anywhere in a tokamak and, in particular, in the pedestal must be nearly this same isothermal Maxwellian solution. That is, the ion temperature variation scale must be much greater than the poloidal ion gyroradius. Consequently, in the banana regime the background radial ion temperature profile cannot have a pedestal similar to that of plasma density

  16. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

    1997-01-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

  17. Plasma engineering analyses of tokamak reactor operating space

    International Nuclear Information System (INIS)

    Houlberg, W.; Attenberger, S.E.

    1981-01-01

    A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models

  18. Plasma residual poloidal rotation in TCABR tokamak

    International Nuclear Information System (INIS)

    Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.

    2003-01-01

    This paper reports the first measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak for a collisional plasma (Pfirsch-Schluter regime), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0)·10 3 m/s at r ∼ 2/3a, (a - limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) · 10 3 m/s for the plasma core to a co-current value of (2.0 ± 1.0) · 10 3 m/s near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner. (author)

  19. A general comparison between tokamak and stellarator plasmas

    Directory of Open Access Journals (Sweden)

    Yuhong Xu

    2016-07-01

    Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.

  20. Stability of tearing modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Hegna, C.C.; Callen, J.D.

    1994-02-01

    The stability properties of m ≥ 2 tearing instabilities in tokamak plasmas are analyzed. A boundary layer theory is used to find asymptotic solutions to the ideal external kink equation which are used to obtain a simple analytic expression for the tearing instability parameter Δ'. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of Δ' are obtained for plasmas with finite beta and for islands that have nonzero width. A simple heuristic estimate is given for the value of the saturated island width when the instability criterion is violated. A connection is made between the calculation of the asymptotic matching parameter in the finite beta and island width case to the nonlinear analog of the Glasser effect

  1. Sensitivity of transient synchrotron radiation to tokamak plasma parameters

    International Nuclear Information System (INIS)

    Fisch, N.J.; Kritz, A.H.

    1988-12-01

    Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs

  2. Technology and plasma-materials interaction processes of tokamak disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.; Kellman, A.G.

    1992-01-01

    A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs

  3. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Medley, S.S.; Young, K.M.

    1988-06-01

    The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10 21 m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs

  4. Shear flow effects on ion thermal transport in tokamaks

    International Nuclear Information System (INIS)

    Tajima, T.; Horton, W.; Dong, J.Q.; Kishimoto, Y.

    1995-03-01

    From various laboratory and numerical experiments, there is clear evidence that under certain conditions the presence of sheared flows in a tokamak plasma can significantly reduce the ion thermal transport. In the presence of plasma fluctuations driven by the ion temperature gradient, the flows of energy and momentum parallel and perpendicular to the magnetic field are coupled with each other. This coupling manifests itself as significant off-diagonal coupling coefficients that give rise to new terms for anomalous transport. The authors derive from the gyrokinetic equation a set of velocity moment equations that describe the interaction among plasma turbulent fluctuations, the temperature gradient, the toroidal velocity shear, and the poloidal flow in a tokamak plasma. Four coupled equations for the amplitudes of the state variables radially extended over the transport region by toroidicity induced coupling are derived. The equations show bifurcations from the low confinement mode without sheared flows to high confinement mode with substantially reduced transport due to strong shear flows. Also discussed is the reduced version with three state variables. In the presence of sheared flows, the radially extended coupled toroidal modes driven by the ion temperature gradient disintegrate into smaller, less elongated vortices. Such a transition to smaller spatial correlation lengths changes the transport from Bohm-like to gyrobohm-like. The properties of these equations are analyzed. The conditions for the improved confined regime are obtained as a function of the momentum-energy deposition rates and profiles. The appearance of a transport barrier is a consequence of the present theory

  5. Plasma residual rotation in the TCABR tokamak

    International Nuclear Information System (INIS)

    Severo, J.H.F.; Nascimento, I.C.; Tsypin, V.S.; Galvao, R.M.O.

    2003-01-01

    This paper reports the first results on the measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak, in the collisional regime (Pfirsch-Schluter), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5±1.0) x 10 3 m s -1 at r ∼ 2/3a,(a-limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) x 10 3 m s -1 , at the plasma core, to a co-current value of (2.0 ± 0.9) x 10 3 m s -1 near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner (1991 Phys. Fluids B 3 2050). (author)

  6. Modeling plasma/material interactions during a tokamak disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1994-10-01

    Disruptions in tokamak reactors are still of serious concern and present a potential obstacle for successful operation and reliable design. Erosion of plasma-facing materials due to thermal energy dump during a disruption can severely limit the lifetime of these components, therefore diminishing the economic feasibility of the reactor. A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The initial burst of energy delivered to facing-material surfaces from direct impact of plasma particles causes sudden ablation of these materials. As a result, a vapor cloud is formed in front of the incident plasma particles. Shortly thereafter, the plasma particles are stopped in the vapor cloud, heating and ionizing it. The energy transmitted to the material surfaces is then dominated by photon radiation. It is the dynamics and the evolution of this vapor cloud that finally determines the net erosion rate and, consequently, the component lifetime. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics, and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed

  7. Theory for neoclassical toroidal plasma viscosity in tokamaks

    International Nuclear Information System (INIS)

    Shaing, K C; Chu, M S; Hsu, C T; Sabbagh, S A; Seol, Jae Chun; Sun, Y

    2012-01-01

    Error fields and magnetohydrodynamic modes break toroidal symmetry in tokamaks. The broken symmetry enhances the toroidal plasma viscosity, which results in a steady-state toroidal plasma flow. A theory for neoclassical toroidal plasma viscosity in the low-collisionality regimes is developed. It extends stellarator transport theory to include multiple modes and to allow for |m − nq| ∼ 1. Here, m is the poloidal mode number, n is the toroidal mode number and q is the safety factor. The bounce averaged drift kinetic equation is solved in several asymptotic limits to obtain transport fluxes. These fluxes depend non-linearly on the radial electric field except for those in the 1/ν regime. Here, ν is the collision frequency. The theory is refined to include the effects of the superbanana plateau resonance at the phase space boundary and the finite ∇B drift on the collisional boundary layer fluxes. Analytical expressions that connect all asymptotic limits are constructed and are in good agreement with the numerical results. The flux–force relations that relate transport fluxes to forces are used to illustrate the roles of transport fluxes in the momentum equation. It is shown that the ambipolar state is reached when the momentum equation is relaxed. It is also shown that the origin of the momentum for plasma flow generated without momentum sources is the local unbalance of particles' momenta and is diamagnetic in nature regardless of the details of the theory. (paper)

  8. 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.

  9. Plasma confinement using biased electrode in the TCABR tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.; Machida, M.

    2005-01-01

    Experimental data obtained on the TCABR tokamak (R = 0.61 m, r = 0.18 m) with an electrally polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1), and hydrogen gas injection adjusted for keeping the electron density at 1.0x10(19) m(-3) without bias. Temporal and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the density, up to a maximum factor of 2.6, while H-alpha hydrogen spectral line intensity decreases, and the CIII impurity stays on the same level. The analysis of temporal and radial profiles of plasma parameters indicates that the confined plasma entered in the H-mode regime. The data analysis shows a maximum enhanced confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the plasma edge, using measured floating potentials and ion saturation currents, show strong decrease in the power spectra and transport. Bifurcation was not observed, and the decrease in the saturation current occurs in 50 microseconds. (author)

  10. Theory-based scaling of the SOL width in circular limited tokamak plasmas

    International Nuclear Information System (INIS)

    Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Silva, C.; Gunn, J.P.; Horacek, J.; Kočan, M.; LaBombard, B.

    2013-01-01

    A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement. (letter)

  11. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

    2005-01-01

    Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

  12. The major tokamak distruption in cylindrical plasma

    International Nuclear Information System (INIS)

    Choi, Jeong Sik; Choi, Eun Ha; Choi, Duk In

    1986-01-01

    The mechanism of the major disruption in tokamak plasma which involves the nonlinear interaction of tearing models is numerically studied in two and three dimensional formulations. In this study, it is found that in the two dimensional case with a flattened current density profile the magnetic islands of the m=2; n=1 mode do not saturate nonlinearly and but strongly interact with the limiter. Thus it is suggested that the helical perturbation of the m=2;n=1 mode plays the dominant role in the major disruption. We also show that the m=2;n=1 mode nonlinearly destablizes other tearing modes, especially the m=3;n=2 mode, from the nonlinear coupling of different helicities as also shown in other studies. The plasma extends across the plasma cross section, and the plasma core shifts inward along the major radius during the major disruption. The numerical result for the major disruption time measured using the nonlinear 3-D procedure for the initial value problem with PLT parameters is about 450 μsec which agrees reasonably well with the experimental value of 500 μsec. (Author)

  13. Tokamak-FED plasma-engineering assessments

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lyon, J.F.; Rutherford, P.H.

    1981-01-01

    A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m 2 . The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts

  14. Energy confinement of tokamak plasma with consideration of bootstrap current effect

    International Nuclear Information System (INIS)

    Yuan Ying; Gao Qingdi

    1992-01-01

    Based on the η i -mode induced anomalous transport model of Lee et al., the energy confinement of tokamak plasmas with auxiliary heating is investigated with consideration of bootstrap current effect. The results indicate that energy confinement time increases with plasma current and tokamak major radius, and decreases with heating power, toroidal field and minor radius. This is in reasonable agreement with the Kaye-Goldston empirical scaling law. Bootstrap current always leads to an improvement of energy confinement and the contraction of inversion radius. When γ, the ratio between bootstrap current and total plasma current, is small, the part of energy confinement time contributed from bootstrap current will be about γ/2

  15. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

  16. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

    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. (auth)

  17. The COMPASS Tokamak Plasma Control Software Performance

    Science.gov (United States)

    Valcarcel, Daniel F.; Neto, André; Carvalho, Ivo S.; Carvalho, Bernardo B.; Fernandes, Horácio; Sousa, Jorge; Janky, Filip; Havlicek, Josef; Beno, Radek; Horacek, Jan; Hron, Martin; Panek, Radomir

    2011-08-01

    The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.

  18. Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)

    2010-09-01

    One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.

  19. Theory of ion heat transport in tokamaks

    International Nuclear Information System (INIS)

    Gott, Y.V.; Yurchenko, E.I.

    1987-01-01

    Experiments which have been carried out in several tokamaks to determine the ion thermal conductivity show that it is several times the value predicted by the neoclassical theory. A possible explanation for this discrepancy is proposed. When the finite width of a banana is taken into account, there are substantial increases in the heat fluxes which stem from the important contribution of superthermal ions to the transport. If the electron diffusive flux is zero, a systematic account of the ions with E>T leads to an ion heat flux with a finite banana width which is two to four times the neoclassical prediction. The effect of the anomalous nature of the electron flux on the ion heat transport is analyzed. An expression is derived for calculating the ion heat transport over the entire range of collision rates

  20. LH-power coupling in advanced tokamak plasmas in JET

    International Nuclear Information System (INIS)

    Joffrin, E.; Erents, K.; Gormezano, C.

    2000-02-01

    Lower Hybrid Current Drive (LHCD) is the most efficient tool to generate non-inductive current in tokamak plasmas. In JET, significant modifications of the current profile have been recently achieved in coupling up to 3MW of LH power in optimised shear discharges. However, the improved particle confinement during optimised shear plasmas results in a sharp decrease of the electron density in front the launcher close or below the cut-off density (ne=1.7.10 17 m -3 for f LH =37GHz) and makes difficult the coupling of the LH power. Deuterium gas near the launcher can help to improve the coupling, but has also the effect of increasing the ELM activity leading to the erosion of the internal transport barrier (ITB). Future development of lower hybrid launcher should include the constraints imposed by scenario such as the optimised shear. (author)

  1. Plasma confinement using biased electrode in the TCABR tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Machida, M.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.

    2005-01-01

    Experimental data obtained on the TCABR tokamak (R = 0.61 m, a = 0.18 m) with an electrically polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1) and hydrogen gas injection adjusted for keeping the electron density at 1.0 x 10 19 m -3 without bias. Time evolution and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the central line-averaged density, up to a maximum factor of 2.6, while H α hydrogen spectral line intensity decreases and the C III impurity stays on the same level. The analysis of temporal behaviour and radial profiles of plasma parameters indicates that the confined plasma enters the H-mode regime. The data analysis shows a maximum enhanced energy confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the Scrape-Off-Layer, using measured floating potentials and ion saturation currents, show a strong decrease in the power spectra and transport. Bifurcation was not observed and the decrease in the saturation current occurs in 50 μs

  2. Edge Plasma Response to Non-Axisymmetric Fields in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, N. M.; Lao, L. L.; Buttery, R. J.; Evans, T. E.; Snyder, P. B.; Wade, M.R., E-mail: ferraro@fusion.gat.com [General Atomics, San Diego (United States); Moyer, R. A.; Orlov, D. M. [University of California San Diego, La Jolla (United States); Lanctot, M. J. [Lawrence Livermore National Laboratory, Livermore (United States)

    2012-09-15

    Full text: The application of non-axisymmetric fields is found to have significant effects on the transport and stability of H-mode tokamak plasmas. These effects include dramatic changes in rotation and particle transport, and may lead to the partial or complete suppression of edge-localized modes (ELMs) under some circumstances. The physical mechanism underlying these effects is presently not well understood, in large part because the response of the plasma to non- axisymmetric fields is significant and complex. Here, recent advances in modeling the plasma response to non-axisymmetric fields are discussed. Calculations using a resistive two-fluid model in diverted toroidal geometry confirm the special role of the perpendicular electron velocity in suppressing the formation of islands in the plasma. The possibility that islands form near the top of the pedestal, where the zero-crossing of the perpendicular electron velocity may coincide with a mode-rational surface, is explored, and the implications for ELM suppression are discussed. Modeling results are compared with empirical data. It is shown that numerical modeling is successful in reproducing some experimentally observed effects of applied non-axisymmetric fields on the edge temperature and density profiles. The numerical model self-consistently includes the plasma, separatrix, and scrape-off layer. Rotation and diamagnetic effects are also included self-consistently. Solutions are calculated using the M3D-C1 extended-MHD code. (and others)

  3. Calculation of the neoclassical conductivity of plasma and fraction of trapped particles for elongated Damavand Tokamak

    International Nuclear Information System (INIS)

    Dini, F.; Khorasani, S.

    2007-01-01

    Configuration of Tokamak plasma has a dominant effect on its parameters. In the calculation of transport, there are some transport coefficients and quantities, where the knowledge of their precise values, according to the system of equations, is essential to be realized. Tokamak has a toroidal configuration, in addition to classical effects, it is necessary to study the neoclassical effects due to the field curvature. The trapped particles in strong electromagnetic fields oscillate on banana-shaped orbits which in turn affect many other collisional transport parameters. Here, a precise estimation of trapped particles based on the standard equilibrium model for an elliptical shape of Tokamak plasma has been carried out using Lin-Liu model. It should be added that in this calculation, the profile of the averaged magnetic field on the flux surfaces has been derived using analytical integration and consideration of an elliptic shape for ellipticity function in the limit of large aspect ratio and zero shift of magnetic flux surfaces. Having the fraction of the trapped particles, by ,following the formulation and using an appropriate model in various collisional regimes, the neoclassical conductivity of plasma in Damavand Tokamak is obtained and the respective variations have been found. The presented results can exploit the computation of transport and other quantities of Damavand Tokamak

  4. Helical temperature perturbations associated with tearing modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    1994-06-01

    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

  5. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    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 ...

  6. Kinetic and transport theory near the tokamak edge

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Catto, P.J.

    1995-12-01

    Conventional transport orderings employed in the core of a tokamak plasma allow large divergence-free flows in flux surfaces, but only weak radial flows. However, alternate orderings are required in the edge region where radial diffusion must balance the rapid loss due to free-streaming to divertor plates or limiters. Kinetic equations commonly used to study the plasma core do not allow such a balance and are, therefore, inapplicable in the plasma edge. Similarly, core transport formulae cannot be extended to the edge region without major, qualitative alteration. Here the authors address the necessary changes. By deriving and solving a novel kinetic equation, they construct distinctive collisional transport laws for the plasma edge. They find that their edge ordering naturally retains the radial diffusion and parallel flow of particles, momentum and heat to lowest order in the conservation equations. To higher order they find a surprising form for parallel transport in the scrape-off layer, in which the parallel flow of particles and heat are driven by a combination of the conventional gradients, viscosity, and new terms involving radial derivatives. The new terms are not relatively small, and could affect understanding of limiter and divertor operation

  7. 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].

  8. Theory-based transport simulation of tokamaks: density scaling

    International Nuclear Information System (INIS)

    Ghanem, E.S.; Kinsey, J.; Singer, C.; Bateman, G.

    1992-01-01

    There has been a sizeable amount of work in the past few years using theoretically based flux-surface-average transport models to simulate various types of experimental tokamak data. Here we report two such studies, concentrating on the response of the plasma to variation of the line averaged electron density. The first study reported here uses a transport model described by Ghanem et al. to examine the response of global energy confinement time in ohmically heated discharges. The second study reported here uses a closely related and more recent transport model described by Bateman to examine the response of temperature profiles to changes in line-average density in neutral-beam-heated discharges. Work on developing a common theoretical model for these and other scaling experiments is in progress. (author) 5 refs., 2 figs

  9. Turbulence, transport and confinement: from tokamaks to star magnetism

    International Nuclear Information System (INIS)

    Strugarek, Antoine

    2012-01-01

    This thesis is part of the general study of self-organization in hot and magnetized plasmas. We focus our work on two specific objects: stars and tokamaks. We use first principle numerical simulations to study turbulence, transport and confinement in these plasmas. The first part of this thesis introduces the main characteristics of stellar and tokamak plasmas. The reasons for studying them together are properly detailed. The second part is focused on stellar aspects. We study the interactions between the 3D turbulent motions in the solar convection zone with an internal magnetic field in the tachocline (the transition region between the instable and stable zones in the Sun). The tachocline is a very thin layer (less than five percent of the solar radius) that acts as a transport barrier of angular momentum. We show that such an internal magnetic field is not likely to explain the observed thickness of the tachocline and we give some insights on how to find alternative mechanisms to constrain it. We also explore the effect of the environment of star on its structure. We develop a methodology to study the influence of stellar wind and of the magnetic coupling of a star with its orbiting planets. We use the same methodology to analyse the magnetic interaction between a stellar wind and a planetary magnetosphere that acts as a transport barrier of matter. Then, the third part is dedicated to fusion oriented research. We present a numerical investigation on the experimental mechanisms that lead to the development of transport barriers in the plasma. These barriers are particularly important for the design of high performance fusion devices. The creation of transport barriers is obtained in turbulent first principle simulations for the very first time. The collaboration between the two scientific teams lead to the results presented in the fourth part of this thesis. An original spectral method is developed to analyse the saturation of stellar convective dynamos and of

  10. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    Atzeni, S.; Vlad, G.

    1985-01-01

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D- 3 He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D- 3 He tokamak plasma

  11. Experimental observations related to the thermodynamic properties of tokamak plasmas

    International Nuclear Information System (INIS)

    Sozzi, C.; Minardi, E.; Lazzaro, E.; Cirant, S.; Mantica, P.; Esposito, B.; Marinucci, M.; Romanelli, M.; Imbeaux, F.

    2005-01-01

    The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)

  12. Plasma Confinement in the UCLA Electric Tokamak.

    Science.gov (United States)

    Taylor, Robert J.

    2001-10-01

    The main goal of the newly constructed large Electric Tokamak (R = 5 m, a = 1 m, BT 8 x 10^12 cm-3 when there is no MHD activity. The electron temperature, derived from the plasma conductivity is > 250 eV with a central electron energy confinement time > 350 msec in ohmic conditions. The sawteeth period is 50 msec. Edge plasma rotation is induced by plasma biasing via electron injection in an analogous manner to that seen in CCT(R.J. Taylor, M.L. Brown, B.D. Fried, H. Grote, J.R. Liberati, G.J. Morales, P. Pribyl, D. Darrow, and M. Ono. Phys. Rev Lett. 63 2365 1989.) and the neoclassical bifurcation is close to that described by Shaing et al(K.C. Shaing and E.C. Crume, Phys. Rev. Lett. 63 2369 (1989).). In the ohmic phase the confinement tends to be MHD limited. The ICRF heating eliminates the MHD disturbances. Under second harmonic heating conditions, we observe an internal confinement peaking characterized by doubling of the core density and a corresponding increase in the central electron temperature. Charge exchange data, Doppler data in visible H-alpha light, and EC radiation all indicate that ICRF heating works much better than expected. The major effort is focused on increasing the power input and controlling the resulting equilibrium. This task appears to be easy since our current pulses are approaching the 3 second mark without RF heating or current drive. Our initial experience with current profile control, needed for high beta plasma equilibrium, will be also discussed.

  13. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1986-01-01

    It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism

  14. Measurements of plasma position in TJ-I Tokamak

    International Nuclear Information System (INIS)

    Qin, J.; Ascasibar, E.; Navarro, A.P.; Ochando, M.A.; Pastor, I.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.; Team, TJ-I.

    1994-01-01

    This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs

  15. Models for impurity production and transport in tokamaks

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1976-01-01

    Models for the edge conditions which are commonly used in tokamak transport codes have been kept simple partly because of a lack of data. A report is presented on an improved model for the particle and energy balance of e - , H 1 + , H 1 0 , H 2 + , H 2 0 , O 0 , O/sup (1 yields 8) + / in the plasma scrape-off region. Experiments should yield the needed data in the near future, and allow one to test the model. The diffusion of impurities has been studied with a neoclassical model. The role of 'anomalous spreading' of the impurity distribution has been studied for the case of Fe. A model is presented for the expulsion of low-Z (oxygen) impurities for cases where q(0) greater than 1, but in which a large shear-free region is produced in the plasma core

  16. Introduction to turbulent transport in fusion plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    2006-01-01

    This introduction presents the main instabilities responsible for turbulence in tokamak plasmas, and the prominent features of the resulting transport. The usual techniques to construct reduced transport models are described. These models can be tested by analysing steady state and transient regimes. Another way to test the theory is to use a similarity principle, similar to the one used in fluid mechanics. Finally, the physics involved in the formation and sustainment of transport barriers is presented. (author)

  17. Tokamak fuelling with pellets: Effect of transport phenomena on the injection requirements

    International Nuclear Information System (INIS)

    Lengyel, L.L.

    1979-01-01

    Results of calculations on pellet-plasma interaction that take into account transport phenomena inherent in tokamak plasmas are analyzed. It is shown that the results obtained by different authors on the optimum pellet penetration depth and required pellet injection frequencies, which are partly contradictory, can be explained by means of the different transport processes taken into account or neglected in the calculations concerned. (orig.)

  18. Control of plasma position in the CASTOR tokamak

    International Nuclear Information System (INIS)

    Valovic, M.

    1988-11-01

    A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

  19. Magnetic diagnostic plasma position in the TCA/BR tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu.K.; Nascimento, I.C.

    1996-01-01

    The cross-section of the plasma column is TCA/BR has a nearly circular plasma shape. This allows implementation of simplified methods of magnetic diagnostics. Although these methods were in may tokamaks and are well described, their accuracies are not clearly defined because the very simplified theoretical model of plasma equilibrium on which they are based differs from the real conditions in tokamaks like TCA/BR. In this paper we present the methods of plasma position diagnostics in TCA/BR from external magnetic measurements with an error analysis. (author). 4 refs., 3 figs

  20. Generation of plasma rotation by ICRH in tokamaks

    International Nuclear Information System (INIS)

    Chang, C.; Phillips, C.K.; White, R.B.; Zweben, S.; Bonoli, P.T.; Rice, J.; Greenwald, M.; Grassie, J.S. de

    2001-01-01

    A physical mechanism to generate plasma rotation by ICRH is presented in a tokamak geometry. By breaking the omnigenity of resonant ion orbits, ICRH can induce a non-ambipolar minor-radial flow of resonant ions. This induces a return current j p r in the plasma, which then drives plasma rotation through the j p r xB force. It is estimated that the fast-wave power in the present-day tokamak experiments can be strong enough to give a significant modification to plasma rotation. (author)

  1. Modelling of electron transport and of sawtooth activity in tokamaks

    International Nuclear Information System (INIS)

    Angioni, C.

    2001-10-01

    Transport phenomena in tokamak plasmas strongly limit the particle and energy confinement and represent a crucial obstacle to controlled thermonuclear fusion. Within the vast framework of transport studies, three topics have been tackled in the present thesis: first, the computation of neoclassical transport coefficients for general axisymmetric equilibria and arbitrary collisionality regime; second, the analysis of the electron temperature behaviour and transport modelling of plasma discharges in the Tokamak a configuration Variable (TCV); third, the modelling and simulation of the sawtooth activity with different plasma heating conditions. The work dedicated to neoclassical theory has been undertaken in order to first analytically identify a set of equations suited for implementation in existing Fokker-Planck codes. Modifications of these codes enabled us to compute the neoclassical transport coefficients considering different realistic magnetic equilibrium configurations and covering a large range of variation of three key parameters: aspect ratio, collisionality, and effective charge number. A comparison of the numerical results with an analytical limit has permitted the identification of two expressions for the trapped particle fraction, capable of encapsulating the geometrical effects and thus enabling each transport coefficient to be fitted with a single analytical function. This has allowed us to provide simple analytical formulae for all the neoclassical transport coefficients valid for arbitrary aspect ratio and collisionality in general realistic geometry. This work is particularly useful for a correct evaluation of the neoclassical contribution in tokamak scenarios with large bootstrap cur- rent fraction, or improved confinement regimes with low anomalous transport and for the determination of the plasma current density profile, since the plasma conductivity is usually assumed neoclassical. These results have been included in the plasma transport code

  2. Thermal stability of the tokamak plasma edge

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1997-01-01

    The general linear, fluid, thermal instability theory for the plasma edge has been extended. An analysis of a two-dimensional fluid model of the plasma edge has identified the importance of many previously unappreciated phenomena associated with parallel and gyroviscous forces in the presence of large radial gradients, with large radial or parallel flows, with the temperature dependence of transport coefficients, and with the coupling of temperature, flow and density perturbations. The radiative condensation effect is generalized to include a further destabilizing condensation effect associated with radial heat conduction. Representative plasma edge neutral and impurity densities are found to be capable of driving thermal instabilities in the edge transport barrier and radiative mantle, respectively. (author)

  3. Parametric dependences of impurity transport in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Parisot, Th.

    2007-09-01

    During this Ph.D. work, a full setup of tools for an experimental investigation of impurity transport has been developed on the Tore Supra tokamak. It includes a laser blow-off system for metallic impurity injections and developments for ITC (Impurity Transport Code), a transport code which allows the extraction of the experimental impurity transport coefficients (diffusion and convection velocity). This tool has been used to perform and analyse several experiments, to evidence parametric dependences of impurity transport. In a first experiment, a confinement time law for nickel in Tore Supra has been obtained as a function of collisionality ν * and normalized Larmor radius ρ * . Then the impurity charge Z role has been investigated in various conditions: ohmic regime with or without sawteeth, and sawtooth less L-mode with LH power. No Z effect is observed, consistently with theoretical predictions, whether neoclassical (NCLASS) or for turbulent transport with both non linear gyro-fluid (TRB) and quasilinear gyrokinetic (QuaLiKiz) simulations. An exception is found for LH heated plasmas where the confinement time seems to decrease for the heaviest impurities. This is not explained by any model available. The observed transport is close to neoclassical between sawtooth relaxations, in the centre (r q-1 ) of ohmic plasmas, turbulent outside. Without sawteeth, it is turbulent in the whole plasma, for ohmic or L mode discharges. The profile shape of the diffusion coefficient is here qualitatively different, with a stronger and deeper transition between the low diffusion central region and a more turbulent peripheral region for LH heated plasmas. (author)

  4. ECRH and electron heat transport in tokamaks

    International Nuclear Information System (INIS)

    Zou, X.L.; Giruzzi, G.; Dumont, R.J.

    2003-01-01

    It has been observed during the ECRH experiments in tokamaks that the shape of the electron temperature profile in stationary regimes is not very sensitive to the ECRH power deposition i.e. the temperature profile remains peaked at the center even though the ECRH power deposition is off-axis. Various models have been invoked for the interpretation of this profile resilience phenomenon: the inward heat pinch, the critical temperature gradient, the Self-Organized Criticality, etc. Except the pinch effect, all of these models need a specific form of the diffusivity in the heat transport equation. In this work, our approach is to solve a simplified time-dependent heat transport equation analytically in cylindrical geometry. The features of this analytical solution are analyzed, in particular the relationship between the temperature profile resilience and the Eigenmode of the physical system with respect to the heat transport phenomenon. Finally, applications of this analytical solution for the determination of the transport coefficient and the polarization of the EC waves are presented. It has been shown that the solution of the simplified transport equation in a finite cylinder is a Fourier-Bessel series. This series represents in fact a decomposition of the heat source in Eigenmode, which are characterized by the Bessel functions of order 0. The physical interpretation of the Eigenmodes is the following: when the heat source is given by a Bessel function of order 0, the temperature profile has exactly the same form as the source at every time. At the beginning of the power injection, the effectiveness of the temperature response is the same for each Eigenmode, and the response in temperature, having the same form as the source, is local. Conversely, in the later phase of the evolution, the effectiveness of the temperature response for each Eigenmode is different: the higher the order, the lower the effectiveness. In this case the response in temperature appears as

  5. Lower hybrid current drive in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  6. Lower hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi

    1999-03-01

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar e - 10 20 m -3 , ALCATOR-C) and the highest current drive efficiency (η CD = 3.5x10 19 m -2 A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  7. Sustained high βN plasmas on EAST tokamak

    Science.gov (United States)

    Gao, Xiang; the EAST team

    2018-05-01

    Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.

  8. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  9. Modelisation of synchrotron radiation losses in realistic tokamak plasmas

    International Nuclear Information System (INIS)

    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)

  10. Spectra of heliumlike krypton from tokamak fusion test reactor plasmas

    International Nuclear Information System (INIS)

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Smith, A.; Fraenkel, B.

    1993-04-01

    Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 Angstrom including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport

  11. Electron thermal transport in tokamak: ETG or TEM turbulences?

    International Nuclear Information System (INIS)

    Lin, Z.; Chen, L.; Nishimura, Y.; Qu, H.; Hahm, T.S.; Lewandowski, J.; Rewoldt, G.; Wang, W.X.; Diamond, P.H.; Holland, C.; Zonca, F.; Li, Y.

    2005-01-01

    This paper reports progress on numerical and theoretical studies of electron transport in tokamak including: (1) electron temperature gradient turbulence; (2) trapped electron mode turbulence; and (3) a new finite element solver for global electromagnetic simulation. In particular, global gyrokinetic particle simulation and nonlinear gyrokinetic theory find that electron temperature gradient (ETG) instability saturates via nonlinear toroidal couplings, which transfer energy successively from unstable modes to damped modes preferably with longer poloidal wavelengths. The electrostatic ETG turbulence is dominated by nonlinearly generated radial streamers. The length of streamers scales with the device size and is much longer than the distance between mode rational surfaces or electron radial excursions. Both fluctuation intensity and transport level are independent of the streamer size. These simulations with realistic plasma parameters find that the electron heat conductivity is much smaller than the experimental value and in contrast with recent findings of flux-tube simulations that ETG turbulence is responsible for the anomalous electron thermal transport in fusion plasmas. The nonlinear toroidal couplings represent a new paradigm for the spectral cascade in plasma turbulence. (author)

  12. Magnetic analysis of tokamak plasma with approximate MHD equilibrium solution

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Hiraki, Naoji

    1993-01-01

    A magnetic analysis method for determining equilibrium configuration parameters (plasma shape, poloidal beta and internal inductance) on a non-circular tokamak is described. The feature is to utilize an approximate MHD equilibrium solution which explicitly relates the configuration parameters with the magnetic fields picked up by magnetic sensors. So this method is suitable for the real-time analysis performed during a tokamak discharge. A least-squares fitting procedure is added to the analytical algorithm in order to reduce the errors in the magnetic analysis. The validity is investigated through the numerical calculation for a tokamak equilibrium model. (author)

  13. Transport simulations of a density limit in radiation-dominated tokamak discharges: profile effects

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1988-01-01

    The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with a magnetohydrodynamic (MHD) equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equaling the input power. The present work is confined to Ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result

  14. Transport simulations of a density limit in radiation-dominated tokamak discharges: Profile effects

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1988-06-01

    The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with an MHD equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equalling the input power. The present work is confined to ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result. 41 refs., 9 figs

  15. He++ transport in the PDX tokamak

    International Nuclear Information System (INIS)

    Fonck, R.J.; Hulse, R.A.

    1983-12-01

    A powerful new approach to the study of particle transport and helium ash confinement in high-temperature fusion plasmas is demonstrated by charge-exchange recombination spectroscopy of He ++ in ohmically heated PDX discharges. Time and space resolved measurements of He ++ density following a short puff of helium gas into the plasma edge are fitted using a diffusive/convective transport model with coefficients D = (2.1 +- 0.9) x 10 4 cm 2 s -1 and v(r)/D = (0.8 +- 0.3) delta (ln n/sub e/)/deltar

  16. Behaviour of metallic droplets in a tokamak plasma

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

    1989-01-01

    Micrometre sized tantalum droplets were injected into a tokamak plasma by a controllable arcing gun located behind the wall. The trajectories of the ablating particles were photographed by a high speed camera. Various possible mechanisms which may explain the observed curvature of the particle paths are discussed. The migration of the ablated material in the tokamak was studied by post-mortem analysis of collector probes and limiters. (author). Letter-to-the-editor. 12 refs, 9 figs

  17. A Midsize Tokamak As Fast Track To Burning Plasmas

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2010-01-01

    This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain ((ge) 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.

  18. Scaling for scrape-off layer plasma in tokamak

    International Nuclear Information System (INIS)

    Shimomura, Yasuo; Maeda, Hikosuke; Kimura, Haruyuki; Azumi, Masashi; Odajima, Kazuo

    1977-12-01

    Scaling for a scrape-off layer plasma in a tokamak is obtained by using DIVA (JFT-2a). The scaling gives the average electron temperature, the width and the mean electron density of the scrape-off layer. The temperature at the edge will be high in a future large tokamak with a small energy-loss by charge-exchange and radiation. The scrape-off layer plasma can easily shield the impurity influx from the wall. The fuel, however, can easily penetrate into the main plasma. (auth.)

  19. Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

    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.

  20. Electron and impurity transport studies in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, D.

    2013-05-15

    In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities

  1. Electron and impurity transport studies in the TCV Tokamak

    International Nuclear Information System (INIS)

    Wagner, D.

    2013-05-01

    In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities

  2. Edge and coupled core/edge transport modelling in tokamaks

    International Nuclear Information System (INIS)

    Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

    1999-01-01

    Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal E x B drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB , the model predicts transitions to supersonic flow at the inboard midplane. 2D simulations show the importance of E x B flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

  3. Edge and coupled core-edge transport modelling in tokamaks

    International Nuclear Information System (INIS)

    Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

    2001-01-01

    Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal ExB drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB, the model predicts transitions to supersonic SOL flow at the inboard midplane. 2D simulations show the importance of ExB flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

  4. User's manual for the Oak Ridge Tokamak Transport Code

    International Nuclear Information System (INIS)

    Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

    1977-02-01

    A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche

  5. Structural Stability of Tokamak Equilibrium: Transport Barriers

    Energy Technology Data Exchange (ETDEWEB)

    Solano, E. R.

    2001-07-01

    A generalised theory of structural stability of differential equations is introduced and applied to the Grad-Shafranov equation. It is discussed how the formation and loss of transport barrier could be associated with the appearance/disappearance of equilibria. The equilibrium conjecture is presented: transport barriers are associated with locally diamagnetic regions in the plasma, and affected by the paramagnetism of the bootstrap current. (Author) 18 refs.

  6. Feedback control of plasma position in the HL-1 tokamak

    International Nuclear Information System (INIS)

    Yuan Baoshan; Jiao Boliang; Yang Kailing

    1991-01-01

    In the HL-1 tokamak with a thick copper shell, the control of plasma position is successfully performed by a feedback-feedforward system with dual mode regulator and the equilibrium field coils outside the shell. The plasma position can be controlled within ±2 mm in both vertical and horizontal directions under the condition that the iron core of transformer is not saturated

  7. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    RAE is the maximum runaway energy emitted during a burst period of tdur. HXR. There being no plasma control feedback system in the SINP tokamak, the dynamics of the plasma equilibrium is time-dependent and the column shift is now made by the discharge dynamics itself. We measured DRAE for the two discharges ...

  8. Plasma fluctuation measurements in tokamaks using beam-plasma interactions

    International Nuclear Information System (INIS)

    Fonck, R.J.; Duperrex, P.A.; Paul, S.F.

    1990-01-01

    High-frequency observations of light emitted from the interactions between plasma ions and injected neutral beam atoms allow the measurement of moderate-wavelength fluctuations in plasma and impurity ion densities. To detect turbulence in the local plasma ion density, the collisionally excited fluorescence from a neutral beam is measured either separately at several spatial points or with a multichannel imaging detector. Similarly, the role of impurity ion density fluctuations is measured using charge exchange recombination excited transitions emitted by the ion species of interest. This technique can access the relatively unexplored region of long-wavelength plasma turbulence with k perpendicular ρ i much-lt 1, and hence complements measurements from scattering experiments. Optimization of neutral beam geometry and optical sightlines can result in very good localization and resolution (Δx≤1 cm) in the hot plasma core region. The detectable fluctuation level is determined by photon statistics, atomic excitation processes, and beam stability, but can be as low as 0.2% in a 100 kHz bandwidth over the 0--1 MHz frequency range. The choices of beam species (e.g., H 0 , He 0 , etc.), observed transition (e.g., H α , L α , He I singlet or triplet transitions, C VI Δn=1, etc.) are dictated by experiment-specific factors such as optical access, flexibility of beam operation, plasma conditions, and detailed experimental goals. Initial tests on the PBX-M tokamak using the H α emissions from a heating neutral beam show low-frequency turbulence in the edge plasma region

  9. Turbulence in tokamak plasmas. Effect of a radial electric field shear; Turbulence dans les plasmas de tokamaks. Effet d`un cisaillement de champ electrique radial

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs.

  10. Study of edge turbulence in tokamak plasmas

    International Nuclear Information System (INIS)

    Sarazin, Y.

    1997-01-01

    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.)

  11. Modeling of noble gas injection into tokamak plasmas

    International Nuclear Information System (INIS)

    Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.

    2005-01-01

    Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)

  12. Formation of transport barriers in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Meyer, H; Field, A R; Akers, R J; Brickley, C; Conway, N J; Patel, A; Carolan, P G; Challis, C; Counsell, G F; Cunningham, G; Helander, P; Kirk, A; Lloyd, B; Maingi, R; Tournianski, M R; Walsh, M J

    2004-01-01

    In the Mega Ampere Spherical Tokamak (MAST) plasmas have been generated with internal (ITB) or edge (ETB) transport barriers. ITBs were achieved in both the electron and the ion energy channel. In the presence of an ITB in the ion energy channel, transport analysis shows that the ion thermal diffusivity, χ i , is reduced to almost neoclassical values while the ITB persists. The widely tested criteria for ITB formation ρ t * =ρ s αlnT/αR>ρ ITB * ∼0.014 (ρ s : Larmor radius at sound speed) obtained from dimensional analysis of JET discharges is easily exceeded on MAST. Even without the evidence of an ρ T * >0.014 often applies, showing that this criterion in its current form is not generally applicable. ETBs are most easily formed in MAST if in a double null divertor configuration the discharge is vertically balanced, so that both X-points are almost on the same flux surface (CDND), and if the plasma is refuelled from the high field side mid-plane. The H-mode threshold power, P thr = 0.5 MW, in connected double null diverted (CDND) is only about half of that in a similar disconnected discharge with the ion ∇ B drift towards the X-point on the last closed flux surface (LDND). P thr scales between lower double null diverted (LDND) and the single null diverted configuration with the plasma surface area on MAST

  13. Fluctuations and transport in fusion plasmas. Final report

    International Nuclear Information System (INIS)

    Gould, R.W.; Liewer, P.C.

    1995-01-01

    The energy confinement in tokamaks in thought to be limited by transport caused by plasma turbulence. Three dimensional plasma particle-in-cell (PIC) codes are used to model the turbulent transport in tokamaks to attempt to understand this phenomena so that tokamaks can be made more efficient. Presently, hundreds of hours of Cray time are used to model these experiments and much bigger and longer runs are desired, to model a large tokamak with realistic parameters is beyond the capability of existing sequential supercomputers. Parallel supercomputers might be a cost effect tool for performing such large scale 3D tokamak simulations. The goal of the work was to develop algorithms for performing PIC codes on coarse-grained message passing parallel computers and to evaluate the performance of such parallel computers on PIC codes. This algorithm would be used in a large scale PIC production code such as the UCLA 3D gyrokinetic code

  14. Poloidal plasma rotation in the presence of RF waves in tokamaks

    International Nuclear Information System (INIS)

    Weyssow, B.; Liu, Caigen

    2001-01-01

    It is well known that one of the consequences of strong RF heating is the deformation of the equilibrium distribution function that induces a change in plasma transport and plasma rotation. The poloidal plasma rotation during RF wave heating in tokamaks is investigated using a moment approach. A set of closed, self-consistent transport and rotation equations is derived and reduced to a single equation for the poloidal particle flux. The formulas are sufficiently general to apply to heating schemes that can be represented by a quasilinear operator. (author)

  15. Study of the electron heat transport in Tore-Supra tokamak; Etude du transport de la chaleur electronique dans le Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Harauchamps, E

    2004-07-01

    This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

  16. Fluctuations and transport in fusion plasmas. Annual progress report, October 1, 1983-September 30, 1984

    International Nuclear Information System (INIS)

    Gould, R.W.

    1984-01-01

    This grant supports an integrated program of experiment and theory in tokamak plasma physics. Emphasis is placed on microscopic fluctuations and anomalous transport. The primary objective is to characterize the properties of the microscopic fluctuations observed in tokamaks and to try to develop an understanding of the fluctuation-induced transport of particles and heat. Anomalous transport, which causes energy losses one to two orders of magnitude larger than predicted by neoclassical transport theory, occurs in all tokamaks and underlies empirical scaling laws

  17. Ion transport in turbulent edge plasmas

    International Nuclear Information System (INIS)

    Helander, P.; Massachusetts Inst. of Tech., Cambridge, MA; Hazeltine, R.D.; Catto, P.J.

    1996-02-01

    Edge plasmas, such as the tokamak scrape-off layer, exist as a consequence of a balance between cross-field diffusion and parallel losses. The former is usually anomalous, and is widely thought to be driven by strong electrostatic turbulence. It is shown that the anomalous diffusion affects the parallel ion transport by giving rise to a new type of thermal force between different ion species. This force is parallel to the magnetic field, but arises entirely because of perpendicular gradients, and could be important for impurity retention in the tokamak divertor. (author)

  18. Edge Plasma Physics and Relevant Diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Devynck, P.; Gunn, J.; Martines, E.; Bonhomme, G.; Van Oost, G.; Hron, Martin; Ďuran, Ivan; Pánek, Radomír; Stejskal, Pavel; Adámek, Jiří

    2004-01-01

    Roč. 3, - (2004), s. 1-6 ISSN 1433-5581. [First Cairo Conference on Plasma Physics & Applications. Cairo, 11.10.2003-15.10.2003] R&D Projects: GA ČR GA202/03/0786; GA ČR GP202/03/P062 Keywords : tokamak * edge plasma * probe diagnostics * biasing * turbulence * polarization Subject RIV: BL - Plasma and Gas Discharge Physics

  19. Kinetic modelling of runaway electron avalanches in tokamak plasmas.

    Czech Academy of Sciences Publication Activity Database

    Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R.S.; Saint-Laurent, F.; Vlainic, Milos

    2015-01-01

    Roč. 57, č. 9 (2015), č. článku 095006. ISSN 0741-3335 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : plasma physics * runaway electrons * knock-on collisions * tokamak * Fokker-Planck * runaway avalanches Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.404, year: 2015

  20. Advanced probes for edge plasma diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Adámek, Jiří; Balan, P.; Hronová-Bilyková, Olena; Brotánková, Jana; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Horáček, Jan; Ionita, C.; Kocan, M.; Martines, E.; Pánek, Radomír; Peleman, P.; Schrittwieser, R.; Van Oost, G.; Žáček, František

    2006-01-01

    Roč. 63, č. 0 (2006), 012001-012002 E-ISSN 1742-6596. [SECOND INTERNATIONAL WORKSHOP AND SUMMER SCHOOL ON PLASMA PHYSICS. Kiten, 03.07.2006-09.07.2006] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak * electric probes * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  1. Classical dissipation and transport in plasmas

    International Nuclear Information System (INIS)

    Hinton, F.L.

    1989-01-01

    This paper reviews the subject of classical and neoclassical transport. The paper is organized into four main parts, dealing with plasma kinetic theory, classical transport, neoclassical transport, and the present state of the subject. The results of the neoclassical theory of transport are still being used to give the lower limit on the transport rates in tokamaks, which would apply if instabilities and turbulence could be suppressed. So far, only the ion thermal conductivity and the current density have been found experimentally to agree with this theory, and only under special conditions. The electron thermal conductivity has been found experimentally to be much larger than the neoclassical prediction

  2. Plasma physics

    International Nuclear Information System (INIS)

    1979-01-01

    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

  3. Effects of fueling profiles on plasma transport

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Mense, A.T.; Attenberger, S.E.; Milora, S.L.

    1977-01-01

    The effects of cold particle fueling profiles on particle and energy transport in an ignition sized tokamak plasma are investigated in this study with a one-dimensional, multifluid transport model. A density gradient driven trapped particle microinstability model for plasma transport is used to demonstrate potential effects of fueling profiles on ignition requirements. Important criteria for the development of improved transport models under the conditions of shallow particle fueling profiles are outlined. A discrete pellet fueling model indicates that large fluctuations in density and temperature may occur in the outer regions of the plasma with large, shallowly penetrating pellets, but fluctuations in the pressure profile are small. The hot central core of the plasma remains unaffected by the large fluctuations near the plasma edge

  4. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Energy Technology Data Exchange (ETDEWEB)

    Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)

    2015-10-15

    Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.

  5. Investigation of small-scale tokamak plasma turbulence by correlative UHR backscattering diagnostics

    International Nuclear Information System (INIS)

    Gusakov, E Z; Gurchenko, A D; Altukhov, A B; Bulanin, V V; Esipov, L A; Kantor, M Yu; Kouprienko, D V; Lashkul, S I; Petrov, A V; Stepanov, A Yu

    2006-01-01

    Fine scale turbulence is considered nowadays as a possible candidate for the explanation of anomalous ion and electron energy transport in magnetized fusion plasmas. The unique correlative upper hybrid resonance backscattering (UHR BS) technique is applied at the FT-2 tokamak for investigation of density fluctuations excited in this turbulence. The measurements are carried out in Ohmic discharge at several values of plasma current and density and during current ramp up experiment. The moveable focusing antennas set have been used in experiments allowing probing out of equatorial plane. The radial wave number spectra of the small-scale component of tokamak turbulence are determined from the correlation data with high spatial resolution. Two small-scale modes possessing substantially different phase velocities are observed in plasma under conditions when the threshold for the electron temperature gradient mode excitation is overcome. The possibility of plasma poloidal velocity profile determination using the UHR BS signal is demonstrated

  6. Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1997-01-01

    Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated

  7. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1986-10-01

    It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism. A model is developed which describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/ ∞ n -1 /sub b/T 3 /sub e/(α 0 /L) 2 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  8. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1987-01-01

    It has been seen that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. This trapping of a gun-injected plasma is explained in terms of a depolarization current mechanism. A model is developed that describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/proportionalT/sup 3/2//sub e/L 2 /n/sub b/α 2 0 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  9. Development of a positronium atom diagnostic beam to study transport in tokamaks

    International Nuclear Information System (INIS)

    Surko, C.M.

    1992-01-01

    Anomalous transport is probably the key physics issue in magnetic confinement fusion. It severely limits our ability to create and maintain a dense, hot, fusion plasma. There is also good evidence that there is a fundamental lack of understanding in this area, namely the transport of particles and energy induced by waves and fluctuations in magnetically confined plasmas. The positronium atom beam diagnostic, if successful, can provide a qualitatively new and different way of studying transport in tokamak fusion plasmas. The use of the positron as a thermalized, electron-mass test particle will allow important new tests of current theories of plasma transport. In particular, it could provide unique insights into the potential role of magnetic fluctuations in producing anomalous transport. This is particularly significant in that, at present, there is essentially no in situ probe of these fluctuations or the transport which they produce. Some results of this study are summarized

  10. Issues in tokamak/stellarator transport and confinement enhancement mechanisms

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1990-08-01

    At present, the mechanism for anomalous energy transport in low-β toroidal plasmas -- tokamaks and stellarators -- remains unclear, although transport by turbulent E x B velocities associated with nonlinear, fine-scale microinstabilities is a leading candidate. This article discusses basic theoretical concepts of various transport and confinement enhancement mechanisms as well as experimental ramifications which would enable one to distinguish among them and hence identify a dominant transport mechanism. While many of the predictions of fine-scale turbulence are born out by experiment, notable contradictions exist. Projections of ignition margin rest both on the scaling properties of the confinement mechanism and on the criteria for entering enhanced confinement regimes. At present, the greatest uncertainties lie with the basis for scaling confinement enhancement criteria. A series of questions, to be answered by new experimental/theoretical work, is posed to resolve these outstanding contradictions (or refute the fine-scale turbulence model) and to establish confinement enhancement criteria. 73 refs., 4 figs., 5 tabs

  11. One-dimensional fluid model for transport in divertor and limiter tokamak scrape-off layers

    International Nuclear Information System (INIS)

    Lipschultz, B.

    1983-11-01

    Single-fluid transport in the plasma scrape-off layer is modeled for poloidal divertor and mechanically limited discharges. This numerical model is one-dimensional along a field line and time-independent. Conductive and convective transport, as well as impurity and neutral source (sink) terms are included. A simple shooting method technique is used for obtaining solutions. Results are shown for the case of the proposed Alcator DCT tokamak

  12. Rippling modes in the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Carreras, B.A.; Callen, J.D.; Gaffney, P.W.; Hicks, H.R.

    1982-02-01

    A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper we develop a computational model for these modes in the cylindrical tokamak approximation and explore the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor scaling as k/sub parallel//sup -4/3/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m/sup -1/, and the radial extent of these modes grows linearly with time due to radial Vector E x Vector B 0 convection. This evolution is found to be terminated by parallel heat conduction

  13. Rippling modes in the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Carreras, B.A.; Gaffney, P.W.; Hicks, H.R.; Callan, J.D.

    1982-01-01

    A promising resistive magnetohydrodynamic candidate for the underlying cause of turbulence in the edge of a tokamak plasma is the rippling instability. In this paper a computational model for these modes in the cylindrical tokamak approximation was developed and the linear growth and single-helicity quasi-linear saturation phases of the rippling modes for parameters appropriate to the edge of a tokamak plasma were explored. Large parallel heat conduction does not stabilize these modes; it only reduces their growth rate by a factor sacling as K/sup -4/3//sub parallel/. Nonlinearly, individual rippling modes are found to saturate by quasi-linear flattening of the resistivity profile. The saturated amplitude of the modes scales as m -1 , and the radial extent of these modes grows linearly with time due to radial E x B 0 convection. This evolution is found to be terminated by parallel heat conduction

  14. Studies on fundamental technologies for producing tokamak-plasma

    International Nuclear Information System (INIS)

    Matsuzaki, Yoshimi

    1987-10-01

    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/cm 3 produced plasma of an electron density of 5 x 10 11 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)

  15. Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

    International Nuclear Information System (INIS)

    Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.

    2017-01-01

    Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

  16. Equilibrium of rotating and nonrotating plasmas in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2003-01-01

    One studied plasma equilibrium in tokamak in case of toroidal rotation. Rotation associated centrifugal force is shown to result in decrease of equilibrium limit as to β. One analyzes unlike opinion and considers its supports. It is shown that in possible case of local improvement of equilibrium conditions associated with special selection of profile of plasma rotation rate, the combined integral effect turns to be negative one. But in case of typical conditions, decrease of equilibrium β caused by plasma rotation is negligible one and one may ignore effect of plasma rotation on its equilibrium for hot plasma [ru

  17. A dimensionless criterion for characterising internal transport barriers in tokamaks

    International Nuclear Information System (INIS)

    Tresset, G.; Litaudon, X.; Moreau, D.

    2000-07-01

    A simple criterion, based on a dimensionless parameter (ρ T * =ρ s /L T ) related to drift wave turbulence stabilisation and anomalous transport theory, is proposed in order to characterise the emergence and the space-time evolution of internal transport barriers (ITB's) during a tokamak discharge. The underlying physics which led us to consider the possible relevance of this parameter as a local indicator of a bifurcated plasma state is the breaking of the gyro-Bohm turbulence scaling by the diamagnetic velocity shear, which has been observed in various numerical simulations [e.g. X. Garbet and R.E. Waltz, Phys. Plasmas 3(1996) 1898]. The presence of an ITB is inferred when ρ T * exceeds a threshold value. The main features like the emergence time, location and even dynamics of ITB's can then be summarized on a single graphical representation consistent with measurement uncertainties. The validity of such a criterion is demonstrated on the Optimized Shear (OS) database of JET in several experimental configurations. Large database analysis and realtime control of OS discharges are envisaged as the most attractive applications. (author)

  18. Experimental methods to study tokamak plasma stability

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1978-01-01

    Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)

  19. Anomalous periodic disruptions in tokamak plasma

    International Nuclear Information System (INIS)

    Montvai, A.; Tegze, M.; Valyi, I.

    1982-09-01

    Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)

  20. On steady poloidal and toroidal flows in tokamak plasmas

    International Nuclear Information System (INIS)

    McClements, K. G.; Hole, M. J.

    2010-01-01

    The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

  1. Detection of tokamak plasma positrons using annihilation photons

    Energy Technology Data Exchange (ETDEWEB)

    Guanying, Yu; Liu, Jian; Xie, Jinlin [University of Science and Technology, Hefei, Anhui, 230027 (China); Li, Jiangang, E-mail: j_li@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2017-05-15

    Highlights: • A design for detection of tokamak plasma positrons is given. • Identify the main obstacle toward experimental confirmation of fusion plasma positrons. • Signal to noise ratio in a plasma disruption is estimated. • Unique potential applications of fusion plasma positrons are discussed. - Abstract: A massive amount of positrons (plasma positrons), produced by the collision between runaway electrons and nuclei during fusion plasma disruption, was first predicted theoretically in 2003. To help confirm this prediction, we report here the design of an experimental system to detect tokamak plasma positrons. Because a substantial amount of positrons (material positrons) are produced when runaway electrons impact plasma-facing materials, we proposed maximizing the ratio of plasma to material positrons by inserting a thin carbon target at the plasma edge as a plasma positron bombing target and producing a plasma disruption scenario triggered by massive gas injection. Meanwhile, the coincidence detection of positron annihilation photons was used to filter out the noise of annihilation photons from locations other than the carbon target and that of bremsstrahlung photons near 511 keV. According to our simulation, the overall signal-to-noise ratio should be more than 10:1.

  2. Development in Diagnostics Application to Control Advanced Tokamak Plasma

    International Nuclear Information System (INIS)

    Koide, Y.

    2008-01-01

    For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

  3. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  4. Fast Low-to-High Confinement Mode Bifurcation Dynamics in a Tokamak Edge Plasma Gyrokinetic Simulation.

    Science.gov (United States)

    Chang, C S; Ku, S; Tynan, G R; Hager, R; Churchill, R M; Cziegler, I; Greenwald, M; Hubbard, A E; Hughes, J W

    2017-04-28

    Transport barrier formation and its relation to sheared flows in fluids and plasmas are of fundamental interest in various natural and laboratory observations and of critical importance in achieving an economical energy production in a magnetic fusion device. Here we report the first observation of an edge transport barrier formation event in an electrostatic gyrokinetic simulation carried out in a realistic diverted tokamak edge geometry under strong forcing by a high rate of heat deposition. The results show that turbulent Reynolds-stress-driven sheared E×B flows act in concert with neoclassical orbit loss to quench turbulent transport and form a transport barrier just inside the last closed magnetic flux surface.

  5. Hybrid model for simulation of plasma jet injection in tokamak

    Science.gov (United States)

    Galkin, Sergei A.; Bogatu, I. N.

    2016-10-01

    Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

  6. Kinetic modelling of runaway electron avalanches in tokamak plasmas

    International Nuclear Information System (INIS)

    Nilsson, E; Peysson, Y; Saint-Laurent, F; Decker, J; Granetz, R S; Vlainic, M

    2015-01-01

    Runaway electrons can be generated in tokamak plasmas if the accelerating force from the toroidal electric field exceeds the collisional drag force owing to Coulomb collisions with the background plasma. In ITER, disruptions are expected to generate runaway electrons mainly through knock-on collisions (Hender et al 2007 Nucl. Fusion 47 S128–202), where enough momentum can be transferred from existing runaways to slow electrons to transport the latter beyond a critical momentum, setting off an avalanche of runaway electrons. Since knock-on runaways are usually scattered off with a significant perpendicular component of the momentum with respect to the local magnetic field direction, these particles are highly magnetized. Consequently, the momentum dynamics require a full 3D kinetic description, since these electrons are highly sensitive to the magnetic non-uniformity of a toroidal configuration. For this purpose, a bounce-averaged knock-on source term is derived. The generation of runaway electrons from the combined effect of Dreicer mechanism and knock-on collision process is studied with the code LUKE, a solver of the 3D linearized bounce-averaged relativistic electron Fokker–Planck equation (Decker and Peysson 2004 DKE: a fast numerical solver for the 3D drift kinetic equation Report EUR-CEA-FC-1736, Euratom-CEA), through the calculation of the response of the electron distribution function to a constant parallel electric field. The model, which has been successfully benchmarked against the standard Dreicer runaway theory now describes the runaway generation by knock-on collisions as proposed by Rosenbluth (Rosenbluth and Putvinski 1997 Nucl. Fusion 37 1355–62). This paper shows that the avalanche effect can be important even in non-disruptive scenarios. Runaway formation through knock-on collisions is found to be strongly reduced when taking place off the magnetic axis, since trapped electrons can not contribute to the runaway electron population. Finally

  7. Impurity transport studies on the FTU tokamak

    International Nuclear Information System (INIS)

    Pacella, D.; Romanelli, F.; Gregory, B.

    1999-01-01

    In this work, the radial profile of the diffusion coefficient D and the convective velocity V in the plasma core (0 2 /s and V ∼ 100 m/s. A model for the anomalous transport induced by electrostatic turbulence is developed. With a typical fluctuation spectrum (ω = 10 5 -2x10 5 Hz), calculations can reproduce very well the experimental results. To investigate the impurity behavior in a non-stationary phase, Kr gas was injected into the plasma. It is found that the total flux of Kr gas flowing into the core is also driven by diffusion but the magnitude is much lower than the single ion fluxes derived for Mo ions. The effect of the turbulence on the single ion is very strong but it is reduced when averaged over many charge states. (author)

  8. Neural net prediction of tokamak plasma disruptions

    International Nuclear Information System (INIS)

    Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.

    1994-10-01

    The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system

  9. Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks

    International Nuclear Information System (INIS)

    Park, Jong-kyu; Boozer, Allen H.; Menard, Jonathan E.; Garofalo, Andrea M.; Schaffer, Michael J.; Hawryluk, Richard J.; Kaye, Stanley M.; Gerhardt, Stefan P.; Sabbagh, Steve A. and the NSTX Team

    2009-01-01

    Tokamaks are sensitive to deviations from axisymmetry as small as (delta)B/B 0 ∼ 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 equivalently 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 sufficiently 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

  10. PELLET: a computer routine for modeling pellet fueling in tokamak plasmas

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Iskra, M.A.; Howe, H.C.; Attenberger, S.E.

    1979-01-01

    Recent experimental results of frozen hydrogenic pellet injection into hot tokamak plasmas and substantial agreement with theoretical predictions have led to a much greater interest in pellets as a means of refueling plasmas. The computer routine PELLET has been developed and used as an aid in assessing pellet ablation models and the effects of pellets on plasma behavior. PELLET provides particle source profiles under various options for the ablation model and can be coupled either to a fluid transport code or to a brief routine which supplies the required input parameters

  11. Anomalous transport in toroidal plasmas

    International Nuclear Information System (INIS)

    Punjabi, A.

    1991-01-01

    We have developed a Monte Carlo method to estimate the transport of different groups of particles for plasmas in toroidal geometries. This method can determine the important transport mechanisms driving the anomalous transport by comparing the numerical results with the experimental data. The important groups of particles whose transport can be estimated by this method include runaway electrons, thermal electrons, both passing and trapped diagnostic beam ions etc. The three basic mechanisms driving the anomalous transport are: spatial variation of magnetic field strength, spatial variation of electrostatic potential within the flux surfaces, and the loss of flux surfaces. The equation of motion are obtained from the drift hamiltonian. The equations of motion are developed in the canonical and in the non-canonical, practical co-ordinates as well. The effects of collisions are represented by appropriate stochastic changes in the constants of motion at each time-step. Here we present the results of application of this method to three cases: superathermal alphas in the rippled field of tokamaks, motion in the magnetic turbulence of takapole II, and transport in the stochastic fields of ZT40. This work is supported by DOE OFE and ORAU HBCU program

  12. Gyrokinetic Simulation of Global Turbulent Transport Properties in Tokamak Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Wang, W.X.; Lin, Z.; Tang, W.M.; Lee, W.W.; Ethier, S.; Lewandowski, J.L.V.; Rewoldt, G.; Hahm, T.S.; Manickam, J.

    2006-01-01

    A general geometry gyro-kinetic model for particle simulation of plasma turbulence in tokamak experiments is described. It incorporates the comprehensive influence of noncircular cross section, realistic plasma profiles, plasma rotation, neoclassical (equilibrium) electric fields, and Coulomb collisions. An interesting result of global turbulence development in a shaped tokamak plasma is presented with regard to nonlinear turbulence spreading into the linearly stable region. The mutual interaction between turbulence and zonal flows in collisionless plasmas is studied with a focus on identifying possible nonlinear saturation mechanisms for zonal flows. A bursting temporal behavior with a period longer than the geodesic acoustic oscillation period is observed even in a collisionless system. Our simulation results suggest that the zonal flows can drive turbulence. However, this process is too weak to be an effective zonal flow saturation mechanism.

  13. ELMy-H mode as limit cycle and chaotic oscillations in tokamak plasmas

    International Nuclear Information System (INIS)

    Itoh Sanae, I.; Itoh, Kimitaka; Fukuyama, Atsushi; Miura, Yukitoshi.

    1991-05-01

    A model of Edge Localized Modes (ELMs) in tokamak plasmas is presented. A limit cycle solution is found in the transport equation (time-dependent Ginzburg-Landau type), which a has hysteresis curve between the gradient and flux. Periodic oscillation of the particle outflux and L/H intermediate state are predicted near the L/H transition boundary. A mesophase in spatial structure appears near edge. Chaotic oscillation is also predicted. (author)

  14. Real-time control of current and pressure profiles in tokamak plasmas

    International Nuclear Information System (INIS)

    Laborde, L.

    2005-12-01

    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)

  15. A model for plasma discharges simulation in Tokamak devices

    International Nuclear Information System (INIS)

    Fonseca, Antonio M.M.; Silva, Ruy P. da; Galvao, Ricardo M.O.; Kusnetzov, Yuri; Nascimento, I.C.; Cuevas, Nelson

    2001-01-01

    In this work, a 'zero-dimensional' model for simulation of discharges in Tokamak machine is presented. The model allows the calculation of the time profiles of important parameters of the discharge. The model was applied to the TCABR Tokamak to study the influence of parameters and physical processes during the discharges. Basically it is constituted of five differential equations: two related to the primary and secondary circuits of the ohmic heating transformer and the other three conservation equations of energy, charge and neutral particles. From the physical model, a computer program has been built with the objective of obtaining the time profiles of plasma current, the current in the primary of the ohmic heating transformer, the electronic temperature, the electronic density and the neutral particle density. It was also possible, with the model, to simulate the effects of gas puffing during the shot. The results of the simulation were compared with the experimental results obtained in the TCABR Tokamak, using hydrogen gas

  16. Analysis of tokamak plasma confinement modes using the fast ...

    Indian Academy of Sciences (India)

    2016-10-20

    Oct 20, 2016 ... ruby laser technology have been extended to deter- mine q. The edge radial electric field is known to ... The control of the shear layer is therefore important to modify transport in tokamaks [26–33]. A ... The calculation of PSD is directly accomplished by the FFT technique presented in the MATLAB software.

  17. Quasilinear kinetic modeling of RMP penetration into a tokamak plasma

    International Nuclear Information System (INIS)

    Heyn, M.F.; Kernbichler, W.; Leitner, P.; Ivanov, I.B.; Kasilov, S.V.

    2013-01-01

    The linear as well as the quasilinear problem of RMP penetration in tokamaks is solved consistently with a particle and energy conserving collision operator. The new collision operator ensures the Onsager symmetry of the quasilinear transport coefficient matrix and avoids artifacts such as fake heat convection connected with simplified collision models.

  18. Turbulence in tokamak plasmas. Effect of a radial electric field shear

    International Nuclear Information System (INIS)

    Payan, J.

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs

  19. Formation of transport barriers in the MAST spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, H [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Field, A R [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Akers, R J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Brickley, C [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Conway, N J [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Patel, A [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Carolan, P G [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Challis, C [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Counsell, G F [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Cunningham, G [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Helander, P [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Kirk, A [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Lloyd, B [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Maingi, R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Tournianski, M R [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Walsh, M J [Walsh Scientific Ltd, Culham Science Centre, Abingdon, Oxfordshire, OX14 3EB (United Kingdom)

    2004-05-01

    In the Mega Ampere Spherical Tokamak (MAST) plasmas have been generated with internal (ITB) or edge (ETB) transport barriers. ITBs were achieved in both the electron and the ion energy channel. In the presence of an ITB in the ion energy channel, transport analysis shows that the ion thermal diffusivity, {chi}{sub i}, is reduced to almost neoclassical values while the ITB persists. The widely tested criteria for ITB formation {rho}{sub t}{sup *}={rho}{sub s}{alpha}lnT/{alpha}R>{rho}{sub ITB}{sup *}{approx}0.014 ({rho}{sub s}: Larmor radius at sound speed) obtained from dimensional analysis of JET discharges is easily exceeded on MAST. Even without the evidence of an {rho}{sub T}{sup *}>0.014 often applies, showing that this criterion in its current form is not generally applicable. ETBs are most easily formed in MAST if in a double null divertor configuration the discharge is vertically balanced, so that both X-points are almost on the same flux surface (CDND), and if the plasma is refuelled from the high field side mid-plane. The H-mode threshold power, P{sub thr} = 0.5 MW, in connected double null diverted (CDND) is only about half of that in a similar disconnected discharge with the ion {nabla} B drift towards the X-point on the last closed flux surface (LDND). P{sub thr} scales between lower double null diverted (LDND) and the single null diverted configuration with the plasma surface area on MAST.

  20. Plasma shaping effects on tokamak scrape-off layer turbulence

    Science.gov (United States)

    Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo

    2017-03-01

    The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\

  1. Pseudo-MHD ballooning modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Callen, J.D.; Hegna, C.C.

    1996-08-01

    The MHD description of a plasma is extended to allow electrons to have both fluid-like and adiabatic-regime responses within an instability eigenmode. In the resultant open-quotes pseudo-MHDclose quotes model, magnetic field line bending is reduced in the adiabatic electron regime. This makes possible a new class of ballooning-type, long parallel extent, MHD-like instabilities in tokamak plasmas for α > s 2 (2 7/3 /9) (r p /R 0 ) or-d√Β/dr > (2 1/6 /3)(s/ R 0q ), which is well below the ideal-MHD stability boundary. The marginally stable pressure profile is similar in both magnitude and shape to that observed in ohmically heated tokamak plasmas

  2. Neoclassical transport coefficients for tokamaks with bean-shaped flux surfaces

    International Nuclear Information System (INIS)

    Chang, C.S.; Kaye, S.M.

    1990-11-01

    Simple analytic representations of the neoclassical transport coefficients for indented flux surfaces are presented. It is shown that a transport coefficient for an indented flux surface can be expressed in terms of a linear combination of the previously known transport coefficients for two nonindented flux surfaces. Numerical calculations based on actual equilibria from the PBX-M tokamak indicate that, even for modestly indented flux surfaces, the ion neoclassical thermal transport can be over a factor of two smaller than in a circular plasma with the same midplane radius or with the equivalent areas. 6 refs., 5 figs., 1 tab

  3. X transport and its effect on H-mode and edge pedestal in tokamaks

    International Nuclear Information System (INIS)

    Chang, C.S.; Darrow, D.; White, R.; Lin, Z.; Lee, W.; Ku, S.H.; Weitzner, H.; Carlstrom, T.N.; Grassie, J.S. de

    2001-01-01

    A new classical non-ambipolar transport mechanism has been identified which can be a dominant source of strong Er and edge pedestal layer formation immediately inside the separatrix in a diverted tokamak. Due to vanishingly small poloidal B-field and grad-B drift toward x-point, plasma ions with small ν parallel in the X-region do not have confined single particle orbits. This leads to a non-ambipolar convective transport in the X-region (X-transport), either collisional or collisionless, inducing a strong negative Er-shear layer. The X-transport can provide basic understanding of many of the experimental observations. (author)

  4. Understanding and Control of Transport in Advanced Tokamak Regimes in DIII-D

    International Nuclear Information System (INIS)

    C.M. Greenfield; J.C. DeBoo; T.C. Luce; B.W. Stallard; E.J. Synakowski; L.R. Baylor; K.H. Burrell; T.A. Casper; E.J. Doyle; D.R. Ernst; J.R. Ferron; P. Gohil; R.J. Groebner; L.L. Lao; M. Makowski; G.R. McKee; M. Murakami; C.C. Petty; R.I. Pinsker; P.A. Politzer; R. Prater; C.L. Rettig; T.L. Rhodes; B.W. Rice; G.L. Schmidt; G.M. Staebler; E.J. Strait; D.M. Thomas; M.R. Wade

    1999-01-01

    Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection takes place when the heating power exceeds a threshold value of about 9 MW, contrasting to CO-NBI injection, where P threshold N H 89 = 9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency MHD activity is identified as the probable cause

  5. Automation of Aditya tokamak plasma position control DC power supply

    Energy Technology Data Exchange (ETDEWEB)

    Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

    2016-11-15

    Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

  6. Metal droplet erosion and shielding plasma layer under plasma flows typical of transient processes in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru [National Research Nuclear University “MEPhI” (Russian Federation)

    2017-03-15

    It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.

  7. Plasma Sprayed Tungsten-based Coatings and their Usage in Edge Plasma Region of Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Dufková, Edita; Piffl, Vojtěch; Peřina, Vratislav

    2006-01-01

    Roč. 51, č. 2 (2006), s. 179-191 ISSN 0001-7043 Grant - others:Evropská unie EFDA Task TW-5-TVM-PSW (EU – Euratom) Institutional research plan: CEZ:AV0Z20430508; CEZ:AV0Z10480505 Keywords : plasma sprayed coatings * fusion * plasma facing components * tungsten * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

  8. ADVANCES IN COMPREHENSIVE GYROKINETIC SIMULATIONS OF TRANSPORT IN TOKAMAKS

    International Nuclear Information System (INIS)

    WALTZ, RE; CANDY, J; HINTON, FL; ESTRADA-MILA, C; KINSEY, JE.

    2004-01-01

    A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite β, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ * ) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or a globally with physical profile variation. Rohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, plasma pinches and impurity flow, and simulations at fixed flow rather than fixed gradient are illustrated and discussed

  9. 'Snowflake' H Mode in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.

    2010-01-01

    An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy (ΔW ELM /W p ) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.

  10. Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

    International Nuclear Information System (INIS)

    Xu, X Q

    2007-01-01

    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 (ψ, θ, ε, μ) 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

  11. On impurities transport in a tokamak

    International Nuclear Information System (INIS)

    Rozhanskij, V.A.

    1980-01-01

    Transport of impurity ions is analitically analized in the case when main plasma is in plateau or banana regimes but impurity ions - in the Pfirsch-Schlutter mode. It is shown that in the large region of parameters the impUrity transport represents a drift in a p oloidal electric field, averaged from magnetic surface with provision for disturbance of concentration on it. Therefore, transport velocity does not depend on Z value and impurity type, as well as collision frequency both in the plateau and banana regimes. A value of flows is determined by the value of poloidal rotation velocity. At the rotation velocity corresponding to the electric field directed from the centre to periphery impurities are thrown out of a discharge, in the reverse case the flow is directed inside. Refusal from the assumption that Zsub(eff) > approximately 2, does not considerably change the results of work. The approach developed in the process of work can be applied to the case when impurity ions are in the plateau or banana modes

  12. Plasma radiation in tokamak disruption simulation experiments

    International Nuclear Information System (INIS)

    Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A.; Wuerz, H.

    1995-01-01

    Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10 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

  13. Advanced transport modeling of toroidal plasmas with transport barriers

    International Nuclear Information System (INIS)

    Fukuyama, A.; Murakami, S.; Honda, M.; Izumi, Y.; Yagi, M.; Nakajima, N.; Nakamura, Y.; Ozeki, T.

    2005-01-01

    Transport modeling of toroidal plasmas is one of the most important issue to predict time evolution of burning plasmas and to develop control schemes in reactor plasmas. In order to describe the plasma rotation and rapid transition self-consistently, we have developed an advanced scheme of transport modeling based on dynamical transport equation and applied it to the analysis of transport barrier formation. First we propose a new transport model and examine its behavior by the use of conventional diffusive transport equation. This model includes the electrostatic toroidal ITG mode and the electromagnetic ballooning mode and successfully describes the formation of internal transport barriers. Then the dynamical transport equation is introduced to describe the plasma rotation and the radial electric field self-consistently. The formation of edge transport barriers is systematically studied and compared with experimental observations. The possibility of kinetic transport modeling in velocity space is also examined. Finally the modular structure of integrated modeling code for tokamaks and helical systems is discussed. (author)

  14. Study of the electron heat transport in Tore-Supra tokamak

    International Nuclear Information System (INIS)

    Harauchamps, E.

    2004-01-01

    This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

  15. Internal transport barrier formation and pellet injection simulation in helical and tokamak reactors

    International Nuclear Information System (INIS)

    Higashiyama, You; Yamazaki, Kozo; Arimoto, Hideki; Garcia, Jeronimo

    2008-01-01

    In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E x B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors. (author)

  16. Tokamak plasma shape identification based on the boundary integral equations

    International Nuclear Information System (INIS)

    Kurihara, Kenichi; Kimura, Toyoaki

    1992-05-01

    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)

  17. Initial plasma production by induction electric field on QUEST tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

    Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)

  18. 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.

  19. Wave trajectory and electron cyclotron heating in tokamak plasmas

    International Nuclear Information System (INIS)

    Tanaka, S.; Maekawa, T.; Terumichi, Y.; Hamada, Y.

    1980-01-01

    Wave trajectories in high density tokamak plasmas are studied numerically. Results show that the ordinary wave injected at an appropriate incident angle can propagate into the dense plasmas and is mode-converted to the extraordinary wave at the plasma cutoff, is further converted to the electron Bernstein wave during passing a loop or a folded curve near the upper hybrid resonance layer, and is cyclotron damped away, resulting in local electron heating before arriving at the cyclotron resonance layer. Similar trajectory and damping are obtained when a microwave in a form of extraordinary wave is injected quasi-perpendicularly in the direction of decreasing toroidal field

  20. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-01-01

    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)

  1. Core transport properties in JT-60U and JET identity plasmas

    NARCIS (Netherlands)

    Litaudon, X.; Sakamoto, Y.; de Vries, P. C.; Salmi, A.; Tala, T.; Angioni, C.; Benkadda, S.; Beurskens, M. N. A.; Bourdelle, C.; Brix, M.; Crombe, K.; Fujita, T.; Futatani, S.; Garbet, X.; Giroud, C.; Hawkes, N. C.; Hayashi, N.; Hoang, G. T.; Hogeweij, G. M. D.; Matsunaga, G.; Nakano, T.; Oyama, N.; Parail, V.; Shinohara, K.; Suzuki, T.; Takechi, M.; Takenaga, H.; Takizuka, T.; Urano, H.; Voitsekhovitch, I.; Yoshida, M.

    2011-01-01

    The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma

  2. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    International Nuclear Information System (INIS)

    Hively, L.M.

    1980-01-01

    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

  3. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    Science.gov (United States)

    Hughes, Paul E.

    It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

  4. Anisotropic plasma with flows in tokamak: Steady state and stability

    International Nuclear Information System (INIS)

    Ilgisonis, V.I.

    1996-01-01

    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

  5. Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

    International Nuclear Information System (INIS)

    Malik, M.A.

    1988-01-01

    There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism

  6. One dimensional simulation on stability of detached plasma in a tokamak divertor

    International Nuclear Information System (INIS)

    Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi

    1999-06-01

    The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is ∼2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)

  7. Scrape-off layer plasma modeling for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Porter, G.D.; Rognlien, T.D.; Allen, S.L.

    1994-09-01

    The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing

  8. MTX [Microwave Tokamak Experiment] plasma diagnostic system

    International Nuclear Information System (INIS)

    Rice, B.W.; Hooper, E.B.; Brooksby, C.A.

    1987-01-01

    In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs

  9. Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

    International Nuclear Information System (INIS)

    Sengoku, Seio

    1985-08-01

    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.)

  10. Stability of high-beta tokamak equilibria and transport in Belt-Pinch IIa

    Energy Technology Data Exchange (ETDEWEB)

    Becker, G; Gruber, O; Krause, H; Mast, F; Wilhelm, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)

    1978-01-01

    In Belt-Pinch IIa, highly elongated equilibria with poloidal beta values up to the aspect ratio have been achieved. In these tokamak-like configurations, no fast-growing MHD instabilities such as external kink and ballooning modes have been observed. Rigid displacement instabilities have been stabilized by an appropriate poloidal magnetic field configuration and by a conducting shell. By comparing simulation experiments using the Garching high-beta transport code with measurements, it has been found that in the collision-dominated plasma no anomalously enhanced transport occurs. Transport theory in the Pfirsch-Schlueter regime, which includes elongation and high-beta effects, has been confirmed by the experiment. In particular, it has been shown that the perpendicular electrical conductivity is also classical. Detailed investigations of oxygen and carbon impurity losses demonstrated that the impurity subprograms commonly used for tokamaks underestimate the radiation losses in the range Tsub(e)=10 to 30 eV.

  11. MAIA, Eigenvalues for MHD Equation of Tokamak Plasma Stability Problems

    International Nuclear Information System (INIS)

    Tanaka, Y.; Azumi, M.; Kurita, G.; Tsunematsu, T.; Takeda, T.

    1986-01-01

    1 - Description of program or function: This program solves an eigenvalue problem zBx=Ax where A and B are real block tri-diagonal matrices. This eigenvalue problem is derived from a reduced set of linear resistive MHD equations which is often employed to study tokamak plasma stability problem. 2 - Method of solution: Both the determinant and inverse iteration methods are employed. 3 - Restrictions on the complexity of the problem: The eigenvalue z must be real

  12. Plasma current profile during current reversal in a tokamak

    International Nuclear Information System (INIS)

    Huang Jianguo; Yang Xuanzong; Zheng Shaobai; Feng Chunhua; Zhang Houxian; Wang Long

    1999-01-01

    Alternating current operation with one full cycle and a current level of 2.5 kA have been achieved in the CT-6B tokamak. The poloidal magnetic field in the plasma is measured with two internal magnetic probes in repeated discharges. The current distribution is reconstructed with an inversion algorithm. The inverse current first appears on the weak field side. The existence of magnetic surfaces and rotational transform provide particle confinement in the current reversal phase

  13. Burn stability of tokamak fusion plasmas with synergetic current drive

    International Nuclear Information System (INIS)

    Anderson, D.; Lisak, M.; Kolesnichenko, Ya.

    1991-01-01

    The stability of thermonuclear burn in Tokamak-reactors with non-inductive current generated with the simultaneous application of various methods is investigated. Particular emphasis is given to the ITER synergetic current drive scenario involving LH waves, neoclassical effects and NB injection. For ITER-like confinement laws, it is shown that this scenario may be unstable on the plasma skin time scale. Figs

  14. Profile formation and sustainment of autonomous tokamak plasma with current hole configuration

    International Nuclear Information System (INIS)

    Hayashi, N.; Takizuka, T.; Ozeki, T.

    2005-01-01

    We have investigated the profile formation and sustainment of tokamak plasmas with the current hole (CH) configuration by using 1.5D time-dependent transport simulations. A model of the current limit inside the CH on the basis of the Axisymmetric Tri-Magnetic-Islands equilibrium is introduced into the transport simulation. We found that a transport model with the sharp reduction of anomalous transport in the reversed-shear (RS) region can reproduce the time evolution of profiles observed in JT-60U experiments. The transport becomes neoclassical-level in the RS region, which results in the formation of profiles with internal transport barrier (ITB) and CH. The CH plasma has an autonomous property because of the strong interaction between a pressure profile and a current profile through the large bootstrap current fraction. The ITB width determined by the neoclassical-level transport agrees well with that measured in JT-60U. The energy confinement inside the ITB agrees with the scaling based on the JT-60U data. The scaling means the autonomous limitation of energy confinement in the CH plasma. The plasma with the large CH is sustained with the full current drive by the bootstrap current. The plasma with the small CH and the small bootstrap current fraction shrinks due to the penetration of inductive current. This shrink is prevented and the CH size can be controlled by the appropriate external current drive (CD). The CH plasma is found to respond autonomically to the external CD. (author)

  15. Tokamak plasma current disruption infrared control system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ulrickson, M.

    1987-01-01

    This patent describes a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of the vacuum chamber and an arrangement for the rapid prediction and control in real time of a major plasma disruption. The arrangement is described which includes: scanning means sensitive to infrared radiation emanating from within the vacuum chamber, the infrared radiation indicating the temperature along a vertical profile of the inner toroidal limiter. The scanning means is arranged to observe the infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along the vertical profile; detection means for analyzing the scanning output signal to detect a first peaked temperature excursion occurring along the profile of the inner toroidal limiter, and to produce a detection output signal in repsonse thereto, the detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing the plasma current driving the plasma, in response to the detection output signal and in anticipation of a subsequent major plasma disruption

  16. 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.

  17. Plasma diagnostics for tokamaks and stellarators

    International Nuclear Information System (INIS)

    Stott, P.E.; Sanchez, J.

    1994-01-01

    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 Sattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma

  18. Effect of alpha drift and instabilities on tokamak plasma edge conditions

    International Nuclear Information System (INIS)

    Miley, G.H.; Choi, C.K.

    1983-01-01

    As suprathermal fusion products slow down in a Tokamak, their average drift is inward. The effect of this drift on the alpha heating and thermalization profiles is examined. In smaller TFTR-type devices, heating in the outer region can be cut in half. Also, the fusion-product energy-distribution near the plasma edge has a positive slope with increasing energy, representing a possible driving mechanism for micro-instabilities. Another instability that can seriously affect outer plasma conditions and shear Alfven transport of alphas is also considered

  19. Impact of magnetic perturbation fields on tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team

    2015-05-01

    Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.

  20. Thermonuclear Tokamak plasmas in the presence of fusion alpha particles

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1988-01-01

    In this overview, we have focused on several results of the thermonuclear plasma research pertaining to the alpha particle physics and diagnostics in a fusion tokamak plasma. As regards the discussion of alpha particle effects, two distinct classes of phenomena have been distinguished: the simpler class containing phenomena exhibited by individual alpha particles under the influence of bulk plasma properties and, the more complex class including collective effects which become important for increasing alpha particle density. We have also discussed several possibilities to investigate alpha particle effects by simulation experiments using an equivalent population of highly energetic ions in the plasma. Generally, we find that the present theoretical knowledge on the role of fusion alpha particles in a fusion tokamak plasma is incomplete. There are still uncertainties and partial lack of quantitative results in this area. Consequently, further theoretical work and, as far a possible, simulation experiments are needed to improve the situation. Concerning the alpha particle diagnostics, the various diagnostic techniques and the status of their development have been discussed in two different contexts: the escaping alpha particles and the confined alpha particles in the fusion plasma. A general conclusion is that many of the different diagnostic methods for alpha particle measurements require further major development. (authors)

  1. Runaway transport studies in the TJ-I tokamak

    International Nuclear Information System (INIS)

    Rodriguez, L.; Vega, J.; Rodriguez-Yunta, A.; Castejon, F.; Pardo, C.; Navarro, A. P.

    1994-01-01

    Runaway diffusion coefficient, Dr, for TJ-I tokamak has been deduced using two different methods: In the first one, Dr is obtained using the steady state approach for values of the runaway confinement time Tτ r deduced from hard-X-ray bremsstrahlung spectra; in the second method, D r is deduced from sawteeth oscillations of HXR flux, and of SXR intensity, recorded simultaneously. Data have been taken in a scan with the toroidal magnetic field Bp Values obtained for Dr=[5-10] m 2 s ,decreasing with B Γ , are in both cases consistent. Assuming that magnetic turbulence is responsible for their transport, from D r we can deduce b/B Γ , the magnetic fluctuations level, and then infer the thermal conductivity coefficient χ c . The radial energy resolution of runaway electrons allow us to infer for b B Γ and χ c values at the plasma edge and for an inner position about r=a/2. Results are in good agreement with those ones obtained in TJ-I using probes, spectroscopic methods, power balance analysis and a coupled analysis for temperature and density pulses. (Author) 32 refs

  2. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

    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) [fr

  3. Experimental investigations of driven Alfven wave resonances in a tokamak plasma using carbon dioxide laser interferometry

    International Nuclear Information System (INIS)

    Evans, T.E.

    1984-09-01

    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

  4. Energy and particle core transport in tokamaks and stellarators compared

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, Marc; Angioni, Clemente; Beidler, Craig; Dinklage, Andreas; Fuchert, Golo; Hirsch, Matthias; Puetterich, Thomas; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald/Garching (Germany)

    2016-07-01

    The paper discusses expectations for core transport in the Wendelstein 7-X stellarator (W7-X) and presents a comparison to tokamaks. In tokamaks, the neoclassical trapped-particle-driven losses are small and turbulence dominates the energy and particle transport. At reactor relevant low collisionality, the heat transport is limited by ion temperature gradient limited turbulence, clamping the temperature gradient. The particle transport is set by an anomalous inward pinch, yielding peaked profiles. A strong edge pedestal adds to the good confinement properties. In traditional stellarators the 3D geometry cause increased trapped orbit losses. At reactor relevant low collisionality and high temperatures, these neoclassical losses would be well above the turbulent transport losses. The W7-X design minimizes neoclassical losses and turbulent transport can become dominant. Moreover, the separation of regions of bad curvature and that of trapped particle orbits in W7-X may have favourable implications on the turbulent electron heat transport. The neoclassical particle thermodiffusion is outward. Without core particle sources the density profile is flat or even hollow. The presence of a turbulence driven inward anomalous particle pinch in W7-X (like in tokamaks) is an open topic of research.

  5. Liquid gallium jet-plasma interaction studies in ISTTOK tokamak

    International Nuclear Information System (INIS)

    Gomes, R.B.; Fernandes, H.; Silva, C.; Sarakovskis, A.; Pereira, T.; Figueiredo, J.; Carvalho, B.; Soares, A.; Duarte, P.; Varandas, C.; Lielausis, O.; Klyukin, A.; Platacis, E.; Tale, I.; Alekseyv, A.

    2009-01-01

    Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma contamination by liquid metal. Additionally the response of an infrared sensor, intended to measure the jet surface temperature increase during its interaction with the plasma, has been studied. The jet power extraction capability is extrapolated from the heat flux profiles measured in ISTTOK plasmas.

  6. Impurity screening of scrape-off plasma in a tokamak

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi; Tani, Keiji; Nakamura, Hiroo

    1981-11-01

    Impurity screening effect of a scrape-off layer has been studied in a tokamak, based on a simple model of wall-released impurity behavior. Wall-sputtered impurities are stopped effectively by the scrape-off plasma for a medium-Z or high-Z wall system while major part of impurities enters the main plasma in a low-Z wall system. The screening becomes inefficient with increase of scrape-off plasma temperature. Successive multiplication of recycling impurities in the scrape-off layer is large for a high-Z wall and is enhanced by a rise of scrape-off plasma temperature. The stability of plasma-wall interaction is determined by a multiplication factor of recycling impurities. (author)

  7. Development of a tokamak plasma optimized for stability and confinement

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement

  8. The role of high speed photography in plasma instability research on the AEC tokamak

    International Nuclear Information System (INIS)

    Fletcher, J.D.; Coster, D.P.; De Villiers, J.A.M.; Kotze, P.B.; Nothnagel, G.; O'Mahony, J.R.; Roberts, D.E.; Sherwell, D.

    1986-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions in fusion research devices like tokamaks. Such a system has been installed on the AEC tokamak. This paper reports some preliminary results obtained during typical plasma discharges

  9. Pellet-plasma interactions in tokamaks

    DEFF Research Database (Denmark)

    Chang, C.T.

    1991-01-01

    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......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...

  10. Electron cyclotron emission imaging in tokamak plasmas

    NARCIS (Netherlands)

    Munsat, T.; Domier, C.W.; Kong, X. Y.; Liang, T. R.; N C Luhmann Jr.,; Tobias, B. J.; Lee, W.; Park, H. K.; Yun, G.; Classen, I.G.J.; Donne, A. J. H.

    2010-01-01

    We discuss the recent history and latest developments of the electron cyclotron emission imaging diagnostic technique, wherein electron temperature is measured in magnetically confined plasmas with two-dimensional spatial resolution. The key enabling technologies for this technique are the

  11. Electron temperature gradient driven instability in the tokamak boundary plasma

    International Nuclear Information System (INIS)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-01-01

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt

  12. Evaluation of the plasma parameters in COMPASS tokamak divertor area

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  13. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  14. Low temperature plasma near a tokamak reactor limiter

    International Nuclear Information System (INIS)

    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 13 cm -3 or less and the thermal diffusivity in the edge region is 2 x 10 4 cm 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

  15. Alpha transport and blistering in tokamaks

    International Nuclear Information System (INIS)

    Bauer, W.; Wilson, K.L.; Bisson, C.L.; Haggmark, L.G.; Goldston, R.J.

    1978-12-01

    The particle flux and angular distribution of 3.5 MeV alpha particles impinging on the first wall from uncontained banana orbits in an axisymmetric tokamak reactor have been calculated. The resulting helium concentration profiles in the first wall can give rise to surface exfoliation under specified conditions. The major mitigating factor is the simultaneous surface recession due to sputtering by the D-T charge exchange neutral flux. For the parameters used in these calculations blistering in high sputtering rate materials such as Be is unlikely whereas in low sputtering rate materials such as Nb, He induced surface deformation is quite probable

  16. Fusion performance analysis of plasmas with reversed magnetic shear in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Ruskov, E.; Bell, M.; Budny, R.V.; McCune, D.C.; Medley, S.S.; Nazikian, R.; Synakowski, E.J.; Goeler, S. von; White, R.B.; Zweben, S.J.

    1999-01-01

    A case for substantial loss of fast ions degrading the performance of tokamak fusion test reactor plasmas [Phys. Plasmas 2, 2176 (1995)] with reversed magnetic shear (RS) is presented. The principal evidence is obtained from an experiment with short (40 - 70 ms) tritium beam pulses injected into deuterium beam heated RS plasmas [Phys. Rev. Lett. 82, 924 (1999)]. Modeling of this experiment indicates that up to 40% beam power is lost on a time scale much shorter than the beam - ion slowing down time. Critical parameters which connect modeling and experiment are: The total 14 MeV neutron emission, its radial profile, and the transverse stored energy. The fusion performance of some plasmas with internal transport barriers is further deteriorated by impurity accumulation in the plasma core. copyright 1999 American Institute of Physics

  17. Experimental validation of a Lyapunov-based controller for the plasma safety factor and plasma pressure in the TCV tokamak

    Science.gov (United States)

    Mavkov, B.; Witrant, E.; Prieur, C.; Maljaars, E.; Felici, F.; Sauter, O.; the TCV-Team

    2018-05-01

    In this paper, model-based closed-loop algorithms are derived for distributed control of the inverse of the safety factor profile and the plasma pressure parameter β of the TCV tokamak. The simultaneous control of the two plasma quantities is performed by combining two different control methods. The control design of the plasma safety factor is based on an infinite-dimensional setting using Lyapunov analysis for partial differential equations, while the control of the plasma pressure parameter is designed using control techniques for single-input and single-output systems. The performance and robustness of the proposed controller is analyzed in simulations using the fast plasma transport simulator RAPTOR. The control is then implemented and tested in experiments in TCV L-mode discharges using the RAPTOR model predicted estimates for the q-profile. The distributed control in TCV is performed using one co-current and one counter-current electron cyclotron heating actuation.

  18. Tokamak plasma variations under rapid compression

    International Nuclear Information System (INIS)

    Holmes, J.A.; Peng, Y.K.M.; Lynch, S.J.

    1980-04-01

    Changes in plasmas undergoing large, rapid compressions are examined numerically over the following range of aspect ratios A:3 greater than or equal to A greater than or equal to 1.5 for major radius compressions of circular, elliptical, and D-shaped cross sections; and 3 less than or equal to A less than or equal to 6 for minor radius compressions of circular and D-shaped cross sections. The numerical approach combines the computation of fixed boundary MHD equilibria with single-fluid, flux-surface-averaged energy balance, particle balance, and magnetic flux diffusion equations. It is found that the dependences of plasma current I/sub p/ and poloidal beta anti β/sub p/ on the compression ratio C differ significantly in major radius compressions from those proposed by Furth and Yoshikawa. The present interpretation is that compression to small A dramatically increases the plasma current, which lowers anti β/sub p/ and makes the plasma more paramagnetic. Despite large values of toroidal beta anti β/sub T/ (greater than or equal to 30% with q/sub axis/ approx. = 1, q/sub edge/ approx. = 3), this tends to concentrate more toroidal flux near the magnetic axis, which means that a reduced minor radius is required to preserve the continuity of the toroidal flux function F at the plasma edge. Minor radius compressions to large aspect ratio agree well with the Furth-Yoshikawa scaling laws

  19. Triangularity effects on the collisional diffusion for elliptic tokamak plasma

    International Nuclear Information System (INIS)

    Martin, P.; Castro, E.

    2007-01-01

    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)

  20. Extremely shaped plasmas to improve the Tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Piras, F.

    2011-04-15

    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

  1. Extremely shaped plasmas to improve the Tokamak concept

    International Nuclear Information System (INIS)

    Piras, F.

    2011-04-01

    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

  2. On fast-ion transport and burn control in Tokamaks

    International Nuclear Information System (INIS)

    Wising, F.

    1994-01-01

    Fast ions, generated by e.g. neutral beam injection (NBI), radio frequency (RF) heating or nuclear reactions, play an important role in all large tokamaks. Several issues related to fast ions and burning fusion plasmas are addressed in this thesis. Firstly, a new model of sawtooth oscillations is developed which explains the recent observations that q 0 remains below unity during the entire sawtooth cycle. The model features full reconnection in two current layers and provides a self-consistent description of the plasma states before and after the sawtooth crash. It is applied to the redistribution of fast NBI-ions in JET and comparisons are made with global as well as line-of-sight integrated D-D neutron measurements. Both the new model and the classical Kadomtsev model are found to be in agreement with the measurements. A simplified redistribution model is developed and applied to the redistribution of tritons and thermal ions, again giving reasonable agreement with D-T/D-D neutron measurements. Using a separate method, earlier results on expulsion of NBI-ions are confirmed. Secondly, a numerical study has been carried out of the coupled nonlinear evolution of alpha-particle driven kinetic Alfven wave turbulence and associated alpha transport. The saturated fluctuation spectrum consists of two peaks and results from nonlinear ion Compton scattering-induced transfer of energy from longer to shorter wavelengths. An analytical solution of the saturated spectrum, and estimates of the anomalous alpha diffusion coefficient, are given. The final paper addresses the problem of determining whether an initial temperature profile, established by e.g. auxiliary heating, will evolve to thermonuclear burn or quench under the influence of alpha particle heating and thermal conduction. Explicit burn criteria are presented and the beneficial effects of density and temperature peaking are discussed. 110 refs

  3. The COMPASS Tokamak Plasma Control Software Performance

    Czech Academy of Sciences Publication Activity Database

    Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír

    2011-01-01

    Roč. 58, č. 4 (2011), s. 1490-1496 ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726

  4. Radial electric field at the plasma edge on the FT-2 Tokamak in regimes with large gradients

    International Nuclear Information System (INIS)

    Lashkul, S.; Popov, A.

    2001-01-01

    The transport barrier formation is widely believed to be the fundamental element of transition into improved confinement regimes (H-mode). Experiments on many tokamaks demonstrate that transport barrier formation is connected with the suppression of turbulent transport by shear of E x B drift. Therefore, the calculation of radial electric field is of great importance. Our work is devoted to progress the neoclassical theory by taking into account electron viscosity and non-linear effects (ion inertia), presented results being valuable for interpretation transition into H-mode at the plasma edge in small tokamaks. Calculations of the electric field profile for FT-2 tokamak (a=8cm, R 0 =55cm, Ioffe Institute, Russia) according found expressions are in the good agreement with experimental results obtained. (orig.)

  5. Characterization and scaling of the tokamak edge transport barrier

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Philip Adrian

    2012-04-24

    The high confinement regime (H-mode) in a tokamak plasma displays a remarkable edge region. On a small spatial scale of 1-2 cm the properties of the plasma change significantly. Certain parameters vary 1-2 orders of magnitude in this region, called the pedestal. Currently, there is no complete understanding of how the pedestal forms or how it is sustained. The goal of this thesis is to contribute to the theoretical understanding of the pedestal and provide scalings towards larger machines, like ITER and DEMO. A pedestal database was built with data from different tokamaks: ASDEX Upgrade, DIIID and JET. The pedestal was characterized with the same method for all three machines. This gives the maximum value, gradient and width of the pedestal in n{sub e}, T{sub e} and T{sub i}. These quantities were analysed along with quantities derived from them, such as the pressure or the confinement time. For this purpose two parameter sets were used: normalized parameters (pressure {beta}, time {nu}{sub *}, length {rho}{sub *}, shape f{sub q}) and machine parameters (size a, magnetic field B{sub t}, plasma current I{sub p}, heating P). All results are dependent on the choice of the coordinate system: normalized poloidal flux {Psi}{sub N} and real space r/a. The most significant result, which was obtained with both parameter sets, shows a different scaling of the pedestal width for the electron temperature and the electron density. The presented scalings predict that in ITER and DEMO the temperature pedestal will be appreciably wider than the density pedestal. The pedestal top scaling for the pressure reveals differences between the electron and the ion pressure. In extrapolations this results in values for T{sub e,ped} of 4 keV (ITER) and 10 keV (DEMO), but significantly lower values for the ion temperature. A two-term method was applied to use the pedestal pressure to determine the pedestal contribution to the global confinement time {tau}{sub E}. The dependencies in the

  6. Characterization and scaling of the tokamak edge transport barrier

    International Nuclear Information System (INIS)

    Schneider, Philip Adrian

    2012-01-01

    The high confinement regime (H-mode) in a tokamak plasma displays a remarkable edge region. On a small spatial scale of 1-2 cm the properties of the plasma change significantly. Certain parameters vary 1-2 orders of magnitude in this region, called the pedestal. Currently, there is no complete understanding of how the pedestal forms or how it is sustained. The goal of this thesis is to contribute to the theoretical understanding of the pedestal and provide scalings towards larger machines, like ITER and DEMO. A pedestal database was built with data from different tokamaks: ASDEX Upgrade, DIIID and JET. The pedestal was characterized with the same method for all three machines. This gives the maximum value, gradient and width of the pedestal in n e , T e and T i . These quantities were analysed along with quantities derived from them, such as the pressure or the confinement time. For this purpose two parameter sets were used: normalized parameters (pressure β, time ν * , length ρ * , shape f q ) and machine parameters (size a, magnetic field B t , plasma current I p , heating P). All results are dependent on the choice of the coordinate system: normalized poloidal flux Ψ N and real space r/a. The most significant result, which was obtained with both parameter sets, shows a different scaling of the pedestal width for the electron temperature and the electron density. The presented scalings predict that in ITER and DEMO the temperature pedestal will be appreciably wider than the density pedestal. The pedestal top scaling for the pressure reveals differences between the electron and the ion pressure. In extrapolations this results in values for T e,ped of 4 keV (ITER) and 10 keV (DEMO), but significantly lower values for the ion temperature. A two-term method was applied to use the pedestal pressure to determine the pedestal contribution to the global confinement time τ E . The dependencies in the scaling for τ E,ped are nearly identical to the IPB98 global

  7. Investigation of plasma equilibrium in HL-1 tokamak

    International Nuclear Information System (INIS)

    Lu Zhihong; Jiang Yunxia; Yang Jinwei; Zhang Baozhu; Qiu Wei; Qin Yunwen

    1987-01-01

    In this paper, the plasma equilibrium in HL-1 tokamak has been discussed. The horizontal and vertical displacement of plasma is measured using a symmetical magneic probe system, and the temporal evolution of displacements is given by a data acquisition system with micro-computer. The influence of various stray fields on plasma equilibrium has been analysed. The direction and value of horizontal stray field induced by the totoidal field coils and primary windings are determined using vertical displacement data. By adjusting parameters of internal and external vertical field, in the flat part of discharge current, the plasma can be kept in its equilibuium state at the place where is 3 cm outer of chamber certre, i.e nearby the centre of limiter

  8. Positional instability analysis of tokamak plasmas by ERATO

    International Nuclear Information System (INIS)

    Kumagai, Michikazu; Tsunematsu, Toshihide; Tokuda, Shinji; Takeda, Tatsuoki

    1983-06-01

    The stability of axisymmetric modes of a tokamak plasma(positional instabilities) is analyzed for the Solov'ev equilibrium by using the linear ideal MHD code ERATO-J. The dependence of the stability on various parameters, i.e., the ellipticity and triangularity of the plasma cross-section, the aspect ratio, the safety factor at the magnetic axis, and the distance between the plasma and a conducting shell is investigated. Comparison of the results with those by the rigid model shows that the stability condition derived from the rigid model in terms of the decay index(n-index) of the external equilibrating field is a good approximation for the plasma with small triangular deformation. Also the results are compared with those of the rigid displacement model and applicability of the various models on the positional instability analyses is discussed. (author)

  9. Plasma vertical instability in a tokamak with rail limiters

    International Nuclear Information System (INIS)

    Belashob, V.I.; Brevnov, N.N.; Gribov, Yu.V.; Putvinskij, S.V.

    1989-01-01

    An effect of currents between rail limiters on plasma equilibrium in the tokamak is studied theoretically and experimentally. Limiter currents can emerge at fast changes of plasma position along rail limiters for example when compression along major radius takes place and result in additional electrodynamic loadings on to the chamber and limiters. It is shown that at high currents between the limiters, the behaviour of discharge depends on limiter voltage polarity. When the plasma - limiter contact points are asymmetrically located respective to an equatorial plane a radial component of the limiter current emerges. The interaction of the component with the toroidal magnetic field can result in a vertical plasma filament instability. 9 refs.; 10 figs

  10. Probabilistic analysis of tokamak plasma disruptions

    International Nuclear Information System (INIS)

    Sanzo, D.L.; Apostolakis, G.E.

    1985-01-01

    An approximate analytical solution to the heat conduction equations used in modeling component melting and vaporization resulting from plasma disruptions is presented. This solution is then used to propagate uncertainties in the input data characterizing disruptions, namely, energy density and disruption time, to obtain a probabilistic description of the output variables of interest, material melted and vaporized. (orig.)

  11. Real-time control of Tokamak plasmas: from control of physics to physics-based control

    International Nuclear Information System (INIS)

    Felici, F. A. A.

    2011-11-01

    shown effective stabilization of both 3/2 and 2/1 NTMs, and have localized the most effective deposition location. Studies of current-profile driven destabilization of tearing modes in TCV plasmas with significant amounts of ECCD show a great sensitivity to details of the current profile, but failed to identify a stationary region in the parameter space in which NTMs are always destabilized, suggesting that transient effects play a role. The simultaneous control of magnetic and kinetic plasma profiles is another key requirement for advanced tokamak operation. While control of kinetic plasma profiles around an operating point can be handled using standard linear control techniques, the strongly nonlinear physics of the coupled profiles complicates the problem. Since internal magnetic quantities are difficult to measure with sufficient spatial and temporal resolution – even after years of diagnostic development – routine control of tokamak plasma profiles remains a daunting and challenging task. In this thesis, physics understanding of plasma current and energy transport is embedded in the control solution. The new lightweight transport code RAPTOR (RApid Plasma Transport simulatOR) has been derived focusing on simplicity and speed of simulation for real-time control. The partial differential equation for current diffusion is solved in real-time during a plasma shot in the TCV control system using RAPTOR. For the first time, this concept is applied experimentally to the tokamak current density profile problem. The real-time simulation gives a physics-model based estimate of key plasma quantities, to be controlled or monitored in real-time by different control systems. Any available diagnostics can be included into the simulation providing additional constraints and removing measurement uncertainties. The real-time simulation approach holds the advantage that knowledge of the plasma profiles is no longer restricted to those points in space and time where they are

  12. Statistical properties of turbulent transport and fluctuations in tokamak and stellarator devices

    Energy Technology Data Exchange (ETDEWEB)

    Hidalgo, C; Pedrosa, M A; Milligen, B Van; Sanchez, E; Balbin, R; Garcia-Cortes, I [Euratom-CIEMAT Association, Madrid (Spain); Bleuel, J; Giannone, L.; Niedermeyer, H [Euratom-IPP Association, Garching (Germany)

    1997-05-01

    The statistical properties of fluctuations and turbulent transport have been studied in the plasma boundary region of stellarator (TJ-IU, W7-AS) and tokamak (TJ-I) devices. The local flux probability distribution function shows the bursty character of the flux and presents a systematic change as a function of the radial location. There exist large amplitude transport bursts that account for a significant part of the total flux. There is a strong similarity between the statistical properties of the turbulent fluxes in different devices. The value of the radial coherence associated with fluctuations and turbulent transport is strongly intermittent. This result emphasizes the importance of measurements with time resolution in understanding the interplay between the edge and the core regions in the plasma. For measurements in the plasma edge region of the TJ-IU torsatron, the turbulent flux does not, in general, show a larger radial coherence than the one associated with the fluctuations. (author). 14 refs, 6 figs.

  13. Measurement of the effective plasma ion mass in large tokamaks

    International Nuclear Information System (INIS)

    Lister, J.B.; Villard, L.; Ridder, G. de

    1997-01-01

    There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs

  14. Magnetohydrodynamic equilibria and local stability of axisymmetric tokamak plasmas

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Dory, R.A.; Nelson, D.B.; Sayer, R.O.

    1976-07-01

    Axisymmetric magnetohydrodynamic equilibria are evaluated in terms of the Mercier Stability Criterion. The parameters of interest include poloidal beta (β/sub p/), current and pressure profile widths, D-shaped and doublet plasmas with elongation (sigma) and triangularity (delta), and the aspect ratio (A). For marginal local stability, the critical values of β, plasma current, and the safety factor q with fixed toroidal field at the geometric center of the plasma are obtained. It is shown that for a wide range of profiles in a D-shaped plasma with A = 3, the highest critical β occurs at β/sub p/ = 2.4, sigma = 1.65, and delta = 0.5. If the toroidal field at the coil surface is fixed, the highest critical pressure occurs near A approximately 3 to 4, given reasonable distance between the coils and the plasma edge. Calculations for a Doublet II-A plasma with sigma = 3 show that with similar pressure profile the highest critical β occurs at β/sub p/ = 1 and is 84 percent of the highest critical β for the D-shaped plasmas. Critical values of ohmic heating power density are also found to be comparable for the two plasma shapes. A D-shaped plasma with the above parameters is suggested for use in future high-β tokamak devices

  15. Nonlinear neoclassical transport in toroidal edge plasmas

    International Nuclear Information System (INIS)

    Fueloep, T.; Helander, P.

    2001-01-01

    In conventional neoclassical theory, the density and temperature gradients are not allowed to be as steep as frequently observed in the tokamak edge. In this paper the theory of neoclassical transport in a collisional, impure plasma is extended to allow for steeper profiles than normally assumed. The dynamics of highly charged impurity ions then becomes nonlinear, which affects the transport of all species. As earlier found in the banana regime, when the bulk plasma gradients are large the impurity ions undergo a poloidal redistribution, which reduces their parallel friction with the bulk ions and suppresses the neoclassical ion particle flux. The neoclassical confinement is thus improved in regions with large radial gradients. When the plasma is collisional and the gradients are large, the impurities accumulate on the inboard side of the torus

  16. Online Plasma Shape Reconstruction for EAST Tokamak

    International Nuclear Information System (INIS)

    Luo Zhengping; Xiao Bingjia; Zhu Yingfei; Yang Fei

    2010-01-01

    An online plasma shape reconstruction, based on the offline version of the EFIT code and MPI library, can be carried out between two adjacent shots in EAST. It combines online data acquisition, parallel calculation, and data storage together. The program on the master node of the cluster detects the termination of the discharge promptly, reads diagnostic data from the EAST mdsplus server on the completion of data storing, and writes the results onto the EFIT mdsplus server after the calculation is finished. These processes run automatically on a nine-nodes IBM blade center. The total time elapsed is about 1 second to several minutes, depending on the duration of the shot. With the results stored in the mdsplus server, it is convenient for operators and physicists to analyze the behavior of plasma using visualization tools.

  17. Simulation of W dust transport in the KSTAR tokamak, comparison with fast camera data

    Directory of Open Access Journals (Sweden)

    A. Autricque

    2017-08-01

    Full Text Available In this paper, dust transport in tokamak plasmas is studied through both experimental and modeling aspects. Image processing routines allowing dust tracking on CCD camera videos are presented. The DUMPRO (DUst Movie PROcessing code features a dust detection method and a trajectory reconstruction algorithm. In addition, a dust transport code named DUMBO (DUst Migration in a plasma BOundary is briefly described. It has been developed at CEA in order to simulate dust grains transport in tokamaks and to evaluate the contribution of dust to the impurity inventory of the plasma. Like other dust transport codes, DUMBO integrates the Orbital Motion Limited (OML approach for dust/plasma interactions modeling. OML gives direct expressions for plasma ions and electrons currents, forces and heat fluxes on a dust grain. The equation of motion is solved, giving access to the dust trajectory. An attempt of model validation is made through comparison of simulated and measured trajectories on the 2015 KSTAR dust injection experiment, where W dust grains were successfully injected in the plasma using a gun-type injector. The trajectories of the injected particles, estimated using the DUMPRO routines applied on videos from the fast CCD camera in KSTAR, show two distinct general dust behaviors, due to different dust sizes. Simulations were made with DUMBO to match the measurements. Plasma parameters were estimated using different diagnostics during the dust injection experiment plasma discharge. The experimental trajectories show longer lifetimes than the simulated ones. This can be due to the substitution of a boiling/sublimation point to the usual vaporization/sublimation cooling, OML limitations (eventual potential barriers in the vicinity of a dust grain are neglected and/or to the lack of a vapor shielding model in DUMBO.

  18. Numerical Simulation of Neoclassical Currents, Parallel Viscosity, and Radial Current Balance in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Kiviniemi, T.

    2001-01-01

    One of the principal problems en route to a fusion reactor is that of insufficient plasma confinement, which has lead to both theoretical and experimental research into transport processes in the parameter range relevant for fusion energy production. The neoclassical theory of tokamak transport is well-established unlike the theory of turbulence driven anomalous transport in which extensive progress has been made during last few years. So far, anomalous transport has been dominant in experiments, but transport may be reduced to the neoclassical level in advanced tokamak scenarios. This thesis reports a numerical study of neoclassical fluxes, parallel viscosity, and neoclassical radial current balance in tokamaks. Neoclassical parallel viscosity and particle fluxes are simulated over a wide range of collisionalities, using the fully kinetic five-dimensional neoclassical orbit-following Monte Carlo code ASCOT. The qualitative behavior of parallel viscosity derived in earlier analytic models is shown to be incorrect for high poloidal Mach numbers. This is because the poloidal dependence of density was neglected. However, in high Mach number regime, it is the convection and compression terms, rather than the parallel viscosity term, that are shown to dominate the momentum balance. For fluxes, a reasonable agreement between numerical and analytical results is found in the collisional parameter regime. Neoclassical particle fluxes are additionally studied in the banana regime using the three-dimensional Fokker-Planck code DEPORA, which solves the drift-kinetic equation with finite differencing. Limitations of the small inverse aspect ratio approximation adopted in the analytic theory are addressed. Assuming that the anomalous transport is ambipolar, the radial electric field and its shear at the tokamak plasma edge can be solved from the neoclassical radial current balance. This is performed both for JET and ASDEX Upgrade tokamaks using the ASCOT code. It is shown that

  19. Helium, iron and electron particle transport and energy transport studies on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Grek, B.; Hill, K.W.; Hulse, R.A.; Johnson, D.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Redi, M.H.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor

  20. Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

    Science.gov (United States)

    Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

  1. Axisymmetric MHD simulation of ITB crash and following disruption dynamics of Tokamak plasmas with high bootstrap current

    International Nuclear Information System (INIS)

    Takei, Nahoko; Tsutsui, Hiroaki; Tsuji-Iio, Shunji; Shimada, Ryuichi; Nakamura, Yukiharu; Kawano, Yasunori; Ozeki, Takahisa; Tobita, Kenji; Sugihara, Masayoshi

    2004-01-01

    Axisymmetric MHD simulation using the Tokamak Simulation Code demonstrated detailed disruption dynamics triggered by a crash of internal transport barrier in high bootstrap current, high β, reversed shear plasmas. Self-consistent time-evolutions of ohmic current bootstrap current and induced loop voltage profiles inside the disrupting plasma were shown from a view point of disruption characterization and mitigation. In contrast with positive shear plasmas, a particular feature of high bootstrap current reversed shear plasma disruption was computed to be a significant change of plasma current profile, which is normally caused due to resistive diffusion of the electric field induced by the crash of internal transport barrier in a region wider than the internal transport barrier. Discussion based on the simulation results was made on the fastest record of the plasma current quench observed in JT-60U reversed shear plasma disruptions. (author)

  2. The effect of non-inductive current drive on tokamak transport

    International Nuclear Information System (INIS)

    Helander, P; Akers, R J; Valovic, M; Peysson, Y

    2005-01-01

    Non-inductive current drive causes cross-field neoclassical transport in a tokamak, in much the same way that the toroidal electric field used to drive the plasma current produces the so-called Ware pinch. This transport can be either inwards or outwards, depending on the current drive mechanism, and can be either larger or smaller than the analogous Ware pinch. A Green's function formalism is used to calculate the transport produced by wave-driven currents, which is found to be inwards for electron-cyclotron and lower-hybrid current drive. Its magnitude is proportional to the collisionality of the current-carrying electrons and therefore smaller than the Ware pinch when the resonant electrons are suprathermal. In contrast, neutral-beam current drive produces outward particle transport when the beams are injected in the same toroidal direction as the plasma current, and inward particle transport otherwise. This transport is somewhat larger than the corresponding Ware pinch. Together, they may explain an observation made on several tokamaks over the years, most recently on MAST, that density profiles tend to be more peaked during counter-injection

  3. Neoclassical transport analysis for a class of high-β tokamak equilibria

    International Nuclear Information System (INIS)

    Rieser, H.; Werthmann, H.; Kuhn, S.

    1995-01-01

    Balescu's neoclassical transport theory is extended to the case of non-circular flux-surface geometries. Modified classical and neoclassical transport equations, governing particle and heat fluxes in the short- and long-mean-free-path regimes, are derived. These equations are shown to coincide to leading order with the corresponding equations given by Hirshman and Sigmar. They are then applied to an ideal MHD equilibrium, suitable as a simplified but analytically tractable model of a high-β tokamak. Numerical results for the radial profiles of the global (i.e. flux-surface integrated) particle and heat fluxes in the classical, Pfirsch-Schlueter and banana regimes are presented for geometry and plasma parameters realized in some tokamaks, like the divertor and injection tokamak experiment (DITE). This spatial representation provides direct insight into the overall collisional transport behaviour of a given equilibrium, whereas the anomalous transport problem is not addressed here. Our results demonstrate that for a given pressure profile the global neoclassical fluxes may depend very sensitively on the temperature profiles and that, in particular, the global classical and neoclassical ion heat fluxes exhibit a characteristic non-monotonic behaviour. (author)

  4. Theoretical issues in tokamak confinement: (i) Internal/edge transport barriers and (ii) runaway avalanche confinement

    International Nuclear Information System (INIS)

    Connor, J.W.; Helander, P.; Thyagaraja, A.; Andersson, F.; Fueloep, T.; Eriksson, L.-G.; Romanelli, M.

    2001-01-01

    This paper summarises a number of distinct, but related, pieces of work on key confinement issues for tokamaks, in particular the formation of internal and edge transport barriers, both within turbulent and neoclassical models, and radial diffusion of avalanching runaway electrons. First-principle simulations of tokamak turbulence and transport using the two-fluid, electromagnetic, global code CUTIE are described. The code has demonstrated the spontaneous formation of internal transport barriers near mode rational surfaces, in qualitative agreement with observations on JET and RTP. The theory of neoclassical transport in an impure, toroidal plasma has been extended to allow for steeper pressure and temperature gradients than are usually considered, and is then found to become nonlinear under conditions typical of the tokamak edge. For instance, the particle flux is found to be a nonmonotonic function of the gradients, thus allowing for a bifurcation in the ion particle flux. Finally, it is shown that radial diffusion caused by magnetic fluctuations can effectively suppress avalanches of runaway electrons if the fluctuation amplitude exceeds δB/B∼10 -3 . (author)

  5. Plasma shape experiments for an optimized tokamak

    International Nuclear Information System (INIS)

    Hyatt, A.W.; Osborne, T.H.; Lazarus, E.A.

    1994-07-01

    In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, βτ E , in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed ''slot'' divertor geometry. power flux to the divertor floor using a closed ''slot'' divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes βτ E and divertor performance. The four shapes studied form a matrix of moderate and high elongations (κ congruent 1.8 and 2.1) and low and high triangularities (δ congruent 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot's transient normalized performance, β N H, where β N ≡ β/(I p )/aB T and H ≡ τ E /τ E ITER-89P , increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D

  6. Plasma shape experiments for an optimized tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hyatt, A.W.; Osborne, T.H. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

    1994-12-31

    In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, {beta}{sub {tau}E}, in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed `slot` divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes {beta}{sub {tau}E} and divertor performance. The four shapes studied form a matrix of moderate and high elongations ({kappa} {approx_equal} 1.8 and 2.1) and low and high triangularities ({delta} {approx_equal} 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot`s transient normalized performance, {beta}{sub N}H, where {beta}{sub N} = {beta}/(I{sub p}/aB{sub T}) and H = {tau}{sub E}/{tau}{sub E}{sup ITER-89P}, increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D. (author) 7 refs., 7 figs.

  7. Self-organized ignition of a tokamak plasma

    International Nuclear Information System (INIS)

    Schoepf, K.

    2007-01-01

    The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product nτ E T, where τ E represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual fusion power deposition [4,5] and its effect of reducing τ E are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself - towards an ignited

  8. The impurity transport in HT-6B tokamak

    International Nuclear Information System (INIS)

    Huang Rong; Xie Jikang; Li Linzhong; He Yexi; Wang Shuya; Deng Chuanbao; Li Guoxiang; Qiu Lijian

    1992-06-01

    The quasi-stationary profiles of the impurity ionization stages in HT-6B tokamak were determined by monitoring the VUV (vacuum ultraviolet) and visible line emissions from impurities. An impurity transport code was set up. The impurity transport coefficients and other parameters of impurities in that device were simulated and determined. From the measurement of impurity emission profiles and simulation analysis, it is concluded that the impurity confinement is improved and the impurity recycling is reduced by the slow magnetic compression. Some characteristics of impurity transport in that device are also discussed

  9. Heavy Neutral Beam Probe for edge plasma analysis in tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The Heavy Neutral Beam Probe project presented in this document is part of 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 Canadian de Fusion Magnetique. The overall objective of the effort is to apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes facility in Montreal, Canada. To achieve this goal, a research and development project was started in December, 1990 to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present, satisfactory progress has been achieved. The ion gun is fully operational with the neutralizer in the final assembly stage in preparation for testing. The beam diagnostics have been completed and mounted in the computer automated test stand. The analyzer design and detailed trajectory calculations are nearing completion to allow for the vacuum interface construction. The CAMAC based data acquisition system hardware was integrated into the test stand. Part of this hardware is a component of the Tokamak de Varennes' contribution to the collaboration. Next steps on the critical path include the beginning of the neutralization tests and the start of the analyzer construction. Anticipated installation of the diagnostic on the tokamak is Spring 1992

  10. Trade studies of plasma elongation for next-step tokamaks

    International Nuclear Information System (INIS)

    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

  11. The evolution of the plasma current during tokamak disruptions

    International Nuclear Information System (INIS)

    Helander, P.; Andersson, F.; Anderson, D.; Lisak, M.; Eriksson, L.G.

    2004-01-01

    In a tokamak disruption, the ohmic plasma current is partly replaced by a current carried by runaway electrons. This process is analysed by combining the equations for runaway electron generation with Maxwell's equations for the evolution of the electric field. This allows a quantitative understanding to be gained of runaway production in present experiments, and extrapolation to be made to ITER. The runaway current typically becomes more peaked on the magnetic axis than the pre-disruption current. In fact, the central current density can rise although the total current falls, which may have implications for post-disruption plasma stability. Furthermore, it is found that the runaway current easily spreads radially in a filament way due to the high sensitivity of the runaway generation efficiency to plasma parameters. (authors)

  12. Plasma rotation under a driven radial current in a tokamak

    International Nuclear Information System (INIS)

    Chang, C.S.

    1999-01-01

    The neoclassical behaviour of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. An ambipolar radial electric field develops instantly in such a way that the driven current is balanced by a return current j p in the plasma. The j p x B torque pushes the plasma into a new rotation state both toroidally and poloidally. An anomalous toroidal viscosity is needed to avoid an extreme toroidal rotation speed. It is shown that the poloidal rotation relaxes to a new equilibrium speed, which is in general smaller than the E x B poloidal speed, and that the timescale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually much longer than the ion-ion collision time. (author)

  13. Full Tokamak discharge simulation and kinetic plasma profile control for ITER

    International Nuclear Information System (INIS)

    Hee Kim, S.

    2009-10-01

    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

  14. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Bell, M.G.; Beer, M.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

  15. On transport and the bootstrap current in toroidal plasmas

    International Nuclear Information System (INIS)

    Connor, J.W.; Taylor, J.B.

    1987-01-01

    The recently reported observation of the bootstrap current in a tokamak plasma highlights the problem of reconciling this neoclassical effect with the anomalous (i.e., non-neoclassical) electron thermal transport. This Comment reviews the bootstrap current and considers the implications of a self-consistent modification of neoclassical theory based on an enhanced electron-electron interaction. (author)

  16. High power RF heating and nonthermal distributions in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Peeters, A.G.

    1994-12-13

    This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL).

  17. High power RF heating and nonthermal distributions in tokamak plasmas

    International Nuclear Information System (INIS)

    Peeters, A.G.

    1994-01-01

    This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL)

  18. Kinetic theory of plasma adiabatic major radius compression in tokamaks

    International Nuclear Information System (INIS)

    Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.

    1998-01-01

    In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics

  19. Neoclassical Physics for Current Drive in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Duthoit, F.X.

    2012-03-01

    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)

  20. Ignition and burn control in tokamak plasmas

    International Nuclear Information System (INIS)

    Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.

    1981-01-01

    Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)

  1. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    Energy Technology Data Exchange (ETDEWEB)

    Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

    2016-10-15

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  2. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    International Nuclear Information System (INIS)

    Mitrishkin, Yuri V.; Pavlova, Evgeniia A.; Kuznetsov, Evgenii A.; Gaydamaka, Kirill I.

    2016-01-01

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  3. Magnetic fluctuations in the plasma of KT-5C tokamak

    International Nuclear Information System (INIS)

    Lu Ronghua; Pan Gesheng; Wang Zhijiang; Wen Yizhi; Yu Changxuan; Wan Shude; Liu Wandong; Wang Jun; Xu Min; Xiao Delong; Yu Yi

    2004-01-01

    A newly developed moveable magnetic probe array was installed on KT-5C tokamak. The profiles of radial and poloidal magnetic fluctuations of the plasma have been measured for (0.5r/a1.1). The experimental results indicate that there is a radial gradient which is greater than relative electrostatic fluctuations and the magnetic fluctuations contribute a little to losses. A strong coherence between fluctuations of 4 mm nearby two points suggests that the magnetic fluctuations have quite a long correlation length

  4. Electron Heating of LHCD Plasma in HT-7 Tokamak

    International Nuclear Information System (INIS)

    Ding Yonghua; Wan Baonian; Lin Shiyao; Chen Zhongyong; Hu Xiwei; Shi Yuejiang; Hu Liqun; Kong Wei; Zhang Xiaoqing

    2006-01-01

    Electron heating via lower hybrid current drive (LHCD) has been investigated in HT-7 superconducting tokamak. Experiments show that the central electron temperature T e0 , the volume averaged electron temperature e > and the peaking factor of the electron temperature Q Te = T e0 / e > increase with the lower hybrid wave (LHW) power. Simultaneously the electron heating efficiency and the electron temperature as the function of the central line-averaged electron density (n e ) and the plasma current (I p ) have also been investigated. The experimental results are in a good agreement with those of the classical collision theory and the LHW power deposition theory

  5. Direction of Impurity Pinch and Auxiliary Heating in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Angioni, C.; Peeters, A.G.

    2006-01-01

    A mechanism of particle pinch for trace impurities in tokamak plasmas, arising from the effect of parallel velocity fluctuations in the presence of a turbulent electrostatic potential, is identified analytically by means of a reduced fluid model and verified numerically with a gyrokinetic code for the first time. The direction of such a pinch reverses as a function of the direction of rotation of the turbulence in agreement with the impurity pinch reversal observed in some experiments when moving from dominant auxiliary ion heating to dominant auxiliary electron heating

  6. 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.)

  7. A theory of the coherent fundamental plasma emission in Tokamaks

    International Nuclear Information System (INIS)

    Alves, M.V.; Chian, A.C.-L.

    1987-01-01

    A theoretical model of coherent radiation near the fundamental plasma frequency in tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be parametrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: travelling wave pump and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electromagnetic decay instabilities (L-> T + S), electromagnetic fusion instabilities (L + S -> T) and electromagnetic oscillating two-stream instabilities (L -> T+- S * , where S * is a purely growing mode). (author) [pt

  8. Information content of transient synchrotron radiation in tokamak plasmas

    International Nuclear Information System (INIS)

    Fisch, N.J.; Kritz, A.H.

    1989-04-01

    A brief, deliberate, perturbation of hot tokamak electrons produces a transient, synchrotron radiation signal, in frequency-time space, with impressive informative potential on plasma parameters; for example, the dc toroidal electric field, not available by other means, may be measurably. Very fast algorithms have been developed, making tractable a statistical analysis that compares essentially all parameter sets that might possibly explain the transient signal. By simulating data numerically, we can estimate the informative worth of data prior to obtaining it. 20 refs., 2 figs

  9. Power supplies for plasma column control in COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Havlíček, Josef; Hauptmann, R.; Peroutka, Oldřich; Tadros, Momtaz; Hron, Martin; Janky, Filip; Vondráček, Petr; Cahyna, Pavel; Mikulín, Ondřej; Šesták, David; Junek, Pavel; Pánek, Radomír

    2013-01-01

    Roč. 88, 9-10 (2013), s. 1640-1645 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * Power supplies * Feedback control * Vertical displacement * Vertical kicks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001543#

  10. A theory of the coherent fundamental plasma emission in Tokamaks

    International Nuclear Information System (INIS)

    Alves, M.V.; Chian, A.C.-L.

    1987-07-01

    A theoretical model of coherent radiation near the fundamental plasma frequency in Tokamaks is proposed. It is shown that, in the presence of runaway electrons, the beam-generated Langmuir waves (L) can be paarmetrically converted into electromagnetic waves (T) through ponderomotive coupling to ion acoustic waves (S). Two types of pumps are considered: traveling wave and standing wave pump. Expressions are derived for the excitation conditions and the growth rates of electomagnetic decay instabilities (L → T + S), electromagnetic fusion instabilities (L + S → T) and electromagnetic oscillating two-stream instabilities (L → T+-S sup(*) is a purely growing mode). (author) [pt

  11. Magnetohydrodynamic stability of tokamak plasmas with poloidal mode coupling

    International Nuclear Information System (INIS)

    Shigueoka, H.; Sakanaka, P.H.

    1987-01-01

    The stability behavior with respect to internal modes is examined for a class of tokamak equilibria with non-circular cross sections. The surfaces of the constant poloidal magnetic flux ψ (R,Z) are obtained numerically by solving the Grad-Shafranov's equation with a specified shape for the outmost plasma surface. The equation of motion for ideal MHD stability is written in a ortogonal coordinate system (ψ, χ, φ). Th e stability analysis is performance numerically in a truncated set of coupled m (poloidal wave number) equations. The calculations involve no approximations, and so all parameters of the equilibrium solution can be arbitrarily varied. (author) [pt

  12. A computational model for the confinement and performance of circular and D-shaped Tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolai, A.; Boerner, P.

    1987-10-01

    A combined one-dimensional and two-dimensional description of toroidal and axisymmetric plasmas is presented which is based essentially on an equilibrium solver resorting to the fast Buneman invertor and two one-dimensional transport codes describing the protium, deuterium, tritium, and plasma energy inventory and accounting for three impurity species; it is employed to compute the time evolution of Tokamak plasmas. The attempt was made to achieve a consistent modelling of the transport and equilibrium phenomena in a plasma which interacts with the peripheral devices for e.g. confinement, plasma heating and limitation of the plasma aperture. The equilibrium solver is connected to a coil submodule computing the poloidal field coil currents maintaining the designed plasma shape approximately. A surface current density standing for the magnetization of the iron core and the yokes is calculated by means of the module for the transformer iron. This module is linked to the equilibrium solver as well so that consistency between the coil currents, the plasma current distribution and the magnetization of the transformer iron is achieved. (orig./GG)

  13. A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

    International Nuclear Information System (INIS)

    Takase, Haruhiko; Senda, Ikuo

    1999-01-01

    A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

  14. Calculation of transport coefficients in an axisymmetric plasma

    International Nuclear Information System (INIS)

    Shumaker, D.E.

    1977-01-01

    A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount

  15. Investigation of radial propagation of electrostatic fluctuations in the IR-T1 tokamak plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Shariatzadeh, R; Ghoranneviss, M; Salem, M K [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University (IAU), PO Box 14665-678, Tehran (Iran, Islamic Republic of); Emami, M, E-mail: rezashariatzadeh@gmail.com [Laser and Optics Research School, NSTRI, AEOI, PO Box 14155-1339, Tehran (Iran, Islamic Republic of)

    2011-01-15

    The radial propagation of electrostatic fluctuation is considered extremely important for understanding cross-field anomalous transport. In this paper, two arrays of Langmuir probes are used to analyze electrostatic fluctuations in the edge of IR-T1 tokamak plasma in both the radial and the poloidal directions. The propagation characteristics of the floating potential fluctuations are analyzed by the two-point correlation technique. The wavenumber spectrum shows that there is a net radially outward propagation of turbulent fluctuations in the edge and scrape-off layer (SOL) regions. Hence, edge turbulence presumably originates from core fluctuations.

  16. Investigation of radial propagation of electrostatic fluctuations in the IR-T1 tokamak plasma edge

    International Nuclear Information System (INIS)

    Shariatzadeh, R; Ghoranneviss, M; Salem, M K; Emami, M

    2011-01-01

    The radial propagation of electrostatic fluctuation is considered extremely important for understanding cross-field anomalous transport. In this paper, two arrays of Langmuir probes are used to analyze electrostatic fluctuations in the edge of IR-T1 tokamak plasma in both the radial and the poloidal directions. The propagation characteristics of the floating potential fluctuations are analyzed by the two-point correlation technique. The wavenumber spectrum shows that there is a net radially outward propagation of turbulent fluctuations in the edge and scrape-off layer (SOL) regions. Hence, edge turbulence presumably originates from core fluctuations.

  17. The effect of ion drifts on the properties of the tokamak scrape-off plasma

    International Nuclear Information System (INIS)

    Petravic, M.; Kuo-Petravic, G.

    1988-09-01

    A plasma fluid model which takes into account ion drifts has been constructed and applied to the scrape-off layer of a tokamak with a poloidal divertor. This model predicts near-sonic toroidal velocities and large poloidal flows in most of the scrapeoff together with steep gradients in the pressure and electrostatic potential along the magnetic field near the X-point, contrary to the predictions of the standard model. The potential step at X-point should reduce parallel heat transport and could act as an H-mode trigger. 12 refs., 4 figs

  18. Aspects of the diffusion of electrons and ions in Tokamak plasma

    International Nuclear Information System (INIS)

    Negrea, M.; Petrisor, I.; Constantinescu, Dana

    2010-01-01

    Two distinct problems concerning the anomalous transport in Tokamak plasma were analysed. The first one is related to the diffusion of ions in a stochastic magnetic field with curvature starting from Langevin equations of the guiding centre approximation. We analysed the influence of the drift Kubo number, the magnetic Kubo number and of the anisotropy on the diffusion of ions. The second problem is related to the diffusion of electrons in a combination of an electrostatic stochastic field with an unperturbed sheared magnetic field in slab geometry. The global effects of the parameters on the diffusion tensor components are exhibited.(authors)

  19. New DIII-D tokamak plasma control system

    International Nuclear Information System (INIS)

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

  20. Progress towards internal transport barriers at high plasma density sustained by pure electron heating and current drive in the FTU tokamak

    International Nuclear Information System (INIS)

    Pericoli Ridolfini, V.; Barbato, E.; Buratti, P.

    2003-01-01

    Strong electron Internal Transport Barriers (ITBs) are obtained in FTU by the combined injection of Lower Hybrid (LH, up to 1.9 MW) and Electron Cyclotron (EC up to 0.8 MW) radio frequency waves. ITBs occur during either the current plateau or the ramp up phase, and both in full and partial current drive (CD) regimes, up to peak densities n e0 >1.2·10 20 m -3 , relevant to ITER operation. Central electron temperatures T e0 >11 keV, at n e0 ∼0.8·10 20 m -3 are sustained longer than 6 confinement times. The ITB extends over a region where a slightly reversed magnetic shear is established by off-axis LHCD and can be as wide as r/a=0.5. The EC power, instead, is used either to benefit from this improved confinement by heating inside the ITB, or to enhance the peripheral LH power deposition and CD with off axis resonance. Collisional ion heating is also observed, but thermal equilibrium with the electrons cannot be attained since the e-i equipartition time is always 4-5 times longer than the energy confinement time. The transport analysis performed with both ASTRA and JETTO codes shows a very good relation between the foot of the barrier and the weak/reversed shear region, which in turn depends on the LH deposition profile. The Bohm-gyroBohm model accounts for the electron transport until T e0 <6 keV, but is pessimistic at higher temperatures, where often also a reduction in the ion thermal conductivity is observed, provided any magneto hydrodynamic activity is suppressed. (author)

  1. Transport barriers in bootstrap-driven tokamaks

    Science.gov (United States)

    Staebler, G. M.; Garofalo, A. M.; Pan, C.; McClenaghan, J.; Van Zeeland, M. A.; Lao, L. L.

    2018-05-01

    Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is caused by the suppression of turbulence primarily from the large Shafranov shift. It is shown that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift-driven barrier formation. Two self-organized states of the internal and edge transport barrier are observed. It is shown that these two states are controlled by the interaction of the bootstrap current with magnetic shear, and the kinetic ballooning mode instability boundary. Election scale energy transport is predicted to be dominant in the inner 60% of the profile. Evidence is presented that energetic particle-driven instabilities could be playing a role in the thermal energy transport in this region.

  2. Electron and current density measurements on tokamak plasmas

    International Nuclear Information System (INIS)

    Lammeren, A.C.A.P. van.

    1991-01-01

    The first part of this thesis describes the Thomson-scattering diagnostic as it was present at the TORTUR tokamak. For the first time with this diagnostic a complete tangential scattering spectrum was recorded during one single laser pulse. From this scattering spectrum the local current density was derived. Small deviations from the expected gaussian scattering spectrum were observed indicating the non-Maxwellian character of the electron-velocity distribution. The second part of this thesis describes the multi-channel interferometer/ polarimeter diagnostic which was constructed, build and operated on the Rijnhuizen Tokamak Project (RTP) tokamak. The diagnostic was operated routinely, yielding the development of the density profiles for every discharge. When ECRH (Electron Cyclotron Resonance Heating) is switched on the density profile broadens, the central density decreases and the total density increases, the opposite takes place when ECRH is switched off. The influence of MHD (magnetohydrodynamics) activity on the density was clearly observable. In the central region of the plasma it was measured that in hydrogen discharges the so-called sawtooth collapse is preceded by an m=1 instability which grows rapidly. An increase in radius of this m=1 mode of 1.5 cm just before the crash is observed. In hydrogen discharges the sawtooth induced density pulse shows an asymmetry for the high- and low-field side propagation. This asymmetry disappeared for helium discharges. From the location of the maximum density variations during an m=2 mode the position of the q=2 surface is derived. The density profiles are measured during the energy quench phase of a plasma disruption. A fast flattening and broadening of the density profile is observed. (author). 95 refs.; 66 figs.; 7 tabs

  3. Impurity injection into tokamak plasmas by erosion probes

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Bakos, J.S.; Buerger, G.; Paszti, F.; Petravich, G.

    1987-08-01

    Exposing special erosion probes into the edge plasma of MT-1 the impurities Li and Ti were released and contaminated the plasma. By the use of collector probes the torodial transport of these impurities were investigated. The results indicate a preferential impurity flow into codirection of the plasma current. However, the asymmetric component of this flow is much larger than expected from the toroidal drift correlated to the plasma current. (author)

  4. Observations of giant recombination edges on PLT tokamak induced by particle transport

    International Nuclear Information System (INIS)

    Brau, K.; von Goeler, S.; Bitter, M.; Cowan, R.D.; Eames, D.; Hill, K.; Sauthoff, N.; Silver, E.; Stodiek, W.

    1980-03-01

    Characteristic steps in the continum spectrum of high temperature tokamak plasmas associated with recombination radiation from impurity ions were observed. During special argon-seeded discharges on the Princeton Large Torus (PLT) tokamak the x-ray spectrum exhibited large enhancements over the bremsstrahlung continuum beginning with energies of 4.1 keV. This corresponds to the radiative capture of free electrons by hydrogen-like argon into the ground state of helium-like argon. A simple particle diffusion model is proposed, with the Ar XVIII radial profiles evaluated from the size of the recombination edges. For the case of moderate density ( approx. 3 x 10 13 cm -3 ) and temperature [T/sub e/(0) approx. 1.5 keV] discharges the outward radial transport velocity is found to be approximately 10 m/sec

  5. The generalized Balescu-Lenard collision operator: A unifying concept for tokamak transport

    International Nuclear Information System (INIS)

    Mynick, H.E.

    1987-08-01

    The generalization of the Balescu-Lenard collision operator to its fully electromagnetic counterpart in Kaufman's action-angle formalism is derived and its properties investigated. The general form may be specialized to any particular geometry where the unperturbed particle motion is integrable, and thus includes cylindrical plasmas, inhomogeneous slabs with nonuniform magnetic fields, tokamaks, and the particularly simple geometry of the standard operator as special cases. The general form points to the commonality between axisymmetric, turbulent, and ripple transport, and implies properties (e.g., intrinsic ambipolarity) which should be shared by them, under appropriate conditions. Along with a turbulent ''anomalous diffusion coefficient'' calculated for tokamaks in previous work, an ''anomalous pinch'' term of closely related structure and scaling is also implied by the generalized operator. 20 refs

  6. Static and dynamic control of plasma equilibrium in a Tokamak

    International Nuclear Information System (INIS)

    Blum, J.; Dei Cas, R.

    1979-01-01

    We are dealing here with the problem of controlling the plasma boundary and its displacements. Static control consists in determining the currents in the external coils of the Tokamak so that the plasma boundary has certain fixed characteristics: radial position, vertical elongation, desired shape. A self-consistent method is proposed here, considering a free plasma boundary, and using the techniques of optimal control of distributed parameter systems to solve the problem. The dynamic control problem considered in the second part of the paper is the control of the plasma radial displacements. An elaborate system of preprogramming and feedback control has been developed to ensure equilibrium and stability of the horizontal plasma motions. Optimal control techniques have been used to calculate the optimal primary coils configuration, the preprogramming voltages and the feedback gains. A new stability diagrams has been obtained which takes into account the erosion of the plasma by the limiter. All these calculations have been applied successfully to TFR 600 where thin liner and the presence of an iron core make the problem of stabilization of the radial displacements very difficult

  7. Effect of neutral atoms on tokamak edge plasmas

    International Nuclear Information System (INIS)

    Fueloep, T.; Catto, Peter J.; Helander, P.

    2001-01-01

    Neutral atoms can significantly influence the physics of tokamak edge plasmas, e.g., by affecting the radial electric field and plasma flow there, which may, in turn, be important for plasma confinement. Earlier work [Fueloep et al., Phys. Plasmas 5, 3969 (1998)], assuming short mean-free path neutrals and Pfirsch-Schlueter ions, has shown that the ion-neutral coupling through charge-exchange affects the neoclassical flow velocity significantly. However, the mean-free path of the neutrals is not always small in comparison with the radial scale length of densities and temperatures in the edge pedestal. It is therefore desirable to determine what happens in the limit when the neutral mean-free path is comparable with the scale length. In the present work a self-similar solution for the neutral distribution function allowing for strong temperature and density variation is used, following the analysis of Helander and Krasheninnikov [Phys. Plasmas 3, 226 (1995)]. The self-similar solution is possible if the ratio of the mean-free path to the temperature and density scale length is constant throughout the edge plasma. The resulting neutral distribution function is used to investigate the neutral effects on the ion flow and electrostatic potential as this ratio varies from much less than one to order unity

  8. Characteristics of equilibrium and perturbed transport coefficients in tokamaks

    International Nuclear Information System (INIS)

    Gentle, K.W.

    1995-01-01

    Although the evolution of a perturbation to a tokamak equilibrium can generally be described by local transport coefficients modestly enhanced above the equilibrium values, there are some significant cases for which this is inadequate. The density profile evolution in ASDEX-U occurs far more rapidly than is consistent with reasonable particle confinement times, and the evolution of cold pulses in TEXT requires nonlocal behavior in the core and some kind of anomaly near the periphery. The experiments are suggesting effects beyond standard local turbulent transport models. (orig.)

  9. A transmission/escape probabilities model for neutral particle transport in the outer regions of a diverted tokamak

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1992-12-01

    A new computational model for neutral particle transport in the outer regions of a diverted tokamak plasma chamber is presented. The model is based on the calculation of transmission and escape probabilities using first-flight integral transport theory and the balancing of fluxes across the surfaces bounding the various regions. The geometrical complexity of the problem is included in precomputed probabilities which depend only on the mean free path of the region

  10. Heavy Neutral Beam Probe for edge plasma analysis in Tokamaks

    International Nuclear Information System (INIS)

    Castracane, J.; Saravia, E.; Beckstead, J.; Aceto, S.

    1993-01-01

    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

  11. 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.)

  12. Nonlinear neoclassical transport in toroidal edge plasmas

    International Nuclear Information System (INIS)

    Fueloep, T.

    2002-01-01

    Edge plasma processes play a critical role for the global confinement of the plasma in a tokamak. In the edge region, where impurity ions are abundant and the temperature and density gradients are large, the assumptions of standard neoclassical theory break down. This paper reviews recent extensions of neoclassical theory to treat this problem, in particular our own work, which focuses on the nonlinear aspects of transport in a plasma with heavy impurity ions. In this theory, the pressure and temperature gradients are allowed to be steeper than in conventional theory neoclassical theory, so that the friction force between the bulk ions and heavy impurities is comparable to the parallel impurity pressure gradient. The impurity ions are then found to undergo a spontaneous rearrangement on each flux surface. This reduces their parallel friction with the bulk ions and causes the neoclassical ion flux to become a non-monotonic function of the gradients for plasma parameters typical of the tokamak edge. Thus, the neoclassical confinement is improved in regions where the gradients are large, such as in the edge pedestal. (orig.)

  13. Results from transient transport experiments in Rijnhuizen tokamak project: Heat convection, transport barriers and 'non-local' effects

    International Nuclear Information System (INIS)

    Mantica, P.; Gorini, G.; Hogeweij, G.M.D.; Kloe, J. de; Lopez Cardozo, N.J.; Schilham, A.M.R.

    2001-01-01

    An overview of experimental transport studies performed on the Rijnhuizen Tokamak Project (RTP) using transient transport techniques in both Ohmic and ECH dominated plasmas is presented. Modulated Electron Cyclotron Heating (ECH) and oblique pellet injection (OPI) have been used to induce electron temperature (T e ) perturbations at different radial locations. These were used to probe the electron transport barriers observed near low order rational magnetic surfaces in ECH dominated steady-state RTP plasmas. Layers of inward electron heat convection in off-axis ECH plasmas were detected with modulated ECH. This suggests that RTP electron transport barriers consist of heat pinch layers rather than layers of low thermal diffusivity. In a different set of experiments, OPI triggered a transient rise of the core T e due to an increase of the T e gradient in the 1< q<2 region. These transient transport barriers were probed with modulated ECH and found to be due to a transient drop of the electron heat diffusivity, except for off-axis ECH plasmas, where a transient inward pinch is also observed. Transient transport studies in RTP could not solve this puzzling interplay between heat diffusion and convection in determining an electron transport barrier. They nevertheless provided challenging experimental evidence both for theoretical modelling and for future experiments. (author)

  14. Development of Tokamak experiment technology - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hang Young; Lee, Soon Chil; Kwon, Gi Chung; Seo, Sung Hun; Jeon, Sang Jin; Heo, Sung Hee; Heo, Eun Gi; Lee, Dae Hang; Lee, Chan Hee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-08-01

    Research objectives are to design and fabricate antenna, measure the property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling. Main heating method is ohmic heating at the KAIST tokamak. So, the plasma current produced is more than 30 kA and, the loop voltage of the plasma is 2 {approx} 3V. The power of the plasma by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, it is needed RF system with more than 100 kW in 7 {approx} 15 MHz. In the first year a RF amplifier with 1 kW in 300 khz {approx} 35 MHz was bought. The manufacture of ICRF system will start from next years. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. Experimentally, we obtain the results that the signal of piezo element is proportional to the square of electric field. In the next year, we will research the type of antenna subsequently. 28 refs., 3 tabs., 18 figs. (author)

  15. Remote network control plasma diagnostic system for Tokamak T-10

    International Nuclear Information System (INIS)

    Troynov, V I; Zimin, A M; Krupin, V A; Notkin, G E; Nurgaliev, M R

    2016-01-01

    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. (paper)

  16. Characterisation of detached plasmas on the MAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, J.R., E-mail: james.harrison@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Lisgo, S.W. [ITER Organization, Route de Vinon-sur-Verdon, St.Paul-lez-Durance, Cedex (France); Gibson, K.J. [Department of Physics, University of York, Heslington, York, YO10 5DD (United Kingdom); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Dowling, J. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

    2011-08-01

    Divertor detachment is an attractive operating regime for the next generation of tokamak devices, as it offers a means of mitigating the steady-state heat flux to plasma facing components. In order to clarify the dominant physical mechanisms that govern detachment, high quality data from several diagnostics are required to constrain theoretical models. To that end, high spatial ({approx}3 mm) and temporal (5 kHz) resolution measurements have been made of the intensity of deuterium Balmer and carbon emission lines during the onset and evolution of detachment of the lower inner strike point in MAST L-mode discharges. Furthermore, spatially-resolved measurements of the shapes and intensities of high-n Balmer lines have been recorded to infer plasma conditions during the detached phase.

  17. Heat flow during sawtooth collapse in tokamak plasmas

    International Nuclear Information System (INIS)

    Hanada, Kazuaki

    1994-01-01

    Heat flow during sawtooth collapse was studied on the WT-3 tokamak by using temporal evolution of soft X-ray intensity profile in the poloidal cross section in a lower hybrid current driven plasma as well as an electron cyclotron heated plasma. Two phase in sawtooth collapses were observed. In the first phases, the hottest spot that is the peak of the soft X-ray distribution approaches the inversion surface and heat flows out through a narrow gate on the inversion surface. In the second phase, the hottest spot stays on the inversion surface, and heat flows out through the whole inversion surface. This suggests that magnetic reconnection as predicted by Kadomtsev's model occurs in the first phase, but in the second phase, a different mechanism dominates heat flow. (author)

  18. Study of the plasma edge turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Laurent, L.; Mourgues, F.; Roubin, J.P.; Samain, A.

    1990-01-01

    The plasma edge in tokamaks is known to be very turbulent. We investigate here the non linear stability of a test mode in presence of an helical potential perturbation, i.e. a pump mode, which simulates the plasma turbulence. The particle trajectories in this perturbed equilibrium are derived using an hamiltonian formalism. The electrons appear to have trapped trajectories in the potential well of the pump mode, while the ions experience a large convective motion. These two effects have a large influence on the test mode stability. First, non linearly trapped electrons supply an energy source for the test mode. Second, the ion convective motion introduces a radial scale of the test mode larger than the ion Larmor radius, in agreement with experimental data. These two phenomena allow a bifurcation in the turbulence level and provide therefore an explanation for the L-H transition

  19. Poloidal asymmetries in the scrape-off layer plasma of the Alcator C tokamak

    International Nuclear Information System (INIS)

    LaBombard, B.; Lipschultz, B.

    1987-01-01

    Large poloidal asymmetries in density, electron temperature, radial density e-folding length and floating potential have been measured in the plasma existing between the limiter radius and the wall of the Alcator C tokamak. Typically, variations in density by factors of about 4-20 and variations in radial density e-folding length by factors of about 3-8 are recorded in discharges which are bounded by poloidally symmetric ring limiters. These poloidal asymmetries show that pressure is a function of poloidal angle on open magnetic flux surfaces in this region of the plasma. Observations of toroidally symmetric MARFE (multifaceted asymmetric radiation from the edge) phenomena further imply that density and perhaps pressure are also a function of poloidal angle on closed flux surfaces existing just inside the limiter radius. The magnitude of these poloidal asymmetries and their dependence on poloidal angle persists independent of machine parameters (central plasma density, plasma current, toroidal field, MARFE versus non-MARFE discharges). Analysis of the data indicates that these asymmetries are caused by poloidal variations in perpendicular particle and heat transport in both the main plasma and the scrape-off layer. A number of possible asymmetric perpendicular transport processes in the scrape-off layer plasma are examined, including diffusion and E-vectorxB-vector plasma convection. (author)

  20. Energy transport requirements for tokamak reactors in the second ballooning stability regime

    International Nuclear Information System (INIS)

    Potok, R.E.; Bromberg, L.; Cohn, D.R.

    1986-01-01

    The authors present an analysis of ignition confinement constraints on a tokamak reactor operating in the second regime of ballooning stability. This regime is characterized by flat plasma pressure profiles, with a sharp pressure gradient near a conducting first wall at the plasma edge. The energy confinement time is determined by transport processes across the pressure gradient region. The authors have found that the required transport needed for ignition in the edge region is very close to the value predicted by neoclassical ion conductivity scaling. Only by carefully tailoring the conductivity scaling across the flux coordinate were the authors able to match both the kink stability and ignition requirements. With optimistic assumptions, R/sub o/ ≅ 7 m appears to be the minimum major radius for an economical tokamak reactor in the second ballooning stability regime. This paper presents a base design case at R/sub o/ = 7 m, and shows how the reactor design varies with changes in major radius, ion transport scaling, and electron transport scaling