WorldWideScience

Sample records for tokamak plasmas invited

  1. 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

  2. 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)

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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).

  11. 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.)

  12. 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

  13. 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.)

  14. 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)

  15. 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

  16. 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 ...

  17. 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

  18. 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)

  19. 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

  20. 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

  1. Summary of the 1982 small tokamak users meeting

    International Nuclear Information System (INIS)

    Sprott, J.C.

    1982-11-01

    On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. Investigating fusion plasma instabilities in the Mega Amp Spherical Tokamak using mega electron volt proton emissions (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Perez, R. V., E-mail: rvale006@fiu.edu; Boeglin, W. U.; Angulo, A.; Avila, P.; Leon, O.; Lopez, C. [Department of Physics, Florida International University, 11200 SW 8 ST, CP204, Miami, Florida 33199 (United States); Darrow, D. S. [Princeton Plasma Physics Laboratory, James Forrestal Campus, P.O. Box 451, Princeton, New Jersey 08543 (United States); Cecconello, M.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, Uppsala SE-751 20 (Sweden); Allan, S. Y.; Akers, R. J.; Keeling, D. L.; McClements, K. G.; Scannell, R.; Conway, N. J. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrasse 2, D-85748, Garching (Germany); Jones, O. M. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Michael, C. A. [Australian National University, Canberra ACT 0200 (Australia)

    2014-11-15

    The proton detector (PD) measures 3 MeV proton yield distributions from deuterium-deuterium fusion reactions within the Mega Amp Spherical Tokamak (MAST). The PD’s compact four-channel system of collimated and individually oriented silicon detectors probes different regions of the plasma, detecting protons (with gyro radii large enough to be unconfined) leaving the plasma on curved trajectories during neutral beam injection. From first PD data obtained during plasma operation in 2013, proton production rates (up to several hundred kHz and 1 ms time resolution) during sawtooth events were compared to the corresponding MAST neutron camera data. Fitted proton emission profiles in the poloidal plane demonstrate the capabilities of this new system.

  9. Dynamics and feedback control of plasma equilibrium position in a tokamak

    International Nuclear Information System (INIS)

    Burenko, O.

    1983-01-01

    A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems

  10. 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].

  11. 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)

  12. 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.

  13. Time-dependent analysis of the resistivity of post-disruption tokamak plasmas

    International Nuclear Information System (INIS)

    Bakhtiari, M.; Whyte, D. G.

    2006-01-01

    The effect of neutrals on plasma resistivity due to electron-neutral collisions is studied with respect to its effect on tokamak disruptions. The resistivity of the tokamak plasma after the thermal quench is critical in determining the current quench rate, the plasma temperature, and runaway electron generation in tokamaks through the electric field, all features which are important for mitigating the damaging effect of disruptions. It is shown that the plasma resistivity during tokamak disruptions is a time-dependent parameter which may vary with disruption time scales due to the increasing fraction of neutrals. However the effect of neutrals on resistivity is found to be small for the expected neutral fraction, mostly due to power balance considerations between radiation and Ohmic heating in the plasma

  14. 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

  15. 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.)

  16. 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)

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

  2. 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

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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.)

  8. 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)

  9. 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.

  10. 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)

  11. 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)

  12. 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.

  13. 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.

  14. Control of plasma poloidal shape and position in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

  15. 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

  16. 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

  17. 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

  18. 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

  19. Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Filipe F.P.W.; Bosco, Edson del

    1994-12-31

    This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.

  20. 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.

  1. 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)

  2. Impurity screening in high density plasmas in tokamaks with a limiter configuration

    International Nuclear Information System (INIS)

    Ferro, C.; Zanino, R.

    1992-01-01

    Impurity screening in high density plasmas in tokamaks with a limiter configuration is investigated by means of a simple semi-analytical model. An iterative scheme is devised, in order to determine self-consistently the values of scrape-off layer thickness, edge electron density and temperature, and main plasma contamination parameter Z eff , as a function of given average electron density and temperature in the main plasma and given input power. The model is applied to the poloidal limiter case of the Frascati Tokamak Upgrade, and results are compared with experimental data. A reasonable agreement between the trends is found, emphasizing the importance of a high edge plasma density for obtaining a clean main plasma in limiter tokamaks. (orig.)

  3. 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

  4. 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

  5. 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

  6. The ‘churning mode’ of plasma convection in the tokamak divertor region

    International Nuclear Information System (INIS)

    Ryutov, D D; Cohen, R H; Farmer, W A; Rognlien, T D; Umansky, M V

    2014-01-01

    The churning mode can arise in a toroidally-symmetric plasma where it causes convection in the vicinity of the poloidal magnetic field null. The mode is driven by the toroidal curvature of magnetic field lines coupled with a pressure gradient. The toroidal equilibrium conditions cannot be satisfied easily in the virtual absence of the poloidal field (PF)—hence the onset of this mode, which ‘churns’ the plasma around the PF null without perturbing the strong toroidal field. We find the conditions under which this mode can be excited in magnetic configurations with first-, second-, and third-order PF nulls (i.e., in the geometry of standard, snowflake and cloverleaf divertors). The size of the affected zone in second- and third-order-null divertors is much larger than in a standard first-order-null divertor. The proposed phenomenological theory allows one to evaluate observable characteristics of the mode, in particular the frequency and amplitude of the PF perturbations. The mode spreads the tokamak heat exhaust between multiple divertor legs and may lead to a broadening of the plasma width in each leg. The mode causes much more intense plasma convection in the poloidal plane than the classical plasma drifts. (invited comment)

  7. 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

  8. Study of Globus-M Tokamak Poloidal System and Plasma Position Control

    Science.gov (United States)

    Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

    2017-12-01

    In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

  9. Impedance of an intense plasma-cathode electron source for tokamak startup

    Science.gov (United States)

    Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

    2016-05-01

    An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

  10. 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

  11. Study of intelligent system for control of the tokamak-ETE plasma positioning

    International Nuclear Information System (INIS)

    Barbosa, Luis Filipe de Faria Pereira Wiltgen

    2003-01-01

    The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)

  12. 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)

  13. 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

  14. 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.

  15. 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%.

  16. 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%

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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.)

  3. 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.)

  4. 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

  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. Total hydrogen and oxygen fluxes in the edge plasma of tokamaks

    International Nuclear Information System (INIS)

    Kastelewicz, H.

    1988-01-01

    A relativistic model of the edge plasma of tokamaks is described considering the primary neutral fluxes emitted from limiter and wall. The primary neutrals, which determine essentially the particle flux balance in the plasma edge, the scrape-off layer plasma and the particles adsorbed at limiter and wall are treated as separate subsystems which are iteratively coupled through the mutual particle sinks and sources. The model is used for the calculation of total hydrogen and oxygen fluxes in edge plasma of tokamaks. The results for different fractions of and contributions to the total fluxes are illustrated and discussed

  7. DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section

    International Nuclear Information System (INIS)

    Amrollahi, R.

    1997-01-01

    The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)

  8. Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak

    International Nuclear Information System (INIS)

    Barbosa, Luis Filipe F.P.W.; Bosco, Edson del.

    1994-01-01

    This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs

  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. Control of tokamak plasma current and equilibrium with hybrid poloidal field coils

    International Nuclear Information System (INIS)

    Shimada, Ryuichi

    1982-01-01

    A control method with hybrid poloidal field system is considered, which comprehensively implements the control of plasma equilibrium and plasma current, those have been treated independently in Tokamak divices. Tokamak equilibrium requires the condition that the magnetic flux function value on plasma surface must be constant. From this, the current to be supplied to each coil is determined. Therefore, each coil current is the resultant of the component related to plasma current excitation and the component required for holding equilibrium. Here, it is intended to show a method by which the current to be supplied to each coil can easily be calculated by the introduction of hybrid control matrix. The text first considers the equilibrium of axi-symmetrical plasma and the equilibrium magnetic field outside plasma, next describes the determination of current using the above hybrid control matrix, and indicates an example of controlling Tokamak plasma current and equilibrium by the hybrid poloidal field coils. It also shows that the excitation of plasma current and the maintenance of plasma equilibrium can basically be available with a single power supply by the appropriate selection of the number of turns of each coil. These considerations determine the basic system configuration as well as decrease the installed capacity of power source for the poloidal field of a Tokamak fusion reactor. Finally, the actual configuration of the power source for hybrid poloidal field coils is shown for the above system. (Wakatsuki, Y.)

  11. 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

  12. Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma

    International Nuclear Information System (INIS)

    Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.

    2000-01-01

    Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)

  13. 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.

  14. 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)

  15. 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/{{\

  16. 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

  17. 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

  18. Bifurcated states of a rotating tokamak plasma in the presence of a static error-field

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    1998-01-01

    The bifurcated states of a rotating tokamak plasma in the presence of a static, resonant, error-field are strongly analogous to the bifurcated states of a conventional induction motor. The two plasma states are the open-quotes unreconnectedclose quotes state, in which the plasma rotates and error-field-driven magnetic reconnection is suppressed, and the open-quotes fully reconnectedclose quotes state, in which the plasma rotation at the rational surface is arrested and driven magnetic reconnection proceeds without hindrance. The response regime of a rotating tokamak plasma in the vicinity of the rational surface to a static, resonant, error-field is determined by three parameters: the normalized plasma viscosity, P, the normalized plasma rotation, Q 0 , and the normalized plasma resistivity, R. There are 11 distinguishable response regimes. The extents of these regimes are calculated in P endash Q 0 endash R space. In addition, an expression for the critical error-field amplitude required to trigger a bifurcation from the open-quotes unreconnectedclose quotes to the open-quotes fully reconnectedclose quotes state is obtained in each regime. The appropriate response regime for low-density, ohmically heated, tokamak plasmas is found to be the nonlinear constant-ψ regime for small tokamaks, and the linear constant-ψ regime for large tokamaks. The critical error-field amplitude required to trigger error-field-driven magnetic reconnection in such plasmas is a rapidly decreasing function of machine size, indicating that particular care may be needed to be taken to reduce resonant error-fields in a reactor-sized tokamak. copyright 1998 American Institute of Physics

  19. 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

  20. Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

    International Nuclear Information System (INIS)

    Caress, R.W.; Mayo, R.M.; Carter, T.A.

    1995-01-01

    Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

  1. 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 ...

  2. Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

    International Nuclear Information System (INIS)

    Wieland, R.M.; Houlberg, W.A.

    1982-02-01

    A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Deposition profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10 13 - 2 x 10 14 cm -3 , and beam orientation ranging from perpendicular to tangential to the inside wall

  3. 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

  4. 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

  5. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  6. 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

  7. Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

    2015-01-01

    Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

  8. 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.

  9. Determination of plasma column transverse section in the TBR-1 tokamak

    International Nuclear Information System (INIS)

    Conde, M.E.

    1986-01-01

    The temporal evolution of plasma column transverse section in the TBR-1 tokamak is determined. The experimental melhod is based on the simulation of toroidal current distribution in plasma by a set of toroidal filaments. The currents in these filaments are determined by minimization of square error between the magnetic field produced by filaments and the field measured into the tokamak vacuum vessel. For the measurement of magnetic field, twenty small magnetic coils were constructed and installated in the region protected by current limiters. The plasma column transverse cross section is determined by poloidal field produced by the currents in filaments. The multipole moments of plasma current distribution and the Λ Shafranov parameter were obtained. (M.C.K.) [pt

  10. 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

  11. Robust control design for the plasma horizontal position control on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

    2013-01-01

    It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

  12. Comparison of bootstrap current and plasma conductivity models applied in a self-consistent equilibrium calculation for Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Maria Celia Ramos; Ludwig, Gerson Otto [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: mcr@plasma.inpe.br

    2004-07-01

    Different bootstrap current formulations are implemented in a self-consistent equilibrium calculation obtained from a direct variational technique in fixed boundary tokamak plasmas. The total plasma current profile is supposed to have contributions of the diamagnetic, Pfirsch-Schlueter, and the neoclassical Ohmic and bootstrap currents. The Ohmic component is calculated in terms of the neoclassical conductivity, compared here among different expressions, and the loop voltage determined consistently in order to give the prescribed value of the total plasma current. A comparison among several bootstrap current models for different viscosity coefficient calculations and distinct forms for the Coulomb collision operator is performed for a variety of plasma parameters of the small aspect ratio tokamak ETE (Experimento Tokamak Esferico) at the Associated Plasma Laboratory of INPE, in Brazil. We have performed this comparison for the ETE tokamak so that the differences among all the models reported here, mainly regarding plasma collisionality, can be better illustrated. The dependence of the bootstrap current ratio upon some plasma parameters in the frame of the self-consistent calculation is also analysed. We emphasize in this paper what we call the Hirshman-Sigmar/Shaing model, valid for all collisionality regimes and aspect ratios, and a fitted formulation proposed by Sauter, which has the same range of validity but is faster to compute than the previous one. The advantages or possible limitations of all these different formulations for the bootstrap current estimate are analysed throughout this work. (author)

  13. DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

    2008-01-01

    Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

  14. 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

  15. 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

  16. Models for Predicting Boundary Conditions in L-Mode Tokamak Plasma

    International Nuclear Information System (INIS)

    Siriwitpreecha, A.; Onjun, T.; Suwanna, S.; Poolyarat, N.; Picha, R.

    2009-07-01

    Full text: The models for predicting temperature and density of ions and electrons at boundary conditions in L-mode tokamak plasma are developed using an empirical approach and optimized against the experimental data obtained from the latest public version of the International Pedestal Database (version 3.2). It is assumed that the temperature and density at boundary of L-mode plasma are functions of engineering parameters such as plasma current, toroidal magnetic field, total heating power, line averaged density, hydrogenic particle mass (A H ), major radius, minor radius, and elongation at the separatrix. Multiple regression analysis is carried out for these parameters with 86 data points in L-mode from Aug (61) and JT60U (25). The RMSE of temperature and density at boundary of L-mode plasma are found to be 24.41% and 18.81%, respectively. These boundary models are implemented in BALDUR code, which will be used to simulate the L-mode plasma in the tokamak

  17. 12th Czechoslovak seminar on plasma physics and technology

    International Nuclear Information System (INIS)

    1983-03-01

    The 12th Czechoslovak seminar on plasma physics and technology was oriented mainly to the problems of high-temperature plasmas and controlled thermonuclear fusion. The proceedings contain 27 invited papers and communications presented in three sections: 1) Inertial controlled fusion, 2) Tokamaks, 3) Theory and miscellaneous topics. The first group of papers deals with various problems of electron-beam, ion-beam, and laser fusion, including physical processes in fusion targets. The tokamak section discusses the latest experimental results achieved in the Russian tokamaks FT-2, Tuman 2-a, T-7 and T-10, in the Czechoslovak tokamak TM-1-MH, and in the Hungarian tokamak MT-1. A detailed survey is presented of work on neutral atom injectors in Novosibirsk. In the third section several papers on theoretical studies of nonlinear and turbulent processes in a hot plasma are presented together with a simulation study of a hybrid tokamak reactor. Several contributions on special diagnostic methods are presented. (J.U.)

  18. 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.

  19. Relaxation of plasma potential and poloidal flows in the boundary of tokamak plasmas

    International Nuclear Information System (INIS)

    Hron, M.; Duran, I.; Stoeckel, J.; Hidalgo, C.; Gunn, J.

    2003-01-01

    The relaxation times of plasma parameters after a sudden change of electrode voltage have been measured in the plasma boundary during polarization experiments on the CASTOR tokamak (R = 0.4 m, a = 75 mm, B t = 1 T, I p ∼ 9 kA, q a ∼ 10). The time evolution of the floating potential after the biasing voltage switch-off can be well fitted by an exponential decay with characteristic time in the range of 10 - 20 μs. The poloidal flow shows a transient behaviour with a time scale of about 10 - 30 μs. These time scales are smaller than the expected damping time based on neoclassical parallel viscosity (which is in the range of 100 νs) and atomic physics via charge exchange (in the range of 100 - 1000 νs). But, they are larger than the correlation time of plasma turbulence (about 5 μs). These findings suggest that anomalous damping rate mechanisms for radial electric fields and poloidal flows may play a role in the boundary of tokamak plasmas. (authors)

  20. 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)

  1. 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

  2. IR and FIR laser polarimetry as a diagnostic tool in high-. beta. and Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, D; Machida, M; Scalabrin, A

    1986-03-01

    The change of the polarization state of an electromagnetic wave (EMW) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the EMW allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-pinch plasmas (lambda approx. 10..mu..m). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-pinch/UNICAMP, using lasers developed at UNICAMP.

  3. 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.

  4. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

    1996-01-01

    The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

  5. 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

  6. 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

  7. 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)

  8. A flexible software design to determine the plasma boundary in Damavand tokamak

    International Nuclear Information System (INIS)

    Ghadiri, R.; Esteki, M.H.; Sadeghi, Y.

    2014-01-01

    A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2002-01-01

    In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate

  15. 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 ...

  16. 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

  17. 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)

  18. 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.

  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. 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.)

  1. 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.).

  2. 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)

  3. 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)

  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.

    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)

  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. Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

    Science.gov (United States)

    Fukuda, Takeshi

    The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

  7. Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma

    Science.gov (United States)

    Alipour, Ramin; Ghanbari, Mohamad R.

    2018-04-01

    Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less than them at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level.

  8. 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.

  9. 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

  10. Beat-wave excitation and current driven in tokamak plasma. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, B F [Plasma physics Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Wave heating current drive in tokamaks is a growing subject in the plasma physics literature. For current drive in tokamaks by electromagnetic waves, different methods have been proposed recently. One of the promising schemes for current drive remains the beat wave scheme. This technique employs two CO- or counterpropagating monochromatic laser beams (or microwaves) whose frequency difference matches the plasma frequency, while the wave number difference (or sum, in the case of counterpropagating) determine the wave number of the resulting plasma beat wave. In this work, the basic analysis of a beat wave current drive scheme in which collinear waves are used is discussed. by assuming a Gaussian profile for the amplitude of these pump waves, the amplitudes of the longitudinal and radial fields of the beat wave due to the nonlinear wave interactions have been calculated. Besides, the transfer of momentum flux that accompanies the transfer of wave action in beat-wave scattering will be used to drive the toroidal radial currents in tokamaks. self-generated magnetic fields due to those currents were also calculated. 1 fig.

  11. Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

    Science.gov (United States)

    Lucia, Matthew James

    The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance

  12. 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

  13. 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.

  14. 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.)

  15. Near-wall effects in improved plasma confinement regimes in tokamak FT-2

    International Nuclear Information System (INIS)

    Budnikov, V.N.; D'yachenko, V.V.; Esipov, L.A.

    1997-01-01

    Transition to the regime of improved plasma confinement (H-mode) revealed in experiments on low hybrid heating in tokamak ft-2 is analyzed. Main attention is paid to processes, taking place in near-wall region. The data are correlated with results of experiments in large tokamaks

  16. Relaxation of potential, flows, and density in the edge plasma of Castor tokamak

    International Nuclear Information System (INIS)

    Hron, M.; Weinzettl, V.; Dufkova, E.; Duran, I.; Stoeckel, J.; Hidalgo, C.

    2004-01-01

    Decay times of plasma flows and plasma profiles have been measured after a sudden biasing switch-off in experiments on the Castor tokamak. A biased electrode has been used to polarize the edge plasma. The edge plasma potential and flows have been characterized by means of Langmuir and Mach probes, the radiation was measured using an array of bolometers. Potential profiles and poloidal flows can be well fitted by an exponential decay time in the range of 10 - 30 μs when the electrode biasing is turned off in the Castor tokamak. The radiation shows a slower time scale (about 1 ms), which is linked to the evolution in the plasma density and particle confinement. (authors)

  17. Determination of the plasma position for its real-time control in the COMPASS tokamak

    International Nuclear Information System (INIS)

    Janky, F.; Havlicek, J.; Valcarcel, D.; Hron, M.; Horacek, J.; Kudlacek, O.; Panek, R.; Carvalho, B.B.

    2011-01-01

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  18. Determination of the plasma position for its real-time control in the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)

    2011-10-15

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  19. The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Ward, D.J.; Jardin, S.C.

    1991-09-01

    The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration. (author) 19 figs., 2 tabs., 30 refs

  20. 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

  1. Soft x-ray imaging system for measurement of noncircular tokamak plasmas

    International Nuclear Information System (INIS)

    Fonck, R.J.; Reusch, M.; Jaehnig, K.P.; Hulse, R.; Roney, P.

    1986-08-01

    A soft x-ray camera and image processing system has been constructed to provide measurements of the internal shape of high temperature tokamak plasmas. The camera consists of a metallic-foil-filtered pinhole aperture and a microchannel plate image intensifier/convertor which produces a visible image for detection by a CCD TV camera. A wide-angle tangential view of the toroidal plasma allows a single compact camera to view the entire plasma cross section. With Be filters 12 to 50 μm thick, the signal from the microchannel plate is produced mostly by nickel L-line emissions which orignate in the hot plasma core. The measured toroidal image is numerically inverted to produce a cross-sectional soft x-ray image of the plasma. Since the internal magnetic flux surfaces are usually isothermal and the nickel emissivity depends strongly on the local electron temperature, the x-ray emission contours reflect the shape of the magnetic surfaces in the plasma interior. Initial results from the PBX tokamak experiment show clear differences in internal plasma shapes for circular and bean-shaped discharges

  2. Effect of plasma density profile of tokamak on Kelvin-Helmholtz instability

    International Nuclear Information System (INIS)

    Tang Fulin

    1984-01-01

    The purpose of this paper is to study the effect of radial distribution of plasma density profile of tokamak on Kelvin-Helmholtz instability caused by toroidal rotation. The effect of radial distribution of plasma rotational velocity on stability is also examine for comparison. It is found that within the range of tokamak parameters the only radial distribution of plasma rotational velocity cannot induce Kelvin-Helmholtz instability. On the contrary, when there is a radial distribution of plasma density, i.e. P 01 =P 0 e -tx and V 0 1 = const, plasma becomes unstable, and instability will increase proportionally to the value of t. Meanwhile when the value of t remains constant, the instability growth rate will decrease if P 0 grows or the distance between plasma and wall of container decreases too. It shows that the Kelvin-Helmoltz instability is not only influenced by the steepness of density profile but also by the inertia of plasma in central region, which is helpful for depressing the instability. (author). 5 refs, 4 figs, 2 tabs

  3. Peeling-off of the external kink modes at tokamak plasma edge

    International Nuclear Information System (INIS)

    Zheng, L. J.; Furukawa, M.

    2014-01-01

    It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture

  4. Peeling-off of the external kink modes at tokamak plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, L. J. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Furukawa, M. [Graduate School of Engineering, Tottori University, Tottori 680-8552 (Japan)

    2014-08-15

    It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.

  5. Magnetic spires for the detection of the position of the plasma column in a Tokamak (linear approximation); Espiras magneticas para la deteccion de la posicion de la columna de plasma en un Tokamak (aproximacion lineal)

    Energy Technology Data Exchange (ETDEWEB)

    Colunga S, S

    1990-07-15

    In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)

  6. 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)

  7. 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.

  8. 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)

  9. Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2008-01-01

    Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434

  10. Development of gyrotrons for plasma diagnostics (invited)

    International Nuclear Information System (INIS)

    Woskoboinikow, P.

    1986-01-01

    Recent advances in high-frequency (>100 GHz) gyrotron technology are reviewed and application to millimeter/submillimeter wave plasma diagnostics is discussed. Gyrotrons have useful capabilities of high-power (>1 kW), long pulse/cw operation, narrow linewidth (<100 kHz), and good spatial mode quality with efficient (--90%) mode converters. These capabilities could be used to significantly improve collective Thomson scattering diagnostics for the study of instabilities, plasma waves, turbulence, and thermal ion fluctuations. Imaging applications with many detector arrays of plasma density, field direction, and microinstabilities may be possible with gyrotons. In a high-field compact ignition tokamak experiment a possible millimeter wave diagnostics window could be exploited by a gyrotron to measure a number of parameters, including alpha particle density and velocity distribution

  11. 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

  12. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.; Pomphrey, N.; Sugiyama, L.E.

    1997-01-01

    The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code, studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data

  13. 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

  14. 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

  15. The importance of the toroidal magnetic field for the feasibility of a tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2000-01-01

    The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER

  16. Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma

    International Nuclear Information System (INIS)

    Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.

    1995-01-01

    The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered

  17. Analysis of tokamak plasma confinement modes using the fast Fourier transformation

    International Nuclear Information System (INIS)

    Mirmoeini, S.R.; Salar Elahi, A.; Ghoranneviss, M.

    2016-01-01

    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 without external field (electric biasing), and then compared it with each other. After the Fourier analysis of Mirnov coil data, the diagram of power spectrum density was depicted in different angles of Mirnov coils in the 'presence of external field' as well as in the 'absence of external field'. The power spectrum density (PSD) interprets the manner of power distribution of a signal with frequency. In this article, the number of plasma modes and the safety factor q were obtained by using the mode number of q = m/n (m is the mode number). The maximum MHD activity was obtained in 30-35 kHz frequency, using the density of the energy spectrum. In addition, the number of different modes across 0-35 ms time was compared with each other in the presence and absence of the external field. (author)

  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. The basics of spherical tokamaks and progress in European research

    International Nuclear Information System (INIS)

    Gusev, V K; Alladio, F; Morris, A W

    2003-01-01

    When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference

  20. Formation and sustainment of a low aspect ratio tokamak by a series of plasma injections

    International Nuclear Information System (INIS)

    Shimamura, Shin; Taniguchi, Makoto; Takahashi, Tsutomu; Nogi, Yasuyuki

    1995-01-01

    A low aspect ratio tokamak plasma was generated and sustained by injecting a series of plasmas from a magnetized coaxial gun into a flux conserver with toroidal field. The magnetized coaxial gun was supplied by an oscillating current with a d.c. component. The first few current pulses injected plasma and helicity into the flux conserver. This pulse helicity injection method worked effectively to maintain the low aspect ratio tokamak. 8 refs., 5 figs

  1. 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.)

  2. 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.)

  3. 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.

  4. Tokamak plasma equilibrium problems with anisotropic pressure and rotation and their numerical solution

    International Nuclear Information System (INIS)

    Ivanov, A. A.; Martynov, A. A.; Medvedev, S. Yu.; Poshekhonov, Yu. Yu.

    2015-01-01

    In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented

  5. Sustainment of spherical tokamak by means of repetitive injection of compact torus plasma

    International Nuclear Information System (INIS)

    Shimamura, Shin; Matsura, Ken; Takahashi, Tsutomu; Nogi, Yasuyuki

    2000-01-01

    Sustainment of spherical tokamak (S.T.) has been studied. A compact torus (C.T.) plasma was injected into confinement region by magnetized coaxial gun. For start-up and sustainment of large main spherical tokamak, single pulsed injection of small C.T. is not sufficient in many cases. C.T.plasma injection of high repetition rate is required. For this purpose magnetized coaxial gun was driven with high repetition rate current. The first injected C.T. plasma could start-up S.T. without other help. The repetitive C.T. injection grew and sustained the S.T. plasma. A CCD camera with fast gated image intensifier took a cross sectional view of S.T. during the repetitive C.T. injection. (author)

  6. 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

  7. 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

  8. A study on the fusion reactor - Development of x-ray spectrometer for diagnosis of tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hong Young; Choi, Duk In; Seo, Sung Hun; Kwon, Gi Chung; Jun, Sang Jin; Heo, Sung Hoi; Lee, Chan Hui [Korea Advanced Institute of Science and Technolgoy, Taejon (Korea, Republic of)

    1996-09-01

    This report of research is on the development of X-ray Photo-Electron Spectrometer (PES) for diagnosis of tokamak plasma. The spectrometer utilizes the fact that the energy of photo-electron is given by the difference between the energy of X-ray and the binding energy of materials. In the research of this year, we constructed two spectrometers; one is operated in KAIST tokamak and the other in KT1 tokamak. In addition, we reviewed the characteristics of the x-ray filter, the photo-electric effect of carbon foils and the detection efficiency of MCP and x-ray radiation of plasma. We measured the x-ray radiation in tokamak and diagnosed the qualitative plasma parameters from the analysis of data. The major interesting plasma parameters, which we can diagnose with the spectrometer, are the electron temperature, Z{sub eff}, the spatial distribution of x-ray radiation and etc. 27 refs., 2 tabs., 20 figs. (author)

  9. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    International Nuclear Information System (INIS)

    Suratia, Pooja; Patel, Jigneshkumar; Rajpal, Rachana; Kotia, Sorum; Govindarajan, J.

    2012-01-01

    Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  10. 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

  11. Asymmetry of edge plasma turbulence in biasing experiments on tokamak TF-2

    International Nuclear Information System (INIS)

    Budaev, V.P.

    1994-01-01

    It was observed in tokamaks the suppression of edge turbulence causes by setting a radial electric field at the plasma edge. The poloidal plasma rotation governed by this electric field is likely to result in changes in edge convention and poloidal asymmetry, however there is no experimental evidence about that of the experimental database concerning the biasing and conditions of edge plasma electrostatic turbulence excitation is not still complete. Also a relation between macroscopic convection and small-scale electrostatic turbulence have not yet revealed both in biasing and non biasing plasmas. In this paper results from biasing experiments carried on on ohmically heated tokamak TF-2 are presented. Changes in both equilibrium and fluctuated edge plasma parameters also convection and turbulence driven particle flux were demonstrated in probe measurements with biasing of electrode immersed within Last Closed Flux Surface (LCFS). Poloidal edge plasma structure and charge in asymmetry have demonstrated in the biasing experiments. (author). 6 refs, 4 figs

  12. 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)

  13. Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

    International Nuclear Information System (INIS)

    Selig, G.

    2012-12-01

    In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

  14. Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    Goutych, I F; Gresillon, D; Sitenko, A G

    1997-12-31

    This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas.

  15. Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Goutych, I.F.; Gresillon, D.; Sitenko, A.G.

    1996-01-01

    This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas

  16. Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas

    International Nuclear Information System (INIS)

    Forster, Michael

    2012-01-01

    Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the

  17. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    International Nuclear Information System (INIS)

    Goodall, D.H.J.

    1982-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

  18. 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

  19. 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)

  20. Analysis of line integrated electron density using plasma position data on Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    Nam, Y. U.; Chung, J.

    2010-01-01

    A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.

  1. Far infrared polarimetry with tokamak plasmas for determination of the poloidal magnetic field distribution

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, W

    1979-01-01

    This study examines the poloidal magnetic field distribution of tokamak plasma, and the elucidation of the radial distribution of the toroidal plasma flow. A numerical and experimental determination of the poloidal field based on the Faraday effect is presented. A method is discussed for measuring the rotation of the polarization plane linear polarized electromagnetic radiation, by passing through a plasma magnetized in the direction of the radiation. The polarization behavior of a linear polarized wave passing through a tokamak plasma is presented theoretically for various wavelengths, along with the experimental investigation of a ferrite modulation procedure through the use of different far infrared detectors.

  2. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a Tokamak plasma diagnostic

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

    Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission

  3. Rapid further heating of tokamak plasma by fast-rising magnetic pulse

    International Nuclear Information System (INIS)

    Inoue, N.; Nihei, H.; Yamazaki, K.; Ichimura, M.; Morikawa, J.; Hoshino, K.; Uchida, T.

    1977-01-01

    The object of the experiment was to study the rapid further heating of a tokamak plasma and its influence on confinement. For this purpose, a high-voltage theta-pinch pulse was applied to a tokamak plasma and production of a high-temperature (keV) plasma was ensured within a microsecond. The magnetic pulse is applied at the plasma current maximum parallel or antiparallel to the study toroidal field. In either case, the pulsed field quickly penetrates the plasma and the plasma resistivity estimated from the penetration time is about 100 times larger than the classical. A burst of energetic neutrals of approximately 1 μs duration was observed and the energy distribution had two components of the order of 1 keV and 0.1 keV in the antiparallel case. Doppler broadening measurement shows heating of ions to a temperature higher than 200 eV; however, the line profile is not always Maxwellian distribution. The X-rays disappear at the moment of applying the magnetic pulse and reappear about 100 μs later with an intensive burst, while both energy levels are the same (approximately 100 keV). (author)

  4. 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

  5. Plasma current sustainment after iron core saturation in the STOR-M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Ding, Y.; Hubeny, M.; Lu, Y.; Onchi, T.; McColl, D.; Xiao, C.; Hirose, A. [Plasma Physics Laboratory, University of Saskatchewan, 116 Science Place, Saskatoon, SK S7N 5E2 (Canada)

    2014-10-15

    Highlights: • Plasma current can be started up by small iron core without central solenoid. • Iron core removes central solenoid. • Plasma current can be maintained after iron core saturation. • Hysteresis curve shows the partial core saturation. • Image field from iron core is estimated during discharge. • Spherical tokamak reactor without CS is proposed using the small iron core. - Abstract: We propose to use of a small iron core transformer to start up the plasma current in a spherical tokamak (ST) reactor without central solenoid (CS). Taking advantage of the high aspect ratio of the STOR-M iron core tokamak, we have demonstrated that the plasma current up to 10–15 kA can be started up using the outer Ohmic heating (OH) coils without CS, and that the plasma current can be maintained further by increasing the outer OH coil current during iron core saturation phase. When the magnetizing current reaches 1.2 kA and the iron core becomes saturated, the third capacitor bank connected to the outer OH coils is discharged to maintain the plasma current. The plasma current is slightly increased and maintained for additional 5 ms as expected from numerical calculations. Core saturation has been clearly observed on the hysteresis curve. This is the first experimental demonstration of the feasibility of slow transition from the iron core to air core transformer phase without CS. The results implies that a plasma current can be initiated by a small iron core and could be ramped up by additional heating and vertical field after iron core saturation in future STs without CS.

  6. Thermal loads on tokamak plasma-facing components during normal operation and disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.

    1990-01-01

    Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

  7. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. Can tokamaks PFC survive a single event of any plasma instabilities?

    Science.gov (United States)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-07-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  15. Can tokamaks PFC survive a single event of any plasma instabilities?

    International Nuclear Information System (INIS)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-01-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time

  16. Can tokamaks PFC survive a single event of any plasma instabilities?

    Energy Technology Data Exchange (ETDEWEB)

    Hassanein, A., E-mail: hassanein@purdue.edu [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States); Sizyuk, V.; Miloshevsky, G.; Sizyuk, T. [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

    2013-07-15

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  17. Magnetic spires for the detection of the position of the plasma column in a Tokamak (linear approximation)

    International Nuclear Information System (INIS)

    Colunga S, S.

    1990-07-01

    In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)

  18. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

  19. Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback

    International Nuclear Information System (INIS)

    Ward, D.J.; Jardin, S.C.; Cheng, C.Z.

    1991-07-01

    A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs

  20. 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)

  1. Electron density and temperature determination in a Tokamak plasma using light scattering

    International Nuclear Information System (INIS)

    Perez-Navarro Gomez, A.; Zurro Hernandez, B.

    1976-01-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthemore, a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, is included in a Tokamak plasma using spectral analysis of the scattered radiation. (author) [es

  2. 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)

  3. 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

  4. The effect of tangled magnetic fields on instabilities in tokamak plasmas

    International Nuclear Information System (INIS)

    Thornton, A J; Kirk, A; Harrison, J R; Chapman, I T; Cahyna, P; Nardon, E

    2014-01-01

    The high pressure gradients in the edge of a tokamak plasma can lead to the formation of explosive plasma instabilities known as edge localised modes (ELMs). The control of ELMs is an important requirement for the next generation of fusion devices such as ITER. Experiments performed on the Mega Amp Spherical Tokamak (MAST) at Culham have shown that the application of non-axisymetric resonant magnetic perturbations (RMPs) can be used to mitigate ELMs. During the application of the RMPs, clear structures are observed in visible- light imaging of the X-point region. These lobes, or tangles, have been observed for the first time and their appearance is correlated with the mitigation of ELMs. Tangle formation is seen to be associated with the RMPs penetrating the plasma and may be important in explaining why the ELM frequency increases during ELM mitigation. Whilst the number and location of the tangles can be explained by vacuum magnetic field modelling, obtaining the correct radial extent of the tangles requires the plasma response to be taken into account

  5. 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

  6. 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

  7. Electron density and temperature determination in a Tokamak plasma using light scattering

    International Nuclear Information System (INIS)

    Perez-Navarro Gomerz, A.; Zurro Hernandez, B.

    1976-01-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs

  8. 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)

  9. 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

  10. 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

  11. 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

  12. I.R. and F.I.R. laser polarimetry as a diagnostic tool in high-β and Tokamak plasmas

    International Nuclear Information System (INIS)

    Pereira, D.; Machida, M.; Scalabrin, A.

    1986-01-01

    The change of the polarization state of an electromagnetic wave (E.M.W.) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the E.M.W. allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-Pinches plasmas (lambda approx. 10μm). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-Pinch/UNICAMP, using lasers developed at UNICAMP. (Author) [pt

  13. 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)

  14. Microwave reflectrometry for electron density measurements in the TJ-1 tokamak plasma

    International Nuclear Information System (INIS)

    Anabitarte, E.; Bustamante, E.G.; Calderon, M.A.G.; Vegas, A.

    1986-01-01

    A study about microwave reflectometry to measure the outside profile of the electron plasma density on tokamak TJ-1 is presented. It is also presented the condition of applicability of this method after the characteristic parameters of the plasma and its resolution. The simulation of the plasma in laboratory by means of a metallic mirror causes the whole characterization of the reflectometer. (author)

  15. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  16. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  17. 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

  18. 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

  19. Probes for edge plasma studies of TFTR (invited)

    International Nuclear Information System (INIS)

    Manos, D.M.; Budny, R.V.; Kilpatrick, S.; Stangeby, P.; Zweben, S.

    1986-01-01

    Tokamak fusion test reactor (TFTR) probes are designed to study the interaction of the plasma with material surfaces such as the wall and limiters, and to study the transport of particles and energy between the core and edge. Present probe heads have evolved from prototypes in Princeton large torus (PLT), poloidal divertor experiment (PDX) [Princeton BETA experiment (PBX)], and the initial phase of TFTR operation. The newest heads are capable of making several simultaneous measurements and include Langmuir probes, heat flux probes, magnetic coils, rotating calorimeter fast ion probes, and sample exposure specimens. This paper describes these probe heads and presents some of the data they and their prototypes have acquired. The paper emphasizes measurement of transient plasma effects such as fast ion loss during auxiliary heating, the evolution of the edge plasma during heating, compression, and free expansion, and fluctuations in the edge plasma

  20. Plasma formation and first OH experiments in GLOBUS-M tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Aleksandrov, S.V.; Burtseva, T.A.

    2001-01-01

    The paper reports results of experimental campaigns on plasma ohmic heating, performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Also discussed is the influence of ECR preioniziation on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid (CS) limited to ∼100 mVs are presented. Ramp-up of the plasma current of 0.25 MA for the time interval ∼0.03 s with about 0.02 s flat-top at the toroidal field (TF) strength of 0.35 T allows the conclusion that power supplies, control system and wall conditioning work well. The same conclusion can be drawn from observation of plasma density behavior the density is completely controlled with external gas puff and the influence of the wall is negligible after boronization. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. (author)

  1. 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

  2. The rate of plasma heating by harmonic ion cyclotron waves in tokamaks

    International Nuclear Information System (INIS)

    Moslehi-Fard, M.; Sobhanian, S.; Solati-Kia, F.

    2002-01-01

    In tokamaks, the toroidal magnetic field, B φ , is due to the current in coils around plasma, and the poloidal magnetic field B p results from the plasma itself. Usually B φ p , and the combination of these two fields forms a nested set of toroidal magnetic surfaces. The equilibrium Grad-Shafranov equation is investigated and it is shown that the particle products of fusion with different pitch angles on these surfaces have different orbital shapes. In the JET tokamak, the α particles with pitch angle θ smaller than 54.8 deg are passing, those with θ between 54.8 deg and 65.1 deg have trapping-passing orbits but for θ greater than 65.1 deg the orbit has a banana form. Other tokamaks such as Alcator and ITER are also considered. The passing, trapping-passing and banana orbits in these tokamaks are traced. The results obtained from this calculation are analyzed. The wave damping has been investigated produced from interaction with particles, particularly α particles, and the rate of heating for l = 1 to 8 harmonics is plotted. The results of calculation show that heating at the fourth harmonic reaches a maximum. For higher harmonics, the heating does not change much from the fourth harmonic. (author)

  3. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Tammen, H.F.

    1995-01-10

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).

  4. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    International Nuclear Information System (INIS)

    Tammen, H.F.

    1995-01-01

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)

  5. Inertia effects on the rigid displacement approximation of tokamak plasma vertical motion

    International Nuclear Information System (INIS)

    Carrera, R.; Khayrutdinov, R.R.; Azizov, E.A.; Montalvo, E.; Dong, J.Q.

    1991-01-01

    Elongated plasmas in tokamaks are unstable to axisymmetric vertical displacements. The vacuum vessel and passive conductors can stabilize the plasma motion in the short time scale. For stabilization of the plasma movement in the long time scale an active feedback control system is required. A widely used method of plasma stability analysis uses the Rigid Displacement Model (RDM) of plasma behavior. In the RDM it is assumed that the plasma displacement is small and usually plasma inertia effects are neglected. In addition, it is considered that no changes in plasma shape, plasma current, and plasma current profile take place throughout the plasma motion. It has been demonstrated that the massless-filament approximation (instantaneous force-balance) accurately reproduces the unstable root of the passive stabilization problem. Then, on the basis that the instantaneous force-balance approximation is correct in the passive stabilization analysis, the massless approximation is utilized also in the study of the plasma vertical stabilization by active feedback. The authors show here that the RDM (without mass effects included) does not provide correct stability results for a tokamak configuration (plasma column, passive conductors, and feedback control coils). Therefore, it is concluded that inertia effects have to be retained in the RDM system of equations. It is shown analytically and numerically that stability diagrams with and without plasma-mass corrections differ significantly. When inertia effects are included, the stability region is more restricted than obtained in the massless approximation

  6. Study on possibility of plasma current profile determination using an analytical model of tokamak equilibrium

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Hiraki, Naoji

    1996-01-01

    The possibility of determining the current profile of tokamak plasma from the external magnetic measurements alone is investigated using an analytical model of tokamak equilibrium. The model, which is based on an approximate solution of the Grad-Shafranov equation, can set a plasma current profile expressed with four free parameters of the total plasma current, the poloidal beta, the plasma internal inductance and the axial safety factor. The analysis done with this model indicates that, for a D-shaped plasma, the boundary poloidal magnetic field prescribing the external magnetic field distribution is dependent on the axial safety factor in spite of keeping the boundary safety factor and the plasma internal inductance constant. This suggests that the plasma current profile is reversely determined from the external magnetic analysis. The possibility and the limitation of current profile determination are discussed through this analytical result. (author)

  7. 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

    Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have

  8. 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

  9. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  10. Wall conditioning of the TBR-1 Tokamak by plasma generated by microwaves

    International Nuclear Information System (INIS)

    Elizondo, J.I.

    1986-01-01

    A new system of vaccum chamber wall conditioning in the TBR-1 Tokamak, using electron cyclotron resonance plasma of hydrogen for the discharge cleaning process is presented. The construction and performance of equipments are described, and the cleaning process to otimize the conditioning efficiency by chase of plasma parameters. (author) [pt

  11. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering ; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0022311517301708

  12. 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)

  13. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (author)

  14. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)

  15. 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

  16. 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

  17. Two-body similarity and its violation in tokamak edge plasmas

    International Nuclear Information System (INIS)

    Catto, P.J.; Knoll, D.A.; Krasheninnikov, S.I.

    1996-01-01

    Scaling laws found under the assumption that two-body collisions dominate can be effectively used to benchmark complex multi-dimensional codes dedicated to investigating tokamak edge plasmas. The applicability of such scaling laws to the interpretation of experimental data, however, is found to be restricted to the relatively low plasma densities ( 19 m -3 ) at which multistep processes, which break the two-body collision approximation, are unimportant. copyright 1996 American Institute of Physics

  18. Simulations of the operational control of a cryogenic plant for a superconducting burning-plasma tokamak

    CERN Document Server

    Mitchell, N

    2001-01-01

    In recent proposals for next generation superconducting tokamaks, such as the ITER project, the nuclear burning plasma is confined by magnetic fields generated from a large set (up to 100 GJ stored energy) of superconducting magnets. These magnets suffer heat loads in operation from thermal and nuclear radiation from the surrounding components and plasma as well as eddy currents and AC losses generated within the magnets, together with the heat conduction through supports and resistive heat generated at the current lead transitions to room temperature. The initial cryoplant for such a tokamak is expected to have a steady state capacity of up to about 85 kW at 4.5 K, comparable to the system installed for LHC at CERN. Experimental tokamaks are expected to operate at least initially in a pulsed mode with 20-30 short plasma pulses and plasma burn periods each day. A conventional cryoplant, consisting of a cold box and a set of primary heat exchangers, is ill-suited to such a mode of operation as the instantaneou...

  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. 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

  1. Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas

    Science.gov (United States)

    Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.

    1980-01-01

    The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.

  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. 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

  4. Analysis of plasma flow in a scrape-off layer in a tokamak

    International Nuclear Information System (INIS)

    Petrov, V.G.

    1988-01-01

    Plasma drift on the periphery of a tokamak in a magnetic tube, on sides of which coins are placed, is considered. Convection caused by toroidal particle drift is taken into account. Distribution of plasma parameters in such a tube is found. Transition from the total poloidal diaphragm to a sectioned one is traced

  5. 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.

  6. 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

  7. 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

  8. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  9. 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

  10. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    International Nuclear Information System (INIS)

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

  11. Plasma jet source parameter optimisation and experiments on injection into Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Semenov, A.A.; Voronin, A.V.

    2005-01-01

    Results of theoretical and experimental research on the plasma sources and injection of plasma and gas jet produced by the modified source into tokamak Globus-M are presented. An experimental test stand was developed for investigation of intense plasma jet generation. Optimisation of pulsed coaxial accelerator parameters by means of analytical calculations is performed with the aim of achieving the highest flow velocity at limited coaxial electrode length and discharge current. The optimal parameters of power supply to generate a plasma jet with minimal impurity contamination and maximum flow velocity were determined. A comparison of experimental and calculation results is made. Plasma jet parameters are measured, such as: impurity species content, pressure distribution across the jet, flow velocity, plasma density, etc. Experiments on the interaction of a higher kinetic energy plasma jet with the magnetic field and plasma of the Globus-M tokamak were performed. Experimental results on plasma and gas jet injection into different Globus-M discharge phases are presented and discussed. Results are presented on the investigation of plasma jet injection as the source for discharge breakdown, plasma current startup and initial density rise. (author)

  12. 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

  13. A study on the heating and diagnostic of a tokamak plasma by electromagnetic waves of the electron cyclotron range of frequencies

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1989-09-01

    A study on the heating and diagnosis of tokamak plasma by electromagnetic waves of electron cyclotron range of frequency is summarized. The main results obtained are as follows. On the engineering and technology, the technology of injecting high frequency, large power millimeter waves into tokamak plasma was established by carrying out the design, manufacture and test of a 60 GHz, 400 kW high frequency heating system, and the design, manufacture and test of a heterodyne type electron cyclotron radiation multi-channel mealsuring system were carried out, and the technology of measuring the radiation from tokamak plasma with the time resolution of 10 μs in multi-channel was established. On nuclear fusion reactor core engineering and plasma physics, the high efficiency electron heating of tokamak plasma by the incidence of fundamental irregular and regular waves at electron cyclotron frequency was verified. The discovery and analysis of the heating by electrostatic waves arising due to mode transformation from electromagnetic waves in upper hybrid resonance layer were carried out. By the incidence of second harmonic waves, the high efficiency electron heating of tokamak plasma was verified, and the heating characteristics were clarified. And others. (K.I.) 179 refs

  14. 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

  15. Stabilization of external kink modes in a tokamak with rotating plasma

    International Nuclear Information System (INIS)

    Mikhailovskii, A.B.; Kuvshinov, B.N.

    1995-01-01

    An analytical theory of stabilization of external kink modes in a tokamak with rotating plasma is developed, which is of interest in connection with experiments on the DIII-D tokamak demonstrating such a stabilization. It is assumed that, in addition to the main poloidal harmonic, the mode includes one or more side-band poloidal harmonics with singular points lying inside the plasma. Near these singular points, plasma inertia and related toroidal effects, the compressible part of plasma pressure and longitudinal viscosity, are allowed for. These effects are described kinetically taking into account the toroidal trapping of the resonant ions, which is essential if the toroidal velocity is small compared to the ion thermal velocity. Thereby, the theory presented includes both ion Landau damping and its weakening due to toroidal trapping. Near the singular points high-beta effects, which result in the finiteness of the Mercier index s, are allowed for. It is shown that the influence of plasma rotation on the external kink modes is most significant in the case of s<0, i.e., when the development of the instability in a non-rotating plasma is most highly favored. In this case, the plasma rotation plays a stabilizing role, even when the ion Landau damping is neglected. The analysis presented also confirms the hypothesis of Bondeson and Ward on the stabilizing effect of ion Landau damping if this damping is not too small

  16. Magnetohydrodynamic helical structures in nominally axisymmetric low-shear tokamak plasmas

    International Nuclear Information System (INIS)

    Graves, J P; Brunetti, D; Cooper, W A; Reimerdes, H; Halpern, F; Pochelon, A; Sauter, O; Chapman, I T

    2013-01-01

    The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER. (paper)

  17. Feedback control of horizontal position and plasma surface shape in a non-circular tokamak

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Nakamura, Kazuo; Nakamura, Yukio; Itoh, Satoshi

    1986-01-01

    The linear model for the coupled horizontal position and plasma surface shape control in the non-circular tokamak device was described. It enables us to estimate easily the displacement and the distortion due to the changes in plasma pressure and current density distribution. The PI-controller and the optimal regulator were designed with the linear model. Transient-response analysis of the control system in the TRIAM-1M tokamak showed that the optimal regulator is superior to the PI-controller with regard to the mutual-interference between the position control system and the elongation control system. (author)

  18. 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)

  19. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D (Research and Development) avenues for their resolution are presented

  20. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  1. Nanoparticle Plasma Jet as Fast Probe for Runaway Electrons in Tokamak Disruptions

    Science.gov (United States)

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

    2017-10-01

    Successful probing of runaway electrons (REs) requires fast (1 - 2 ms) high-speed injection of enough mass able to penetrate through tokamak toroidal B-field (2 - 5 T) over 1 - 2 m distance with large assimilation fraction in core plasma. A nanoparticle plasma jet (NPPJ) from a plasma gun is a unique combination of millisecond trigger-to-delivery response and mass-velocity of 100 mg at several km/s for deep direct injection into current channel of rapidly ( 1 ms) cooling post-TQ core plasma. After C60 NPPJ test bed demonstration we started to work on ITER-compatible boron nitride (BN) NPPJ. Once injected into plasma, BN NP undergoes ablative sublimation, thermally decomposes into B and N, and releases abundant B and N high-charge ions along plasma-traversing path and into the core. We present basic characteristics of our BN NPPJ concept and first results from B and N ions on Zeff > 1 effect on REs dynamics by using a self-consistent model for RE current density. Simulation results of BNQ+ NPPJ penetration through tokamak B-field to RE beam location performed with Hybrid Electro-Magnetic code (HEM-2D) are also presented. Work supported by U.S. DOE SBIR Grant.

  2. 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)

  3. 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.

  4. 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.

  5. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  6. Reactor aspects of counterstreaming-ion tokamak plasmas

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1975-06-01

    Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)

  7. 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)

  8. 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

  9. Twentyseventh European physical society conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Igitkhanov, Y.

    2000-01-01

    The twentyseventh European physical society conference on controlled fusion and plasma physics was held in Budapest, 12-16 June 2000. About 10 invited papers were presented, covering a wide range of problems in plasma physics, including confinement and transport issues in fusion devices, astrophysics and industrial application of plasmas. More than 100 papers were presented on plasma theory and experiments from tokamaks and stellarators. Some of the ITER-relevant issues covered are described in this newsletter

  10. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  11. 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)

  12. Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas

    International Nuclear Information System (INIS)

    Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.

    1989-06-01

    The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs

  13. 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)

  14. Neutral-beam requirements for compression-boosted ignited tokamak plasmas

    International Nuclear Information System (INIS)

    Cohn, D.R.; Jassby, D.L.; Kreischer, K.

    1977-12-01

    Neutral-beam energies of 200 to 500-keV D 0 may be required to insure adequate penetration into the center of ignition-sized tokamak plasmas. However, the beam energy requirement can be reduced by using a start-up scenario in which the final plasma is formed by major-radius compression of a beam-heated plasma whose density-radius product, na, is determined by satisfactory neutral-beam penetration. ''Compression boosting'' is attractive only for plasmas in which ntau/sub E/ increases with na, because a major-radius compression C increases na by C 3 / 2 . The dependence on C of beam energy and beam power for plasmas which obey ''empirical scaling laws'' of the type ntau/sub E/ varies as (na) 2 is analyzed. The dependences on C of stored magnetic energy and TF-coil power dissipation are also determined. It is found that a compression ratio of 1.5 to attain the ignited plasma permits adequate penetration by 150-keV D 0 beams

  15. Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

    International Nuclear Information System (INIS)

    Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.

    2005-01-01

    Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution

  16. 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)

  17. 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.

  18. 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.

  19. On thermionic emission from plasma-facing components in tokamak-relevant conditions.

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Ratynskaia, S.; Tolias, P.; Cavalier, Jordan; Dejarnac, Renaud; Gunn, J. P.; Podolník, Aleš

    2017-01-01

    Roč. 59, č. 9 (2017), č. článku 094002. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * PIC * tungsten * tokamak * thermionic emission * plasma facing components * particle-in-cell Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/1361-6587/aa78c4/pdf

  20. 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)

  1. 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

  2. Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Chrystal, C. [University of California-San Diego, La Jolla, California 92186-5608 (United States); Burrell, K. H.; Groebner, R. J.; Kaplan, D. H. [General Atomics, San Diego, California 92186-5608 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

    2012-10-15

    To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

  3. Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma.

    Science.gov (United States)

    Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H

    2012-10-01

    To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

  4. Internal helical modes with m > 1 in a tokamak with a small shear and high plasma pressure

    International Nuclear Information System (INIS)

    Mikha lovskij, A.B.; Aburdzhaniya, G.D.; Krymskij, A.M.

    1979-01-01

    Internal helical modes with m>1 in a circular cross-section tokamak with a small shear and large value of the parameter β (β is the ratio between the mean plasma pressure and the mean pressure of the poloidal magnetic field) are investigated. The equations obtained are used to study the destabilizing effects leading to helical instabilities. The role of destabilizing effects is regarded both in local and in a nonlocal approximations on the assumption that the radial plasma pressure is distributed parabolically and that the radial current distribution is also parabolic though slightly varying. It has been established that the profiling of current may lead to the tokamak plasma stability with respect to the modes under investigation. A tokamak with a small shear has been shown to be more stable relative to these modes than that with a large shear

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. Control oriented modeling and simulation of the sawtooth instability in nuclear fusion tokamak plasmas

    NARCIS (Netherlands)

    Witvoet, G.; Westerhof, E.; Steinbuch, M.; Doelman, N.J.; Baar, de M.R.

    2009-01-01

    Tokamak plasmas in nuclear fusion are subject to various instabilities. A clear example is the sawtooth instability, which has both positive and negative effects on the plasma. To optimize between these effects control of the sawtooth period is necessary. This paper presents a simple control

  10. Plasma rotation measurement in small tokamaks using an optical spectrometer and a single photomultiplier as detector.

    Science.gov (United States)

    Severo, J H F; Nascimento, I C; Kuznetov, Yu K; Tsypin, V S; Galvão, R M O; Tendler, M

    2007-04-01

    The method for plasma rotation measurement in the tokamak TCABR is reported in this article. During a discharge, an optical spectrometer is used to scan sequentially spectral lines of plasma impurities and spectral lines of a calibration lamp. Knowing the scanning velocity of the diffraction grating of the spectrometer with adequate precision, the Doppler shifts of impurity lines are determined. The photomultiplier output voltage signals are recorded with adequate sampling rate. With this method the residual poloidal and toroidal plasma rotation velocities were determined, assuming that they are the same as those of the impurity ions. The results show reasonable agreement with the neoclassical theory and with results from similar tokamaks.

  11. Unified Ideal Stability Limits for Advanced Tokamak and Spherical Torus Plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Sabbagh, S.A.; Fredrickson, E.D.; Jardin, S.C.; Maingi, R.; Manickam, J.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, Y.-K.M.; Soukhanovskii, V.; Stutman, D.; Synakowski, E.J.

    2003-01-01

    Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants

  12. Plasma heating and fuelling in the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Barsukov, A.G.; Belyakov, V.A.

    2005-01-01

    The results of the last two years of plasma investigations at Globus-M are presented. Described are improvements helping to achieve high performance OH plasmas, which are used as the target for auxiliary heating and fuelling experiments. Increased energy content, high beta poloidal and good confinement are reported. Experiments on NBI plasma heating with a wide range of plasma parameters were performed. Some results are presented and analyzed. Experiments on RF plasma heating in the frequency range of fundamental ion cyclotron harmonics are described. In some experiments which were performed for the first time in spherical tokamaks, promising results were achieved. Noticeable ion heating was recorded at low launched power and a high concentration of hydrogen minority in deuterium plasmas. Simulations of RF wave absorption are briefly discussed. Described also are modification of the plasma gun and test-stand experiments. Fuelling experiments performed at Globus-M are discussed. (author)

  13. On lateral deflection of the SOL plasma in tokamaks during giant ELMs

    International Nuclear Information System (INIS)

    Landman, I.S.; Wuerz, H.

    2000-06-01

    In recent H-mode experiments at JET with giant ELMs a lateral deflection of hot tokamak plasma leaving the scrape-off layer and striking the divertor plate has been observed. This deflection can effect the divertor erosion caused by the hot plasma irradiation, because of enlarging the irradiated area. A simplified MHD model of the vapor shield plasma and of the hot plasma initially formed at time t → -∞ is analyzed. At t = -∞ both plasmas are assumed to stay on rest and to be separated by a boundary, which is parallel to the plate surface. The interaction between plasmas is assumed to develop gradually ('adiabatically') as exp(t/t 0 ) with t 0 ∝ 10 2 μs the ELM duration time. Electrical insulation of the core tokamak plasma is assumed everywhere except for the contact with the divertor. Electric currents are flowing only in the toroidal direction. These currents developing in the interaction zone of the hot plasma and the rather cold target plasma are calculated for inclined impact of the magnetized hot plasma. At such conditions the J x B force in the lateral direction accelerates the interacting plasmas. The motion of the cold plasma and the gradual increase of the plasma interaction intensity are shown to be important for the appropriate deflection magnitude. Adiabatically responding against the increase of the interaction intensity the cold plasma motion compensates significantly the currents thus decreasing the deflection compared to motionless approach. The calculated magnitude of the hot plasma deflection is comparable to the observed one. The results of the modeling are discussed in relation to the experiments. It is shown that sudden switching on of the interaction produces Alfven oscillations of large amplitudes causing much larger amplitudes of the magnetic field induced by the currents than in the adiabatic case. (orig.)

  14. Thermal and nonthermal electron cyclotron emission by high-temperature tokamak plasmas

    International Nuclear Information System (INIS)

    Airoldi, A.; Ramponi, G.

    1997-01-01

    An analysis of the electron cyclotron emission (ECE) spectra emitted by a high-temperature tokamak plasma in the frequency range of the second and third harmonic of the electron cyclotron frequency is made, both in purely Maxwellian and in non-Maxwellian cases (i.e., in the presence of a current-carrying superthermal tail). The work is motivated mainly by the experimental observations made in the supershot plasmas of the Tokamak Fusion Test Reactor (TFTR), where a systematic disagreement is found between the T e measurements by second-harmonic ECE and Thomson scattering. We show that, by properly taking into account the overlap of superthermals-emitted third harmonic with second-harmonic bulk emission, the radiation temperature observed about the central frequency of the second harmonic may be enhanced up to 30%endash 40% compared to the corresponding thermal value. Moreover we show that, for parameters relevant to the International Thermonuclear Experimental Reactor (ITER) with T e (0)>7 keV, the overlap between the second and the downshifted third harmonic seriously affects the central plasma region, so that the X-mode emission at the second harmonic becomes unsuitable for local T e measurements. copyright 1997 American Institute of Physics

  15. Spectra of germanium and selenium in the 50-350 A region from the PLT tokamak plasma

    International Nuclear Information System (INIS)

    Stratton, B.C.; Hodge, W.L.; Moos, H.W.; Schwob, J.L.; Suckewer, S.; Finkenthal, M.; Cohen, S.

    1983-03-01

    Spectra of germanium and selenium injected into the PLT tokamak plasma were observed in the 50 to 350 A region for GeXIV-XXV (KI to OI-like) and SeXVI-XXIV (KI to NaI-like). A number of 3p/sup k/-3p/sup k-1/3d transitions predicted by isoelectronic sequence extrapolation have been identified. Also, previously identified lines from ions in the AlI to OI-like and KI-like isoelectronic sequences have been observed in the tokamak plasma

  16. 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)

  17. The conversion of the thermal energy of plasma in the SOL of tokamaks

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Nenova, N.V.

    2008-01-01

    When the plasma expands across the confining magnetic field, a part of its thermal energy is converted to electrical energy. In the SOL of tokamaks, the motion of the plasma across the field due to turbulent processes is accompanied by its departure along the open lines of the magnetic field. The conversion of thermal energy is taken into account in theoretical studies devoted to the physics of plasma in the SOL; however, this conversion is ignored in numerical models, for example, in B2-SOLPS4.0. This paper deals with thermal-to-electrical energy conversion in the SOL of tokamaks. It is demonstrated that the part of the thermal energy subjected to conversion to electrical energy forms an appreciable part of the total energy flowing in the SOL. In ITER, this fraction may be as high as 20-25%. The electrical energy generated in the SOL volume is liberated in the form of Joule heat in a relatively cold plasma in the vicinity of diverter plates or directly on these plates. (letter)

  18. 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

  19. Design of an O-mode frequency modulated reflectometry system for the measurement of Alborz Tokamak plasma density profile

    Energy Technology Data Exchange (ETDEWEB)

    Koohestani, Saeideh [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Amrollahi, Reza, E-mail: amrollahi@aut.ac.ir [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Moradi, Gholamreza [Department of Electrical Engineering, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of)

    2016-12-15

    Reflectometry is a common method for plasma diagnostic, in which microwaves are launched into the plasma and reflected at the critical surfaces. Comparing the reflected microwave signals with the launched waves would give rise to the plasma density profiles. In the present study, an ordinary mode (O-mode) frequency modulation (FM) reflectometry system has been designed for the electron density profile measurement of the Alborz Tokamak plasma. This system has been considered to operate at K-band (18–26.5 GHz) frequency range and scan the frequency band between 18 to 26 GHz in 40 μS. The density profile from major radius r = 47.9–51.55 cm can be measured in Alborz Tokamak plasma. Based on the Alborz Tokamak operational conditions, the characteristic frequencies, and some dimensional limitations, all parts of reflectometer have been designed so that an appropriate efficiency with minimum attenuation, especially in transmitting/receiving system would be achieved. A dual antenna and an oversized waveguide of X-band (8–12 GHz) for transmitting and receiving purposes and a balanced detector for absolute phase determination have been utilized. The details of the Alborz Tokamak FM reflectometry components focusing on the antenna and waveguide design and mounting are described in this paper. Additionally, the procedure of plasma profile reconstruction using the system output signal is discussed. This system uses signal phase shift to determine the position of the cutoff layer.

  20. The diagnostic neutral beam injector with arc-discharge plasma source on the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Karpushov, Alexander N. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)], E-mail: alexander.karpushov@epfl.ch; Andrebe, Yanis; Duval, Basil P.; Bortolon, Alessandro [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)

    2009-06-15

    The diagnostic neutral beam injector (DNBI) together with a charge exchange recombination spectroscopy (CXRS) system has been used on the TCV Tokamak as a diagnostic tool for local measurements of plasma ion temperature, velocity and carbon impurity density based on analysis of the beam induced impurity radiation emission since 2000. To improve the performance of the CXRS diagnostic, several upgrades of both the optical system and the neutral beam were performed. An increase of the plasma source size together with beam optimization in 2003 resulted in a twofold increase the beam current. The RF plasma generator was replaced by an arc-discharge plasma source together with a new ion optical system (IOS) in 2006 and subsequent beam optimization is presented herein. This was designed to increase the line brightness of the beam in the CXRS observation region without increasing of the injected power (to avoid plasma perturbation by the beam). The beam characteristics are measured by a multi-chord scanning of Doppler-shifted H{sub {alpha}} emission, thermal measurements on a movable calorimeter and visible optical measurements inside the Tokamak vessel.

  1. Plasma fluctuation measurements in tokamaks using beam-plasma interactions (abstract)

    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

  2. High kinetic energy plasma jet generation and its injection into the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Voronin, A.V.; Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Abramova, K.B.; Sklyarova, E.M.; Tolstyakov, S.Yu.

    2005-01-01

    Progress in the theoretical and experimental development of the plasma jet source and injection of hydrogen plasma and neutral gas jets into the Globus-M spherical tokamak is discussed. An experimental test bed is described for investigation of intense plasma jets that are generated by a double-stage plasma gun consisting of an intense source for neutral gas production and a conventional pulsed coaxial accelerator. A procedure for optimizing the accelerator parameters so as to achieve the maximum possible flow velocity with a limited discharge current and a reasonable length of the coaxial electrodes is presented. The calculations are compared with experiment. Plasma jet parameters, among them pressure distribution across the jet, flow velocity, plasma density, etc, were measured. Plasma jets with densities of up to 10 22 m -3 , total numbers of accelerated particles (1-5) x 10 19 , and flow velocities of 50-100 km s -1 were successfully injected into the plasma column of the Globus-M tokamak. Interferometric and Thomson scattering measurements confirmed deep jet penetration and a fast density rise ( 19 to 1 x 10 19 ) did not result in plasma degradation

  3. 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.

  4. Large potential change induced by pellet injection in JIPP T-IIU tokamak plasmas

    International Nuclear Information System (INIS)

    Hamada, Y.; Sato, K.N.; Sakakita, H.

    1995-05-01

    A large, rapid change in the local plasma potential is found to be induced by off-axis hydrogen ice-pellet injection into a tokamak plasma. The polarity of the rapid change is reversed when the pellet is injected into the upper and lower halves of the poloidal plasma cross-section. This change can be interpreted as being due to the gradient-B drift of particles in the high-density plasmas of the pellet cloud, before the increase of the plasma density due to the ablation becomes uniform on the magnetic surface. (author)

  5. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

    Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

    1992-01-01

    Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

  6. Injection of electrons with predominantly perpendicular energy into an area of toroidal field ripple in a tokamak plasma to improve plasma confinement

    Science.gov (United States)

    Ono, Masayuki; Furth, Harold

    1993-01-01

    An electron injection scheme for controlling transport in a tokamak plasma. Electrons with predominantly perpendicular energy are injected into a ripple field region created by a group of localized poloidal field bending magnets. The trapped electrons then grad-B drift vertically toward the plasma interior until they are detrapped, charging the plasma negative. Calculations indicate that the highly perpendicular velocity electrons can remain stable against kinetic instabilities in the regime of interest for tokamak experiments. The penetration distance can be controlled by controlling the "ripple mirror ratio", the energy of the injected electrons, and their v.sub..perp. /v.sub.51 ratio. In this scheme, the poloidal torque due to the injected radial current is taken by the magnets and not by the plasma. Injection is accomplished by the flat cathode containing an ECH cavity to pump electrons to high v.sub..perp..

  7. Fast Waves Mode Conversion and Energy Deposition in Simulated, Pre-Heated, Neoclassical, Tight Aspect Ratio Tokamak Plasmas

    International Nuclear Information System (INIS)

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

    1999-01-01

    Some basic aspects of wave-plasma interaction of interest for tight aspect ratio spherical tokamaks are investigated theoretically. The following scenario is considered: A. Fast magnetosonic waves are launched by an external antenna into a simulated spherical Tokamak plasma; these waves are converted to Alfven waves at points (layer) satisfying the Alfven resonance condition. B. The simulated spherical tokamaks-plasma has a circular cross-section and toroidicity effects are simulated by Grad-Shafranov type, radially dependent axial magnetic field and its shear. (J. Actual equilibrium profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. D. The study is based on the numerical solution of the full e.m. wave equation which includes a quite general resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. Two kinds of results will be presented: I. Proofs validating the computational algorithm used and including convergence and energy conservation. II. Exact quantitative results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited p over . The dependence of the results on the launched wave characteristics (wave numbers, frequency and intensity) as well as on those of the equilibrium plasma (equilibrium current, neoclassical resistivity and electron inertia) will be discussed

  8. Study on the bywalled plasma near the FT-2 tokamak limiter

    International Nuclear Information System (INIS)

    Budnikov, V.N.; Gerasimenko, G.V.; Esipov, L.A.; Levitskij, A.N.; Sakharov, I.E.; Shatalin, S.V.

    1987-01-01

    Results od investigation into the bywalled plasma near the FT-2 tokamak limiter, conducted using the ternary probe method are presented. Plasma charged particle concentration radial profiles, results of measuring the electron temperature and azimuthal electric field are presented. It is shown, that at l > 2mm, where l is the distance up to the limiter, concentration practically does not depend on the longitudinal coordinate and near the limiter (l < 20 mm) a sharp drop of plasma charged particle concentration measured values (∼ 2-fold) is observed. Neoclassical mechanism of filling the depletion area (the probe shade) is confirmed

  9. Resistive internal kink modes in a tokamak with high-pressure plasma

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Mikhajlovskij, A.B.; Tatarinov, E.G.

    1988-01-01

    Theory of resistive internal kink modes in a tokamak with high-pressure plasma is developed. Equation for Fourie-image of disturbed displacment in a resistive layer ie derived with regard to effects of the fourth order by plasma pressure within the framework of single-liquid approach. In its structure this equation coincides with a similar equation for resistive balloon modes and has an exact solution expressed by degenerated hypergeometric function. A general dispersion equation for resistive kink modes is derived with regard to the effects indicated. It is shown that plasma pressure finiteness leads to the reduction of reconnection and tyring-mode increments

  10. Fast potential change in sawteeth in JIPP T-IIU tokamak plasmas

    International Nuclear Information System (INIS)

    Hamada, Y.; Nishizawa, A.; Kawasumi, Y.

    1994-12-01

    Fast changes of electric potential with different polarities are observed during sawtooth oscillation in a core region of a tokamak plasma using a heavy ion beam probe. The potential change inside the inversion radius is found to be positive. The change is negative outside the inversion radius and shows clearly a propagation nature. The observed potential can be interpreted by the mixture of the potentials of two origins. One of them drives the fast MHD plasma motion through E/B drift and the other is a barrier potential induced by mixing of hot and cold plasmas at sawtooth crash. (author)

  11. 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

  12. Bifurcation of radial electric field in tokamak edge plasmas due to ion orbit loss

    International Nuclear Information System (INIS)

    Wu, G.J.; Zhang, X.D.

    2015-01-01

    The ion orbit loss and the formation of radial electric field Er in tokamak edge region are calculated. The ion orbit loss generates a negative Er, which in turn affects the ion loss. As a result, Er can saturates at either a low or a high value, depending on the plasma parameters. When the ion temperature in the plasma edge is higher than a threshold a self-sustaining growth in both the ion loss and Er is found, leading to a high saturation value of Er in the milliseconds time. This mechanism provides a possible explanation for the formation of the edge radial electric field during the L to H-mode transition observed in tokamak experiments. (author)

  13. 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

  14. 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)

  15. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  16. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  17. The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas

    International Nuclear Information System (INIS)

    Park, H.K.; Batha, S.

    1997-02-01

    Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  18. Energy balance in the TCA tokamak plasma with Alfven wave heating

    International Nuclear Information System (INIS)

    Ding Ning; Qu Wenxiao; Huang Li; Long Yongxing; Qiu Xiaoming

    1993-01-01

    The energy balance in TCA tokamak plasma with Alfven wave heating is studied, in which the equivalent electron thermal conductivity is determined by using the profile consistency principle. The results are in good agreement with experiments. It is shown that this method is applicable to various devices and other heating methods

  19. Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.; Brooks, J.; Mattas, R.

    1980-01-01

    A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification

  20. High-power heating experiment of spherical tokamaks by use of plasma merging

    International Nuclear Information System (INIS)

    Ueda, Yoshinobu; Ono, Yasushi

    1999-01-01

    High-power heating of spherical tokamaks (STs) has been investigated experimentally by use of plasma merging effect. When two STs were coaxially collided, thermal energy of a colliding ST was injected into a target ST during short reconnection time (Alfven time). Though the thermal energy increment increased with decreasing plasma q value, thermal energy loss during the following relaxation, tended to be smaller with increasing q. The produced high-β STs had hallower current profiles and weaker paramagnetic toroidal field than those of single STs. Those heating properties indicate the plasma merging to be a promising initial heating method of ST plasmas. (author)

  1. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

    During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

  2. 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).

  3. 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)

  4. 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

  5. Tokamak-7 operation in experiments with a plasma

    International Nuclear Information System (INIS)

    Buzanki, V.V.; Bychkov, A.V.; Denisov, V.F.

    1982-01-01

    The results of experiments with plasma at the Tokamak-7 (T-7) device are presented. The experiments have been carried out with a constant diaphragm of 31,5 cm radius and two movable graphite diaphragms at the 26-28 cm plasma filament radius and 1,6-1,9 T magnetic field. Two stable regimes with 150 and 200 kA and 250 ms discharge current length have been investigated. It is shown that the strongest poloidal filed perturhations have been observed at the beginning of the discharge. Electron plasma temperature Tsub(e) has been determined from the spectrum analysis of soft X radiation by the foil method. Stable plasma regimes with current up to 200 kA, bypass voltage being equal 1,58V electron density -0,5-5,0 x 10 13 cm -3 , Tsub(e)=1,1-1,3 keV ion temperature-490 eV. The range between discharge pulses has reached 3 min. at the discharge current-240 kA. No considerable effect of magnetic field variables on the superconducting magnetic system has been observed

  6. 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)

  7. Experimental device for the X-ray energetic distribution measurement in a tokamak plasma

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1977-01-01

    An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author) [es

  8. Plasma behavior with molecular beam injection in the HL-1m tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Tang Nianyi; Cui Zhengying; Xu Deming; Deng Zhongchao; Ding Xuantong; Luo Junlin; Dong Jiafu; Guo Gancheng; Yang Shikun; Cui Chenghe; Xiao Zhenggui; Liu Dequan; Chen Xiaoping; Yan Longwen; Yan Donghai; Wang Enyao; Deng Xiwen

    1999-01-01

    The authors report effect of the new fueling method of high speed molecular beam injection on Tokamak confinement improvement. The present method is an improvement of conventional gas puffing, with performance comparable to the small pellet injection in HL-1M and also to the slow pellet in ASDEX. The fact that a shallower fueling can lead to similar confinement improvement as a deep one suggests that there may exist a critical position in a Tokamak plasma such that any kind of fueling will have a better confinement as long as it can give rise to density peaking at the critical position

  9. 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

  10. Role of turbulence and electric fields in the establishment of improved confinement in tokamak plasmas

    Czech Academy of Sciences Publication Activity Database

    Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.

    2006-01-01

    Roč. 12, č. 6 (2006), s. 14-19 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html

  11. Selected methods of electron-and ion-diagnostics in tokamak scrape-off-layer

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This invited paper considers reasons why exact measurements of fast electron and ion losses in tokamaks, and particularly i n a scrape-off-layer and near a divertor region, are necessary in order to master nuclear fusion energy production. Attention is also paid to direct measurements of escaping fusion products from D-D and D-T reactions, and in particular of fast alphas which might be used for plasma heating. The second part describes the generation of so-called runaway and ripple-born electrons which might induce high energy losses and cause severe damages of internal walls in fusion facilities. Advantages and disadvantages of different diagnostic methods applied for studies of such fast electrons are discussed. Particular attention is paid to development of a direct measuring technique based on the Cherenkov effect which might be induced by fast electrons in appropriate radiators. There are presented various versions of Cherenkov-type probes which have been developed by the NCBJ team and applied in different tokamak experiments. The third part is devoted to direct measurements of fast ions (including those produced by the nuclear fusion reactions which can escape from a high-temperature plasma region. Investigation of fast fusion-produced protons from tokamak discharges is reported. New ion probes, which were developed by the NCBJ team, are also presented. For the first time there is given a detailed description of an ion pinhole camera, which enables irradiation of several nuclear track detectors during a single tokamak discharge, and a miniature Thomson-type mass-spectrometer, which can be used for ion measurements at plasma borders.

  12. Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

    Science.gov (United States)

    Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

    2012-01-10

    Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

  13. 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

  14. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  15. 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

  16. 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

  17. Electron density and temperature determination in a Tokamak plasma using light scattering; Determinacion de la densidad y temperatura electronicas en un Tokamak mediante difusion luminosa

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Navarro Gomerz, A; Zurro Hernandez, B

    1976-07-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs.

  18. Confinement of ohmically heated plasmas and turbulent heating in high-magnetic field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

    Hiraki, N; Itoh, S; Kawai, Y; Toi, K; Nakamura, K [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

    1979-12-01

    TRIAM-1, the tokamak device with high toroidal magnetic field, has been constructed to establish the scaling laws of advanced tokamak devices such as Alcator, and to study the possibility of the turbulent heating as a further economical heating method of the fusion oriented plasmas. The plasma parameters obtained by ohmic heating alone are as follows; central electron temperature T sub(e0) = 640 eV, central ion temperature T sub(i0) = 280 eV and line-average electron density n average sub(e) = 2.2 x 10/sup 14/ cm/sup -3/. The empirical scaling laws are investigated concerning T sub(e0), T sub(i0) and n average sub(e). The turbulent heating has been carried out by applying the high electric field in the toroidal direction to the typical tokamak discharge with T sub(i0) asymptotically equals 200 eV. The efficient ion heating is observed and T sub(i0) attains to about 600 eV.

  19. 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)

  20. 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)

  1. ICRF Mode Conversion Current Drive for Plasma Stability Control in Tokamaks

    International Nuclear Information System (INIS)

    Grekov, D.; Kock, R.; Lyssoivan, A.; Noterdaeme, J. M.; Ongena, J.

    2007-01-01

    There is a substantial incentive for the International Thermonuclear Experimental Reactor (ITER) to operate at the highest attainable beta (plasma pressure normalized to magnetic pressure), a point emphasized by requirements of attractive economics in a reactor. Recent experiments aiming at stationary high beta discharges in tokamak plasmas have shown that maximum achievable beta value is often limited by the onset of instabilities at rational magnetic surfaces (neoclassical tearing modes). So, methods of effective control of these instabilities have to be developed. One possible way for neoclassical tearing modes control is an external current drive in the island to locally replace the missing bootstrap current and thus to suppress the instability. Also, a significant control of the sawtooth behaviour was demonstrated when the magnetic shear was modified by driven current at the magnetic surface where safety factor equals to 1. In the ion cyclotron range of frequencies (ICRF), the mode conversion regime can be used to drive the local external current near the position of the fast-to-slow wave conversion layer, thus providing an efficient means of plasma stability control. The slow wave energy is effectively absorbed in the vicinity of mode conversion layer by electrons with such parallel to confining magnetic field velocities that the Landau resonance condition is satisfied. For parameters of present day tokamaks and for ITER parameters the slow wave phase velocity is rather low, so the large ratio of momentum to energy content would yield high current drive efficiency. In order to achieve noticeable current drive effect, it is necessary to create asymmetry in the ICRF power absorption between top and bottom parts of the plasma minor cross-section. Such asymmetric electron heating may be realized using: - shifted from the torus midplane ICRF antenna in TEXTOR tokamak; - plasma displacement in vertical direction that is feasible in ASDEX-Upgrade; - the

  2. 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)

  3. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  4. Identification and control of plasma vertical position using neural network in Damavand tokamak

    International Nuclear Information System (INIS)

    Rasouli, H.; Rasouli, C.; Koohi, A.

    2013-01-01

    In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg–Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant.

  5. Identification and control of plasma vertical position using neural network in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Rasouli, H. [School of Plasma Physics and Nuclear Fusion, Institute of Nuclear Science and Technology, AEOI, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Advanced Process Automation and Control (APAC) Research Group, Faculty of Electrical Engineering, K.N. Toosi University of Technology, P.O. Box 16315-1355, Tehran (Iran, Islamic Republic of); Rasouli, C.; Koohi, A. [School of Plasma Physics and Nuclear Fusion, Institute of Nuclear Science and Technology, AEOI, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

    2013-02-15

    In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg-Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant.

  6. Studies of feedback stabilization of axisymmetric modes in deformable tokamak plasmas

    International Nuclear Information System (INIS)

    Ward, D.J.

    1991-01-01

    A new linear MHD stability code, NOVA-W, is described and applied to the study of the feedback stabilization of the axisymmetric mode in deformable tokamak plasma. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The code has been tested for the case of passive stabilization against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. The NOVA-W code is used to examine the effects of plasma deformability on feedback stabilization. It is seen that plasmas with shaped cross sections have unstable motion different from a rigid shift. Plasma equilibria with large triangularity show particularly significant deviations from a uniform rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the motion in a way that reduces the stabilizing effects of these conductors. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops. These non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations

  7. 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

  8. Magnetic sensorless control of plasma position and shape in a tokamak

    International Nuclear Information System (INIS)

    Nakamura, K.; Luo, J.R.; Wang, H.Z.

    2005-01-01

    Magnetic sensorless sensing and control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of control experiments were carried out, to keep the position constant and swing the position in a triangular waveform. And magnetic sensorless sensing of plasma shape is discussed. (author)

  9. Characteristics of disruptive plasma current decay in the HT-2 tokamak

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Takeuchi, Kazuhiro; Otsuka, Michio

    1993-01-01

    Motions of plasma current channel and time evolutions of eddy current distribution on the vacuum vessel during disruptive plasma current decay were studied experimentally in the Hitachi tokamak HT-2. The plasmas are vertically elongated and circularly shaped plasmas. A disruptive plasma current decay has three phases. During the first phase, a large displacement of the plasma position without plasma current decay is observed. Rapid plasma current decay is observed during the second phase and the decay rate is roughly constant with time. The eddy current distribution is like that due to the shell effect which creates a poloidal field to reduce the plasma displacement. During the third phase, the plasma current decays exponentially. The second phase is observed in slightly elongated and high plasma current (> 20 kA) circularly shaped plasmas. The plasma current decay rates in the second phase depend on the plasma cross sectional shape, but they do not in the third phase. The magnetic axis moves from the plasma area to the vacuum vessel wall between the second and third phases. (author)

  10. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  11. Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Doyle, E J [University of California, Los Angeles, CA 90095-1597 (United States); Austin, M E [University of Texas at Austin, Austin, TX 78712 (United States); DeGrassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Gohil, P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Greenfield, C M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Jayakumar, R [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Kaplan, D H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lao, L L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); McKee, G R [University of Wisconsin, Madison, WI 53706-1687 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Rhodes, T L [University of California, Los Angeles, CA 90095-1597 (United States); Wade, M R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wang, G [University of California, Los Angeles, CA 90095-1597 (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Zeng, L [University of California, Los Angeles, CA 90095-1597 (United States)

    2004-05-01

    H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the {beta} limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, E{sub r}, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge E{sub r} well. The E{sub r} at the bottom of the well in these cases is about -100 kV m{sup -1} compared with -20 to -30 kV m{sup -1} in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal {beta} and {nu}{sup *} values matching the values desired for ITER have been achieved. In

  12. 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

  13. Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent

    2007-01-01

    Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

  14. 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

  15. 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

  16. 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)

  17. 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

  18. Development of a six channel Fabry-Perot interferometer for continuous measurement of electron temperature of Tokamak plasma. Application to current diffusion study

    International Nuclear Information System (INIS)

    Talvard, M.

    1984-10-01

    It is shown how the properties of the electron cyclotron emission of a tokamak plasma can be used to measure the electron temperature. The design of a six channel Fabry-Perot interferometer is then described. This interferometer allows the measurement of the time evolution of the electron temperature profile of the plasma in the TFR tokamak. Using this technique interesting results have been obtained concerning the current penetration during the start up phase of a tokamak discharge [fr

  19. 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

  20. Development of fast video recording of plasma interaction with a lithium limiter on T-11M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, V.B., E-mail: v_lazarev@triniti.ru [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Dzhurik, A.S.; Shcherbak, A.N. [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Belov, A.M. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

    2016-11-15

    Highlights: • The paper presents the results of the study of tokamak plasma interaction with lithium capillary-porous system limiters and PFC by high-speed color camera. • Registration of emission near the target in SOL in neutral lithium light and e-folding length for neutral Lithium measurements. • Registration of effect of MHD instabilities on CPS Lithium limiter. • A sequence of frames shows evolution of lithium bubble on the surface of lithium limiter. • View of filament structure near the plasma edge in ohmic mode. - Abstract: A new high-speed color camera with interference filters was installed for fast video recording of plasma-surface interaction with a Lithium limiter on the base of capillary-porous system (CPS) in T-11M tokamak vessel. The paper presents the results of the study of tokamak plasma interaction (frame exposure time up to 4 μs) with CPS Lithium limiter in a stable stationary phase, unstable regimes with internal disruption and results of processing of the image of the light emission around the probe, i.e. e-folding length for neutral Lithium penetration and e-folding length for Lithium ion flux in SOL region.

  1. 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)

  2. 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

  3. 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)

  4. Preliminary experiment of non-induced plasma current startup on SUNIST spherical tokamak

    International Nuclear Information System (INIS)

    He Yexi; Zhang Liang; Xie Lifeng; Tang Yi; Yang Xuanzong; Fu Hongjun

    2005-01-01

    Non-inductive plasma current startup is an important motivation on the SUNIST spherical tokamak. In this experiment, a 100 kW, 2.45 GHz magnetron microwave system has been applied to the plasma current startup. Besides the toroidal field, a vertical field was applied to generate a preliminary toroidal plasma current without action of the central solenoid. As the evidence of the plasma current startup by the vertical field drift effect, the direction of the plasma current is changed with the changing direction of the vertical field during ECR startup discharge. We have also observed the plasma current maximum by scanning the vertical field in both directions. Additionally, we have used electrode discharge to assist the ECR current startup. (author)

  5. Poloidal rotation induced by injecting lower hybrid waves in tokamak plasma edge

    International Nuclear Information System (INIS)

    Jiao Yiming; Gao Qingdi; Shi Bingren

    2001-01-01

    The poloidal rotation of the magnetized edge plasma in tokamak driven by the ponderomotive force which is generated by injecting lower hybrid wave (LHW) electric field has been studied. The LHW is launched from a waveguide in the plasma edge, and by Brambilla's grill theory, analytic expressions for the wave electric field in the slab model of an inhomogeneous cold plasma have been derived. It is shown that a strong wave electric field will be generated in the plasma edge by injecting LH wave of the power in MW magnitude, and this electric field will induce a poloidal rotation with a sheared poloidal velocity

  6. Neutral molecules in tokamak edge plasma - role of vibrationally excited hydrogen molecules

    International Nuclear Information System (INIS)

    Cadez, I.; Cercek, M.; Pelicon, P.; Razpet, A.

    2003-01-01

    The role of neutral molecules in edge plasma is discussed with special emphasis on the vibrationally excited hydrogen. Neutral molecules are formed mostly by surface processes on the walls and then released to the edge plasma where they take part in volumetric reactions with other particles. Typically these molecules are formed in excited states and data are needed for their reactions on the wall and in the volume. Processes in edge plasma determine particle and energy flux what is especially critical issue in tokamak divertor region. Various cross sections and reaction rates are needed for modelling edge plasma and its interaction with walls. (author)

  7. Experimental observation of current generation by asymmetrical heating of ions in a tokamak plasma

    International Nuclear Information System (INIS)

    Gahl, J.; Ishihara, O.; Wong, K.L.; Kristiansen, M.; Hagler, M.

    1986-01-01

    The first experimental observation of current generation by asymmetrical heating of ions is reported. Ions were asymmetrically heated by a unidirectional fast Alfven wave launched by a slow wave antenna inside a tokamak. Current generation was detected by measuring the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column

  8. 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)

  9. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

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

    Czech Academy of Sciences Publication Activity Database

    Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan

    96-97, October (2015), s. 974-979 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044

  11. Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

    International Nuclear Information System (INIS)

    Isayama, Akihiko; Tanaka, Hitoshi

    2009-01-01

    Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

  12. Losses of runaway electrons in MHD-active plasmas of the COMPASS tokamak.

    Czech Academy of Sciences Publication Activity Database

    Ficker, Ondřej; Mlynář, Jan; Vlainic, Milos; Čeřovský, Jaroslav; Urban, Jakub; Vondráček, Petr; Weinzettl, Vladimír; Macúšová, Eva; Decker, J.; Gospodarczyk, M.; Martin, P.; Nardon, E.; Papp, G.; Plyusnin, V.V.; Reux, C.; Saint-Laurent, F.; Sommariva, C.; Cavalier, Jordan; Havlíček, Josef; Havránek, Aleš; Hronová-Bilyková, Olena; Imríšek, Martin; Markovič, Tomáš; Varju, Jozef; Papřok, Richard; Pánek, Radomír; Hron, Martin

    2017-01-01

    Roč. 57, č. 7 (2017), č. článku 076002. ISSN 0029-5515 R&D Projects: GA MŠk LG14002; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamaks * runaway electrons * MHD instabilities * disruptions Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016

  13. Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Berta, Miklós; Anda, G.; Bencze, A.; Dunai, D.; Háček, Pavel; Hron, Martin; Kovácsik, A.; Krbec, Jaroslav; Pánek, Radomír; Réfy, D.; Véres, G.; Weinzettl, Vladimír; Zoletnik, S.

    96-97, October (2015), s. 795-798 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : BES * plasma diagnostics * COMPASS tokamak * density fluctuations * plasma density profile Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615300442

  14. 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)

  15. Influence of the helical resonant fields on the plasma potential in the TBR-1 Tokamak

    International Nuclear Information System (INIS)

    Ribeiro, C.; Silva, R.P. da; Caldas, I.L.; Fagundes, A.N.; Sanada, E.K.

    1990-01-01

    This work describes an experimental work that are in progress in TBR-1 tokamak about the influence of resonant helical fields on the plasma potential. TBR-1 is a small tokamak in operation in the Physics Institute of University of Sao Paulo and used for basic research, diagnostic development and personal formation. Its main parameters are: R(Major Radius) = 0.30 m; a v (Vessel Radius) = 0.11 m; a(Plasma Radius) = 0.08 m; R/a(Aspect Ratio) = 3.75; B φ (Toroidal Field) = 5 kG; n e0 (Central Electron Density) ≅ 7 x 10 18 m -3 ; T e0 (central electron temperature) ≅ 200 eV. (Author)

  16. 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)

  17. 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)

  18. Overview of Tokamak Results

    International Nuclear Information System (INIS)

    Unterberg, Bernhard; Samm, Ulrich

    2004-01-01

    An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

  19. 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

  20. Measurement of local, internal magnetic fluctuations via cross-polarization scattering in the DIII-D tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Barada, K., E-mail: kshitish@ucla.edu; Rhodes, T. L.; Crocker, N. A.; Peebles, W. A. [University of California-Los Angeles, P.O. Box 957099, Los Angeles, California 90095 (United States)

    2016-11-15

    We present new measurements of internal magnetic fluctuations obtained with a novel eight channel cross polarization scattering (CPS) system installed on the DIII-D tokamak. Measurements of internal, localized magnetic fluctuations provide a window on an important physics quantity that we heretofore have had little information on. Importantly, these measurements provide a new ability to challenge and test linear and nonlinear simulations and basic theory. The CPS method, based upon the scattering of an incident microwave beam into the opposite polarization by magnetic fluctuations, has been significantly extended and improved over the method as originally developed on the Tore Supra tokamak. A new scattering geometry, provided by a unique probe beam, is utilized to improve the spatial localization and wavenumber range. Remotely controllable polarizer and mirror angles allow polarization matching and wavenumber selection for a range of plasma conditions. The quasi-optical system design, its advantages and challenges, as well as important physics validation tests are presented and discussed. Effect of plasma beta (ratio of kinetic to magnetic pressure) on both density and magnetic fluctuations is studied and it is observed that internal magnetic fluctuations increase with beta. During certain quiescent high confinement operational regimes, coherent low frequency modes not detected by magnetic probes are detected locally by CPS diagnostics.

  1. 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.

  2. 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)

  3. 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

  4. Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)

    2016-11-01

    In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the Tokamak Simulation Code (TSC). With the numerical models and the source plasma current, the dynamic response of both the VV structure and the test module were calculated. The numerical results show that the maximum stress of the test module is in safe range, and the magnetic damping effect can weaken vibration of the test module. In addition, simulation without considering the coupling effect was carried out, which shows that the influence of coupling effect is not significant for the peak stress of the J-TEXT disruption problem.

  5. A dynamic state observer for real-time reconstruction of the tokamak plasma profile state and disturbances

    NARCIS (Netherlands)

    Felici, F.; De Baar, M.; Steinbuch, M.

    2014-01-01

    A dynamic observer is presented which can reconstruct the internal state of a tokamak fusion plasma, consisting of the spatial distribution of current and temperature, from measurements. Today, the internal plasma state is usually reconstructed by solving an ill-conditioned inversion problem using a

  6. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  7. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  8. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  9. Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh

    2012-01-01

    We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations...... with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated...... the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas...

  10. Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks

    International Nuclear Information System (INIS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-02-01

    Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)

  11. 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

  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. Generación y dinámica de electrones runaway en plasmas tokamak

    OpenAIRE

    Fernández Gómez, Isabel

    2016-01-01

    La dinámica y generación de electrones runaway en plasmas tokamak constituye el tema central de esta tesis. En un tokamak, el fenómeno runaway es el resultado de la existencia de un campo eléctrico en dirección toroidal. Aquellos electrones cuya velocidad excede un cierto valor crítico se aceleran de forma continua, ya que la e ciencia de las colisiones para disipar la energía ganada en el campo disminuye con la velocidad (∼ ⁻¹) . Se tiene entonces lo que se conoce como un electrón runaway. ...

  14. Observation of magnetic field perturbations during sawtooth activity in tokamak plasmas

    International Nuclear Information System (INIS)

    Soltwisch, H.; Koslowski, H.R.

    1997-01-01

    Sawtooth activity is a prominent example of a global plasma instability which is observed in virtually all tokamak devices. Despite numerous experimental and theoretical investigations, the phenomenon is still barely understood. As far as experimental effort is concerned, much attention has been paid to soft X-ray emission from the plasma and to its analysis in terms of two-dimensional contour plots, because it is thought to reflect the shape and temporal behaviour of magnetic flux surfaces during a sawtooth cycle. Recently, more direct methods of detecting sawtooth-related changes in the magnetic field structure have become available and have added new facets to the general picture. In this picture, some observations made on the Juelich tokamak TEXTOR by means of a Faraday rotation diagnostic technique will be reported. First, in correlation with the sawtooth collapse a localized periodic perturbation of the magnetic field with principal mode numbers m = 1 and n = 0 has been detected which, in the presence of an m = n = 1 island, may give rise to magnetic field line stochastization and thereby contribute significantly to a rapid expulsion of electronic energy from the plasma core region. Second, the so-called precursor oscillations prior to a sawtooth crash have been investigated and estimates have been obtained for the growth rate and width of a magnetic island forming immediately before the collapse. (Author)

  15. 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)

  16. Pellet-plasma interaction: Local disturbances caused by pellet ablation in tokamaks

    International Nuclear Information System (INIS)

    Lengyel, L.L.

    1989-01-01

    The local disturbance amplitudes caused by ablating pellets in tokamaks are computed in the framework of a magnetohydrodynamic model supplemented by the neutral gas plasma shielding ablation model. The model computes, for a given number of pellet particles locally deposited, the time histories of the ablatant cloud parameters, such as cloud radius, cloud length, electron density, temperature and cloud beta, at a succession of magnetic flux surfaces. The cloud radius thus determined may be fed back into the ablation model, thus adjusting the effect of the shielding cloud on the ablation rate. The model is applied to typical plasma parameter ranges of existing and planned tokamaks. The results show that the ablating pellets may cause massive local disturbances in tokamaks, depending upon the number of particles locally deposited. The peaks of these disturbances are of a spike nature, lasting only a few microseconds (Alfven time-scale). The characteristic decay time of the 'quasi-steady' disturbance values that characterize the after-spike period is of the order of several milliseconds (hydrodynamic time-scale). The peak electron density values may be as high as 10 23 to 10 25 m -3 , with the associated beta peaks exceeding unity. The 'quasi-steady' values of the electron density and the ablatant beta may be of the order of 10 22 to 10 24 m -3 and unity, respectively. Furthermore, the results show the strong dependence of the ablation rate on the dynamic characteristics of the ablatant cloud surrounding the pellet. (author). 25 refs, 6 figs, 2 tabs

  17. 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

  18. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    International Nuclear Information System (INIS)

    Gomes, R.B.; Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I.; Lielausis, O.; Klyukin, A.; Platacis, E.

    2011-01-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  19. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, R.B., E-mail: gomes@ipfn.ist.utl.pt [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Lielausis, O.; Klyukin, A.; Platacis, E. [Association EURATOM/University of Latvia, Institute of Solid State Physics, 8 Kengaraga Str., LV-1063 Riga (Latvia)

    2011-08-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  20. Feedback stabilization of the axisymmetric instability of a deformable tokamak plasma

    International Nuclear Information System (INIS)

    Pomphrey, N.; Jardin, S.C.; Ward, D.J.

    1989-01-01

    The paper presents an analysis of the magnetohydrodynamic stability of the axisymmetric system consisting of a free boundary tokamak plasma with non-circular cross-section, finite resistivity passive conductors, and an active feedback system with magnetic flux pickup loops, a proportional amplifier with gain G and current carrying poloidal field coils. A numerical simulation of the system when G is set to zero identifies flux loop locations which correctly sense the plasma motion. However, when certain of these locations are incorporated into an active feedback scheme, the plasma fails to be stabilized, no matter what value of the gain is chosen. Analysis on the basis of an extended energy principle indicates that this failure is due to the deformability of the plasma cross-section. (author). 14 refs, 7 figs

  1. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2011-01-01

    The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

  2. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  3. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

  4. Effect of neoclassical toroidal viscosity on error-field penetration thresholds in tokamak plasmas.

    Science.gov (United States)

    Cole, A J; Hegna, C C; Callen, J D

    2007-08-10

    A model for field-error penetration is developed that includes nonresonant as well as the usual resonant field-error effects. The nonresonant components cause a neoclassical toroidal viscous torque that keeps the plasma rotating at a rate comparable to the ion diamagnetic frequency. The new theory is used to examine resonant error-field penetration threshold scaling in Ohmic tokamak plasmas. Compared to previous theoretical results, we find the plasma is less susceptible to error-field penetration and locking, by a factor that depends on the nonresonant error-field amplitude.

  5. Effect of ECRH and resonant magnetic fields on formation of magnetic islands in the T-10 tokamak plasma

    Science.gov (United States)

    Shestakov, E. A.; Savrukhin, P. V.

    2017-10-01

    Experiments in the T-10 tokamak demonstrated possibility of controlling the plasma current during disruption instability using the electron cyclotron resonance heating (ECRH) and the controlled operation of the ohmic current-holding system. Quasistable plasma discharge with repeating sawtooth oscillations can be restored after energy quench using auxiliary ECRH power when PEC / POH > 2-5. The external magnetic field generation system consisted of eight saddle coils that were arranged symmetrically relative to the equatorial plane of the torus outside of the vacuum vessel of the T-10 tokamak to study the possible resonant magnetic field effects on the rotation frequency of magnetic islands. The saddle coils power supply system is based on four thyristor converters with a total power of 300 kW. The power supply control system is based on Siemens S7 controllers. As shown by preliminary experiments, the interaction efficiency of external magnetic fields with plasma depends on the plasma magnetic configuration. Optimal conditions for slowing the rotation of magnetic islands were determined. Additionally, the direction of the error magnetic field in the T-10 tokamak was determined, and the threshold value of the external magnetic field was determined.

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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#

  11. Current density distribution during disruptions and sawteeth in a simple model of plasma current in a tokamak

    International Nuclear Information System (INIS)

    Stefanovskii, A. M.

    2011-01-01

    The processes that are likely to accompany discharge disruptions and sawteeth in a tokamak are considered in a simple plasma current model. The redistribution of the current density in plasma is supposed to be primarily governed by the onset of the MHD-instability-driven turbulent plasma mixing in a finite region of the current column. For different disruption conditions, the variation in the total plasma current (the appearance of a characteristic spike) is also calculated. It is found that the numerical shape and amplitude of the total current spikes during disruptions approximately coincide with those measured in some tokamak experiments. Under the assumptions adopted in the model, the physical mechanism for the formation of the spikes is determined. The mechanism is attributed to the diffusion of the negative current density at the column edge into the zero-conductivity region. The numerical current density distributions in the plasma during the sawteeth differ from the literature data.

  12. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  13. A review of the methods to measure the ion temperature in a tokamak plasma

    International Nuclear Information System (INIS)

    Zurro Hernandez, B.; Perez-Navarro Gomez, A.

    1976-01-01

    The most important methods to measure the ion temperatu--re in a Tokamak plasma are reviewed, e.g. energy analysis of the fast neutrals which leave out the plasma, Doppler broadening of the emision spectral lines and fusion neutron analysis. It is discussed their bounds so as the advantages and drawbacks of each one. Other methods of some interest in the future are outlined. (author) [es

  14. 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

  15. MHD instabilities in astrophysical plasmas: very different from MHD instabilities in tokamaks!

    Science.gov (United States)

    Goedbloed, J. P.

    2018-01-01

    The extensive studies of MHD instabilities in thermonuclear magnetic confinement experiments, in particular of the tokamak as the most promising candidate for a future energy producing machine, have led to an ‘intuitive’ description based on the energy principle that is very misleading for most astrophysical plasmas. The ‘intuitive’ picture almost directly singles out the dominant stabilizing field line bending energy of the Alfvén waves and, consequently, concentrates on expansion schemes that minimize that contribution. This happens when the wave vector {{k}}0 of the perturbations, on average, is perpendicular to the magnetic field {B}. Hence, all macroscopic instabilities of tokamaks (kinks, interchanges, ballooning modes, ELMs, neoclassical tearing modes, etc) are characterized by satisfying the condition {{k}}0 \\perp {B}, or nearly so. In contrast, some of the major macroscopic instabilities of astrophysical plasmas (the Parker instability and the magneto-rotational instability) occur when precisely the opposite condition is satisfied: {{k}}0 \\parallel {B}. How do those instabilities escape from the dominance of the stabilizing Alfvén wave? The answer to that question involves, foremost, the recognition that MHD spectral theory of waves and instabilities of laboratory plasmas could be developed to such great depth since those plasmas are assumed to be in static equilibrium. This assumption is invalid for astrophysical plasmas where rotational and gravitational accelerations produce equilibria that are at best stationary, and the associated spectral theory is widely, and incorrectly, believed to be non-self adjoint. These complications are addressed, and cured, in the theory of the Spectral Web, recently developed by the author. Using this method, an extensive survey of instabilities of astrophysical plasmas demonstrates how the Alfvén wave is pushed into insignificance under these conditions to give rise to a host of instabilities that do not

  16. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  17. Critical condition for current-driven instability excited in turbulent heating of TRIAM-1 tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Y; Watanabe, T; Nagao, A; Nakamura, K; Kikuchi, M; Aoki, T; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Mitarai, O

    1982-02-01

    Critical condition for current-driven instability excited in turbulently heated TRIAM-1 tokamak plasma is investigated experimentally. Resistive hump in loop voltage, plasma density fluctuation and rapid increase of electron temperature in a skin layer are simultaneously observed at the time when the electron drift velocity amounts to the critical drift velocity for low-frequency ion acoustic instability.

  18. A study of quasi-mode parametric excitations in lower-hybrid heating of tokamak plasmas

    International Nuclear Information System (INIS)

    Villalon, E.; Bers, A.

    1980-01-01

    A detailed linear and non-linear analysis of quasi-mode parametric excitations relevant to experiments in supplementary heating of tokamak plasmas is presented. The linear analysis includes the full ion-cyclotron harmonic quasi-mode spectrum. The non-linear analysis, considering depletion of the pump electric field, is applied to the recent Alcator A heating experiment. Because of the very different characteristics of a tokamak plasma near the wall (in the shadow of the limiter) and inside, the quasi-mode excitations are studied independently for the plasma edge and the main bulk of the plasma, and for two typical regimes in overall density, the low (peak in density, n 0 =1.5x10 14 cm -3 ) and high (n 0 =5x10 14 cm -3 ) density regimes. At the edge of the plasma and for the low-density regime, it is found that higher nsub(z)(nsub(z)=cksub(z)/ω) than those predicted by the linear theory are strongly excited. Inside the plasma, the excitation of higher wave numbers is also significant. These results indicate that a large amount of the RF-power may not penetrate to the plasma centre, but will rather be either Landau-damped on the electrons or mode-converted into thermal modes, close to the plasma edge. Moreover, for sufficiently high peaks in density, it is found that all the RF-power is mode-converted before reaching the plasma centre. Inside the plasma, the power density of the excited sideband fields is shown to be always very small in comparison with their excitation at the plasma edge. (author)

  19. Study of the linear and non-linear coupling of the LH wave to the tokamak plasmas

    International Nuclear Information System (INIS)

    Preynas, M.

    2012-10-01

    In order to achieve long pulse operation with a tokamak, additional heating and current drive systems are necessary. High frequency antennas, which deliver several megawatts of power to the plasma, are currently used in several tokamaks. Moreover, a good control of the coupling of the wave launched by the antenna to the edge plasma is required to optimize the efficiency of heating and current drive LH systems. However, non-linear effects which depend on the level of injected power in the plasma strongly damage the coupling of the LH wave at particular edge parameters (density and temperature profiles). Results presented in the manuscript deal with the study of the linear and non-linear coupling of the LH wave to the plasma. In the framework of the commissioning of the Passive Active Multijunction antenna in 2009 on the Tore Supra tokamak aiming at validating the LH system suggested for ITER, the characterisation of its coupling properties was realized from low power experiments. The experimental results, which are compared with the linear coupling code ALOHA, have validated the theoretical predictions of good coupling at edge plasma density around the cut-off density. Besides, the ponderomotive effect is clearly identified as responsible for the deterioration in the coupling of the wave, which is measured under particular edge plasma conditions. A theoretical model combining the coupling of the LH wave with the ponderomotive force is suggested to explain the experimental observations. Thus, a new full wave code (named PICCOLO-2D) was developed and results from simulations validate the working hypothesis of the contribution of the ponderomotive effect in the non-linear observations of LHCD coupling on Tore Supra. (author)

  20. Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Anda, G.; Bencze, A.; Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Háček, P., E-mail: hacek@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Kovácsik, A. [Wigner – RCP, HAS, Budapest (Hungary); Department of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Pánek, R. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Réfy, D.; Veres, G. [Wigner – RCP, HAS, Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2015-10-15

    Highlights: • Li-BES detection system on the COMPASS tokamak is optimized observation system with high temporal resolution. • High sensitivity to low level light fluctuations. • Optics and detectors with electronics are placed in thermally stabilized compact box. • Fast deflection system allows us to measure background corrected electron density profiles on microsecond time-scale. - Abstract: A new Li beam emission spectroscopy (Li-BES) diagnostic system with a ∼ cm spatial resolution, and with beam energy ranging from 10 keV up to 120 keV and a 18 channel Avalanche photo diode (APD) detector system sampled at 2 MHz has been recently installed and tested on the COMPASS tokamak. This diagnostic allows to reconstruct density profile based on directly measured light profiles, and to follow turbulent behaviour of the edge plasma. The paper reports technical capabilities of this new system designed for fine spatio-temporal measurements of plasma electron density. Focusing on turbulence-induced fluctuation measurements, we demonstrate how physically relevant information can be extracted using the COMPASS Li-BES system.

  1. Computation of electromagnetic effects in a tokamak due to a plasma disruption

    International Nuclear Information System (INIS)

    Turner, L.R.

    1989-01-01

    To model the consequences of a plasma disruption in a tokamak one must combine a code that computes the detailed MHD behavior of the plasma with one that treats the three-dimensional features of the conducting toroidal components around the plasma. The NET (Next European Torus) Team have undertaken a treatment of electromagnetic effects from plasma disruptions using both open loop and closed loop integration of codes. In America, workers at Oak Ridge National Laboratory, Idaho National Engineering Laboratory, and Argonne National Laboratory have looked at plasma disruption effects on the ITER blanket using the codes TSC and EDDYNET. Results show how the forces on a blanket segment depend on the number and size of the segments and on the gap between them. 9 refs., 4 figs., 1 tab

  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. Development of the 'JFT-2' tokamak plasma position control system

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

    1980-01-01

    Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

  4. 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

  5. Interpretive modelling of scrape-off plasmas on the MAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); University of York, Heslington, York (United Kingdom)], E-mail: james.harrison@ukaea.org.uk; Lisgo, S. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Counsell, G.F. [Fusion for Energy, Barcelona (Spain); Gibson, K. [University of York, Heslington, York (United Kingdom); Dowling, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Trojan, L. [University of Manchester, Oxford Road, Manchester (United Kingdom); Reiter, D. [IPP, Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany)

    2009-06-15

    Electrical currents in the scrape-off layer (SOL) of MAST are modelled using an interpretive Onion-Skin Model (OSM) constrained with experimental data from MAST diagnostics. The model was extended to include the effects of the magnetic mirror force, which has a strong influence on the particle and momentum balance in spherical tokamaks, such as MAST . These modifications serve to more accurately model the parallel electric fields present in the MAST SOL, which can alter plasma dynamics via the E x B drift. Simulations show that the electrical current at the divertor targets is predominantly thermoelectric, whereas Pfirsch-Schlueter currents have a greater contribution to the total current in the bulk of the SOL plasma.

  6. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  7. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

  8. DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS

    International Nuclear Information System (INIS)

    LUCE, T.C.; WADE, M.R.; FERRON, J.R.; HYATT, A.W.; KELLMAN, A.G.; KINSEY, J.E.; LAHAY, R.J.; LASNIER, C.J.; MURAKAMI, M.; POLITZER, P.A.; SCOVILLE, J.T.

    2002-01-01

    OAK A271 DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak in stationary conditions with relatively low plasma current (q 95 > 4). A figure of merit for fusion gain Β N H 89 /q 95 2 has been maintained at values corresponding to Q = 10 operation in a burning plasma for > 6 s or 36 τ E and 2 τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1, which allows stable operation up to the no-wall ideal β limit. These plasmas maintain particle balance by active pumping rather than transient wall conditions. The reduced current lessens significantly the potential for structural damage in the event of a major disruption

  9. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  10. 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

  11. Thermonuclear-driven fast magnetosonic-wave heating in tokamak plasmas

    International Nuclear Information System (INIS)

    Sutton, W.R. III.

    1982-01-01

    A thermonuclear driven fast magnetosonic wave instability is investigated in tokamak plasmas for propagation transverse to the external magnetic field at frequencies of several times the alpha particle gyro rate: ω approx. = L Ω/sub α/ = k/sub perpendicular/ v/sub A/, L approx. 4 to 8, k/sub parallel/ << k/sub perpendicular/. The 2-D differential quasi-linear diffusion equation is derived in circular cylindrical, v/sub perpendicular/-v/sub parallel/ geometry. We perform an expansion in the small parameter k/sub parallel/k/sub perpendicucular/ of the quasi-linear diffusion coefficients

  12. 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

  13. Role of boundary plasma in lower-hybrid-frequency heating of a tokamak

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Yamamoto, Takumi; Fujii, Tsuneyuki

    1982-01-01

    Boundary plasma of a circular tokamak has been investigated by means of electrostatic probes during lower-hybrid heating. The reflection coefficient is affected by the density gradient in front of the launcher. An effective ion heating is performed in the main plasma region when the boundary electron temperature is relatively high enough to suppress the parametric decay instabilities. The simultaneous injection of neutral beams as well as the lower-hybrid wave brings the suppression of instabilities with increase of the electron temperature coming from the neutral beam heating. (author)

  14. 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.

  15. 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.)

  16. Frequency spectral broadening of lower hybrid waves in tokamak plasmas - causes and effects

    Energy Technology Data Exchange (ETDEWEB)

    Pericoli Ridolfini, V; Giannone, L.; Bartiromo, R [Associazione Euratom-ENEA sulla Fusione, Rome (Italy). Centro Ricerche Energia Frascati

    1994-04-01

    The frequency spectral broadening of lower hybrid (LH) waves injected into tokamak plasmas is extensively analyzed with reference mostly to experimental data from the ASDEX tokamak. The link between the magnitude of the pump spectral width and the degradation of the LH current drive efficiency (up to a factor of 2), pointed out in previous works, is explained. The experimental behaviour of LH power absorption is also well reproduced, even in situations when the access of the launched LH waves to the core plasma should be largely forbidden. Experiments are described that are aimed at determined whether parametric decay instabilities (PDIs) or scattering of LH waves by density fluctuations in the plasma edge are causes of the broadening of the LH pump frequency spectrum. Fluctuations emerge as the largely dominant process, while no signature of PDI processes is observed. Careful measurements of the density fluctuations in the ASDEX scrape-off layer plasma allow the analytical description given by Andrews and Perkins to be assumed as the appropriate model of LH scattering. Indeed, it supplies the correct magnitude for the frequency spectral width of the LH pump, and explains quantitatively, together with a ray tracing code, why the CD efficiency decreases with the broadening of the pump spectrum. It can also account for the observed LH power absorption coefficient. (author). 48 refs, 13 figs, 2 tabs.

  17. Structural instability of sheath potential distribution and its possible implications for the L/H transition in tokamak plasmas

    International Nuclear Information System (INIS)

    Yoshida, Zensho; Yamada, Hiroshi.

    1988-07-01

    The Bohm equation of electrostatic potential distributions in one-dimensional plasmas has been studied for various Mach numbers and plasma potentials. Solvability and structural stability have been discussed using the Sagdeev potential. Implications of the structural stability for the L/H transitions in tokamak plasmas has been also discussed. (author)

  18. 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

  19. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  20. 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)

  1. Strong Scattering of High Power Millimeter Waves in Tokamak Plasmas with Tearing Modes

    DEFF Research Database (Denmark)

    Westerhof, E.; Nielsen, Stefan Kragh; Oosterbeek, J.W.

    2009-01-01

    In tokamak plasmas with a tearing mode, strong scattering of high power millimeter waves, as used for heating and noninductive current drive, is shown to occur. This new wave scattering phenomenon is shown to be related to the passage of the O point of a magnetic island through the high power...

  2. High-resolution Thomson scattering system on the COMPASS tokamak: Evaluation of plasma parameters and error analysis

    Czech Academy of Sciences Publication Activity Database

    Aftanas, Milan; Böhm, Petr; Bílková, Petra; Weinzettl, Vladimír; Zajac, Jaromír; Žáček, František; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Walsh, M.

    2012-01-01

    Roč. 83, č. 10 (2012), 10E350-10E350 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] R&D Projects: GA ČR GA202/09/1467; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : error analysis * Monte Carlo methods * plasma density * plasma diagnostics * plasma temperature * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4743956

  3. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  4. Calculation of poloidal rotation in the edge plasma of limiter tokamaks

    International Nuclear Information System (INIS)

    Gerhauser, H.; Claassen, H.A.

    1987-05-01

    The existing 2-d two-fluid code for computing the plasma profiles in the scrape-off layer of limiter tokamaks has been further developed to include the effect of poloidal rotation in the basic equations. This rotation is produced by radial electric fields which arise in the limiter shadow due to radial gradients in the Langmuir sheath potential in front of the limiter. As a consequence slight deviations from ambipolar motion must occur. A strong increase of rotation near the separatrix is connected with an electric current circuit closed via the limiter edge. The 2-d profiles of all relevant quantities are calculated and discussed for TEXTOR-typical parameters including also the effect of limiter recycled neutrals. The results agree well with the known experimental evidence on poloidal rotation and should be transferable to all limiter tokamaks. (orig.)

  5. The analysis of Alfven wave current drive and plasma heating in TCABR tokamak

    International Nuclear Information System (INIS)

    Ruchko, L.F.; Lerche, E.A.; Galvao, R.M.O.; Elfimov, A.G.; Nascimento, I.C.; Sa, W.P. de; Sanada, E.; Elizondo, J.I.; Ferreira, A.A.; Saettone, E.A.; Severo, J.H.F.; Bellintani, V.; Usuriaga, O.N.

    2002-01-01

    The results of experiments on Alfven wave current drive and plasma heating in the TCABR tokamak are analyzed with the help of a numerical code for simulation of the diffusion of the toroidal electric field. It permits to find radial distributions of plasma current density and conductivity, which match the experimentally measured total plasma current and loop voltage changes, and thus to study the performance of the RF system during Alfven wave plasma heating and current drive experiments. Regimes with efficient RF power input in TCABR have been analyzed and revealed the possibility of noninductive current generation with magnitudes up to ∼8 kA. The increase of plasma energy content due to RF power input is consistent with the diamagnetic measurements. (author)

  6. An advanced plasma control system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J.; Lazarus, E.

    1991-11-01

    An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as β p , ell i and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 μs intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 μs

  7. Combined confinement system applied to tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, Tihiro

    1986-01-01

    From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

  8. 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

  9. 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.

  10. New fluctuation phenomena in the H-mode regime of PDX tokamak plasmas

    International Nuclear Information System (INIS)

    Slusher, R.E.; Surko, C.M.; Valley, J.F.; Crowley, T.; Mazzucato, E.; McGuire, K.

    1984-05-01

    A new kind of quasi-coherent fluctuation is observed near the edge of plasmas in the PDX tokamak during H-mode operation. (The H-mode occurs in neutral beam heated divertor plasmas and is characterized by improved energy containment as well as large density and temperature gradients near the plasma edge.) These fluctuations are evidenced as VUV and density fluctuation bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the frequency range between 50 and 180 kHz. They affect the edge temperature-density product, and therefore they may be important for understanding the relationship between the large edge density and temperature gradients and the improved energy confinement

  11. Multi-scale semi-ideal magnetohydrodynamics of a tokamak plasma

    International Nuclear Information System (INIS)

    Bazdenkov, S.; Sato, Tetsuya; Watanabe, Kunihiko.

    1995-09-01

    An analytical model of fast spatial flattening of the toroidal current density and q-profile at the nonlinear stage of (m = 1/n = 1) kink instability of a tokamak plasma is presented. The flattening is shown to be an essentially multi-scale phenomenon which is characterized by, at least, two magnetic Reynolds numbers. The ordinary one, R m , is related with a characteristic radial scale-length, while the other, R m * , corresponds to a characteristic scale-length of plasma inhomogeneity along the magnetic field line. In a highly conducting plasma inside the q = 1 magnetic surface, where q value does not much differ from unity, plasma evolution is governed by a multi-scale non-ideal dynamics characterized by two well-separated magnetic Reynolds numbers, R m and R m * ≡ (1 - q) R m , where R m * - O(1) and R m >> 1. This dynamics consistently explains two seemingly contradictory features recently observed in a numerical simulation [Watanabe et al., 1995]: i) the current profile (q-profile) is flattened in the magnetohydrodynamic time scale within the q = 1 rational surface; ii) the magnetic surface keeps its initial circular shape during this evolution. (author)

  12. Space-resolved vacuum ultra-violet spectroscopy on T.F.R. Tokamak plasmas

    International Nuclear Information System (INIS)

    1978-01-01

    Results are reported of space-resolved vacuum-ultraviolet spectroscopy (between 100 A and 2000A) on T.F.R. Tokamak plasmas and examples are given of profiles for both heavy and light impurity ions. The experimental method and the associated uncertainties and problems are stressed. The great importance of numerical calculations in the interpretation of the impurity profiles is pointed out. (author)

  13. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1996-01-01

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  14. 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

  15. 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)

  16. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

    Rebut, P.H.; Green, B.J.

    1977-01-01

    The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

  17. 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

  18. 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.

  19. Determination of the plasma column shape in the Tokamak Novillo cross section by magnetic probes

    International Nuclear Information System (INIS)

    Hernandez Omana, A.

    1993-01-01

    The determination of plasma cross section shape in Tokamaks is an important diagnostic method for equilibrium conditions analysis. In this work, it is obtained a time dependent variation of the plasma column cross section in Novillo Tokamak. The experimental method is based on using one magnetic probe, which is installed inside of the vacuum vessel in a 1 mm. wall thickness stainless steel tube, in the protected region of the limiter shadow. The plasma column cross section is determined measuring the poloidal magnetic field produced by the plasma current. This method, now running for determining the plasma column shape, requires the measurement of magnetic present field outside plasma column. The measurements are carried out from a set of small coils, which are located inside the vacuum chamber in the radial and poloidal direction, so we can measure magnetic field with no current attenuations produced by the penetration time of the stainless steel vacuum chamber. The magnetic probe detect a real time variation of magnetic flux passing through them. In order to obtain the magnetic field values, it is required that the electric signals coming from the magnetic probe be integrated, this operation is carried out by active circuits located between the probe signal and one oscilloscope. The integrated signals can be exhibited photographed on the oscilloscope display. (Author)

  20. 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)

  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.; 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)

  2. Plasma density determination by microwave interferometry. The 2 mm interferometer of the TJ-1 tokamak

    International Nuclear Information System (INIS)

    Manero, F.; Martin, R.

    1984-01-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density profiles from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (author)

  3. The O-X-B mode conversion scheme for ECRH of a high-density Tokamak plasma

    DEFF Research Database (Denmark)

    Hansen, F. R.; Lynov, Jens-Peter; Michelsen, Poul

    1985-01-01

    A method to apply electron cyclotron resonance heating (ECRH) to a Tokamak plasma with central density higher than the critical density for cut-off of the ordinary mode (O-mode) has been investigated. This method involves two mode conversions, from an O-mode via an extraordinary mode (X......-mode) into an electron Bernstein mode (B-mode). Radial profiles for the power deposition and the wave-drive current due to the B-waves are calculated for realistic antenna radiation patterns with parameters corresponding to the Danish DANTE Tokamak and to Princeton's PLT....

  4. Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak

    International Nuclear Information System (INIS)

    Martin, R.; Manero, F.

    1984-01-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs

  5. 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

  6. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  7. Real-time horizontal position control for Aditya-upgrade tokamak

    International Nuclear Information System (INIS)

    Kumar, Rohit; Ghosh, Joydeep; Tanna, Rakesh L.

    2015-01-01

    Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

  8. Improvement of tokamak performance by injection of electrons

    International Nuclear Information System (INIS)

    Ono, Masayuki.

    1992-12-01

    Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas

  9. Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak; Determinacion de la densidad de un plasma por interferometria de microondas. El interferometro de 2 mm del Tokamak TJ-1

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R; Manero, F

    1984-07-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs.

  10. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  11. Development of plasma current waveform adjusting system ZLJ for tokamak device HL-1

    International Nuclear Information System (INIS)

    Wang Shangbing; Hu Haotian; Tang Fangqun; Zhou Yongzheng; Chu Xiuzhong; Cheng Jiashun; Gao Yunxia

    1989-12-01

    The control of some typical Tokamak discharge waveforms has been achieved by using plasma current waveform adjusting system ZLJ in the ohmic heating of HL-1. The discharge waveforms include a series of regular plasma current waveforms with various slow rising rate, such as 80 kA, 450 ms long flat-topping; 100 kA, 200 ms rising; 200 ms falt-topping and 180 kA, 400 ms slow rising etc. The design principle of the system and the initial experimental results are described

  12. Electron cyclotron resonance hydrogen/helium plasma characterization and simulation of pumping in tokamaks

    International Nuclear Information System (INIS)

    Outten, C.A.

    1992-01-01

    Electron Cyclotron Resonance (ECR) plasmas have been employed to simulate the plasma conditions at the edge of a tokamak in order to investigate hydrogen/helium uptake in thin metal films. The process of microwave power absorption, important to characterizing the ECR plasma source, was investigated by measuring the electron density and temperature with a Langmuir probe and optical spectroscopy as a function of the magnetic field gradient and incident microwave power. A novel diagnostic, carbon resistance probe, provided a direct measure of the ion energy and fluence while measurements from a Langmuir probe were used for comparison. The Langmuir probe gave a plasma potential minus floating potential of 30 ± 5 eV, in good agreement with the carbon resistance probe result of ion energy ≤ 40 eV. The measured ion energy was consistent with the ion energy predicted from a model based upon divergent magnetic field extraction. Also, based upon physical sputtering of the carbon, the hydrogen fluence rate was determined to be 1 x 10 16 /cm 2 -sec for 50 Watts of incident microwave power. ECR hydrogen/helium plasmas were used to study preferential pumping of helium in candidate materials for tokamak pump-limiters: nickel, vanadium, aluminum, and nickel/aluminum multi-layers. Nickel and vanadium exhibited similar pumping capacities whereas aluminum showed a reduced capacity due to increased sputtering. A helium retention model based upon ion implantation ranges and sputtering rates agreed with the experimental data. A new multilayer/bilayer pumping concept showed improved pumping above that for single element films

  13. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

    Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.

    1987-01-01

    Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs

  14. 2D full-wave simulation of waves in space and tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Kim Eun-Hwa

    2017-01-01

    Full Text Available Simulation results using a 2D full-wave code (FW2D for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF waves in the scape-off layer (SOL of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  15. 2D full-wave simulation of waves in space and tokamak plasmas

    Science.gov (United States)

    Kim, Eun-Hwa; Bertelli, Nicola; Johnson, Jay; Valeo, Ernest; Hosea, Joel

    2017-10-01

    Simulation results using a 2D full-wave code (FW2D) for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry) and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC) waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF) waves in the scape-off layer (SOL) of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  16. 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.)

  17. Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

    The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  18. Joule Heating of Plasma in the Toroidal Tokamak-3 Device; Chauffage du Plasma par Effet Joule, dans l'Installation Torique; Dzhoulev nagrev plazmy na toroidal'noj ustanovke Tokamak-3; Calentamiento de un Plasma por Efeto Joule en la Instalacion Toroidal Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Arcimovich, L. A.; Afrosimov, V. V.; Gladkovskij, I. P.; Mirnov, S. V.; Petrov, M. P.; Strelkov, V. S. [Institut Atomnoj Ehnergii, Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

    1966-04-15

    In Tokamak-3, a hydrogen plasma is formed and heated by an annular electric current of 40 to 60 kA. The time of current flow is 20 to 30 ms. The bulk of the experiments were performed with a 25-kOe stabilizing longitudinal magnetic field. The transverse component of the stray magnetic field was compensated for by using special correcting loops. In the course of the discharge there was no appreciable displacement of the centre of the column from the equatorial plane of the torus, and we observed a plasma-column drift ''to the outside'' (increase of the large radius of the loop). This motion can be caused by a change in the radius of the current column, by plasma heating or by the damping of the Foucault currents in the conducting vessel. We succeeded in obtaining a macroscopically stable plasma column in Tokamak-3 under these conditions, but the plasma temperature was lower than should be expected for the case with no anomalous energy losses. The interaction processes between the plasma column and the diaphragm must lead to large energy losses. We succeeded in attenuating this interaction by applying a diaphragm of special shape and by using the property of the column to shift to the outside in the course of the process. The problem of the investigations was to determine various plasma parameters under these conditions, including electron and ion temperatures. To determine the electron temperature from the change of plasma resistance, it is necessary to know the column-radius changes with time, as well as the electric characteristics of the column. By using a computer to solve a set of equations, including the electrical engineering and the equilibrium equations, we can determine the time-dependence of changes in plasma temperature and density and also in the radius and displacement of the plasma, column, making use of experimentally measured time dependencies of plasma current, voltage and magnetic probe and radio interferometer signals. The computed temperature

  19. Sawtooth Pacing by Real-Time Auxiliary Power Control in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Goodman, T. P.; Felici, F.; Sauter, O.; Graves, J. P.

    2011-01-01

    In the standard scenario of tokamak plasma operation, sawtooth crashes are the main perturbations that can trigger performance-degrading, and potentially disruption-generating, neoclassical tearing modes. This Letter demonstrates sawtooth pacing by real-time control of the auxiliary power. It is shown that the sawtooth crash takes place in a reproducible manner shortly after the removal of that power, and this can be used to precisely prescribe, i.e., pace, the individual sawteeth. In combination with preemptive stabilization of the neoclassical tearing modes, sawtooth pacing provides a new sawtooth control paradigm for improved performance in burning plasmas.

  20. 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