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Sample records for continuous current tokamak

  1. 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

  2. Mechanical impacts of poloidal eddy currents on the continuous vacuum vessel of a tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo.

    1996-11-01

    Poloidal eddy currents are induced on the continuous torus vacuum vessel by changes of the toroidal field during the machine start-up (toroidal field coil charge), shut-down (toroidal field coil discharge) and plasma disruption (plasma diamagnetism change). Analytic forms for the eddy currents flowing on the vessel, consequent pressures and forces acting on it are presented in this report. The results are applied to typical operation modes of the KT-2 tokamak. Stress analysis for two typical operation modes of toroidal field damping during a machine shut-gown and plasma energy quench during a plasma disruption were carried out using 3D FEM code (ANSYS 5.2). (author). 5 tabs., 22 figs., 9 refs

  3. Natural current profiles in tokamaks

    International Nuclear Information System (INIS)

    Biskamp, D.

    1986-01-01

    It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)

  4. 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

  5. Operating tokamaks with steady-state toroidal current

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1981-04-01

    Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements

  6. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

    Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios

  7. Electric conductivity and bootstrap current in tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Wang Maoquan

    1996-12-01

    A modified Ohm's law for the electric conductivity calculation is presented, where the modified ohmic current can be compensated by the bootstrap current. A comparison of TEXT tokamak experiment with the theories shows that the modified Ohm's law is a more close approximation to the tokamak experiments than the classical and neoclassical theories and can not lead to the absurd result of Z eff <1, and the extended neoclassical theory would be not necessary. (3 figs.)

  8. Lower hybrid current drive in shaped tokamaks

    International Nuclear Information System (INIS)

    Kesner, J.

    1993-01-01

    A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

  9. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

    Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

  10. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1986-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500 to 700 keV are needed for this device

  11. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1987-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500-700 keV are needed for this device

  12. Noninductive current drive in tokamaks

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1985-01-01

    Various current drive mechanisms may be grouped into four classes: (1) injection of energetic particle beams; (2) launching of rf waves; (3) hybrid schemes, which are combinations of various rf schemes (rf plus beams, rf and/or beam plus ohmic heating, etc.); and (4) other schemes, some of which are specific to reactor plasma conditions requiring the presence of alpha particle or intense synchrotron radiation. Particle injection schemes include current drive by neutral beams and relativistic electron beams. The rf schemes include current drive by the lower hybrid (LH) waves, the electron waves, the waves in the ion cyclotron range of frequencies, etc. Only a few of these approaches, however, have been tested experimentally, with the broadest data base available for LH waves. Included in this report are (1) efficiency criteria for current drive, (2) current drive by neutral beam injection, (3) LH current drive, (4) electron cyclotron current drive, (5) current drive by ion cyclotron waves - minority species heating, and (6) current drive by other schemes (such as hybrids and low frequency waves)

  13. Continuous tokamak operation with an internal transformer

    International Nuclear Information System (INIS)

    Singer, C.E.; Mikkelsen, D.R.

    1982-10-01

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia timescales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors

  14. 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

  15. Study of the non inductive current generation in Tore Supra and application to the operational scenario of a continuous tokamak; Etude de la generation de courant non inductive dans Tore Supra et application aux scenarios operationnels d`un tokamak continu

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian-Vibert, F.

    1996-07-05

    Lower Hybrid Current Drive in tokamak plasmas allows to obtain continuous operations, which constitute a necessary step towards a definition of a thermonuclear fusion reactor. The objectives of this work is to define and study fully non inductive steady-state scenarios on Tore Supra. The current diffusion equation is solved to determined precisely the inductive and non inductive current density profiles and their influence on thee time evolution of a discharge. Then, a new operation mode is studied theoretically and experimentally. In this scenario, the transformer primary circuit voltage is controlled in such a way that the flux consumption vanishes. It allows to achieve full steady-state discharges in a fast and reproducible manner. A theoretical flux consumption scaling law during plasma current ramp-up assisted by Lower-Hybrid waves is presented and validated by experimental data, in view to minimized this consumption. The influence of a non monotonic current profile on the confinement and the transport of energy in the plasma is also clearly illustrated by experiments. (author). 138 refs., 16 figs., 1 tab.

  16. MDSplus integration at TCABR tokamak: Current status

    International Nuclear Information System (INIS)

    Sá, W.P. de; Ronchi, G.

    2016-01-01

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  17. MDSplus integration at TCABR tokamak: Current status

    Energy Technology Data Exchange (ETDEWEB)

    Sá, W.P. de, E-mail: pires@if.usp.br; Ronchi, G., E-mail: gronchi@if.usp.br

    2016-11-15

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  18. Lower hybrid current drive in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  19. Lower hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi

    1999-03-01

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

  20. Bootstrap currents in stellarators and tokamaks

    International Nuclear Information System (INIS)

    Okamoto, Masao; Nakajima, Noriyoshi.

    1990-09-01

    The remarkable feature of the bootstrap current in stellarators is it's strong dependence on the magnetic field configuration. Neoclassical bootstrap currents in a large helical device of torsatron/heliotron type (L = 2, M = 10, R = 4 m, B = 4 T) is evaluated in the banana (1/ν) and the plateau regime. Various vacuum magnetic field configurations are studied with a view to minimizing the bootstrap current. It is found that in the banana regime, shifting of the magnetic axis and shaping of magnetic surfaces have a remarkable influence on the bootstrap current; a small outward shift of the magnetic axis and vertically elongated magnetic surfaces are favourable for a reduction of the bootstrap current. It is noted, however, that the ripple diffusion in the 1/ν regime has opposite tendency to the bootstrap current; it increases with the outward shift and increases as the plasma cross section is vertically elongated. The comparison will be made between bootstrap currents in stellarators and tokamaks. (author)

  1. Physics issues of high bootstrap current tokamaks

    International Nuclear Information System (INIS)

    Ozeki, T.; Azumi, M.; Ishii, Y.

    1997-01-01

    Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of q min is close to the region of high ∇n e . Finally, a method for stable start-up for a plasma with a hollow current profile is describe, and stable sustainment of a steady-state plasma with high bootstrap current is discussed. (Author)

  2. Current drive by spheromak injection into a tokamak

    International Nuclear Information System (INIS)

    Brown, M.R.; Bellan, P.M.

    1990-01-01

    The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

  3. Combined RF current drive and bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Schultz, S. D.; Bers, A.; Ram, A. K.

    1999-01-01

    By calculating radio frequency current drive (RFCD) and the bootstrap current in a consistent kinetic manner, we find synergistic effects in the total noninductive current density in tokamaks [1]. We include quasilinear diffusion in the Drift Kinetic Equation (DKE) in order to generalize neoclassical theory to highly non-Maxwellian electron distributions due to RFCD. The parallel plasma current is evaluated numerically with the help of the FASTEP Fokker-Planck code [2]. Current drive efficiency is found to be significantly affected by neoclassical effects, even in cases where only circulating electrons interact with the waves. Predictions of the current drive efficiency are made for lower hybrid and electron cyclotron wave current drive scenarios in the presence of bootstrap current

  4. Joint Czechoslovak-Soviet workshop on current drive in tokamaks

    International Nuclear Information System (INIS)

    1985-10-01

    At the Joint Czechoslovak-Soviet Workshop on Current Drive in Tokamaks, five papers dealing with issues of general interest were presented. In a theoretical paper by Klima and Pavlo a one-dimensional model of the lower-hybrid current drive is described and the results of its analysis are used in a numerical simulation using T-7 tokamak parameters. In the second theoretical paper by Vojtsekhovich, Parail and Pereverzev the influence of the LH wave spectrum on the efficiency of the current drive is studied. Two papers deal with a new microwave system designed for experiments on LHCD in the T-7 tokamak. In particular, the power spectra of new four-waveguide grills are computed. In the last paper the non-inductive start-up of the discharge in the T-7 tokamak by means of electron cyclotron waves is investigated. (J.U.)

  5. ELECTRON CYCLOTRON CURRENT DRIVE EFFICIENCY IN GENERAL TOKAMAK GEOMETRY

    International Nuclear Information System (INIS)

    LIN-LUI, Y.R; CHAN, V.S; PRATER, R.

    2003-01-01

    Green's-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasi-linear rf diffusion operator describes wave-particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the non-inductive current drive of electron cyclotron waves

  6. Current drive by Alfven waves in elongated cross section tokamak

    International Nuclear Information System (INIS)

    Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A.; Assis, A.S. de

    1997-01-01

    Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves 9GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory

  7. Current drive by Alfven waves in elongated cross section tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Fisica; Assis, A.S. de [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves (GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory

  8. Electron cyclotron current drive efficiency in an axisymmetric tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez-Tapia, C.; Beltran-Plata, M. [Instituto Nacional de Investigaciones Nucleares, Dept. de Fisica, Mexico D.F. (Mexico)

    2004-07-01

    The neoclassical transport theory is applied to calculate electron cyclotron current drive (ECCD) efficiency in an axisymmetric tokamak in the low-collisionality regime. The tokamak ordering is used to obtain a system of equations that describe the dynamics of the plasma where the nonlinear ponderomotive (PM) force due to high-power radio-frequency (RF) waves is included. The PM force is produced around an electron cyclotron resonant surface at a specific poloidal location. The ECCD efficiency is analyzed in the cases of first and second harmonics (for different impinging angles of the RF waves) and it is validated using experimental parameter values from TCV and T-10 tokamaks. The results are in agreement with those obtained by means of Green's function techniques. (authors)

  9. Computer simulation of transport driven current in tokamaks

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Fast wave current drive in reactor scale tokamaks

    International Nuclear Information System (INIS)

    Moreau, D.

    1992-01-01

    The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

  11. Effects of magnetic shear on current penetration in a tokamak

    International Nuclear Information System (INIS)

    Zhang Pengyun; Wang Long

    2001-01-01

    The penetrations of the parallel and perpendicular components of plasma currents are interrelated to each other due to the existence of magnetic shear in a tokamak configuration. Effects of the shear on the penetration of Fourier components of toroidal plasma current are analysed in a cylindrical column model. The current penetration is obviously strengthened by the shear for a bell-bike conductivity profile and low safety factor and low aspect ratio

  12. Toroidal current asymmetry in tokamak disruptions

    Science.gov (United States)

    Strauss, H. R.

    2014-10-01

    It was discovered on JET that disruptions were accompanied by toroidal asymmetry of the toroidal plasma current I ϕ. It was found that the toroidal current asymmetry was proportional to the vertical current moment asymmetry with positive sign for an upward vertical displacement event (VDE) and negative sign for a downward VDE. It was observed that greater displacement leads to greater measured I ϕ asymmetry. Here, it is shown that this is essentially a kinematic effect produced by a VDE interacting with three dimensional MHD perturbations. The relation of toroidal current asymmetry and vertical current moment is calculated analytically and is verified by numerical simulations. It is shown analytically that the toroidal variation of the toroidal plasma current is accompanied by an equal and opposite variation of the toroidal current flowing in a thin wall surrounding the plasma. These currents are connected by 3D halo current, which is π/2 radians out of phase with the n = 1 toroidal current variations.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  14. Disruption-induced poloidal currents in the tokamak wall

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2017-01-01

    Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

  15. Analysis on Θ pumping for tokamak current drive

    International Nuclear Information System (INIS)

    Miyamoto, Kenro; Naito, Osamu

    1986-01-01

    Analytical results of Θ pumping for the tokamak current drive are presented. Diffusion of externally applied oscillating electric field into the tokamak plasma is examined when the plasma is normal. When the oscillating electric field is parallel to the stationary toroidal plasma current and the induced current density by the applied electric field becomes larger than the average density of the toroidal plasma current over the plasma cross section, the radial profile of the safety factor has the extremum near the plasma boundary region and MHD instabilities are excited. It is assumed that anomalous diffusion of the induced current localized in the plasma boundary region takes place, so that the extreme value in the radial profile of the safety factor disappears. The anomalously diffused electric field due to this relaxation process has net d. c component and its non-zero value of the time average is estimated. Then the condition of the tokamak current drive by Θ pumping is derived. Some numerical results are presented for an example of a fusion grade plasma. (author)

  16. Disruption-induced poloidal currents in the tokamak wall

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V.D., E-mail: Pustovitov_VD@nrcki.ru [National Research Centre ‘Kurchatov Institute’, Pl. Kurchatova 1, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow 115409, Russia (Russian Federation)

    2017-04-15

    Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

  17. 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

  18. 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

  19. Resonant fields created by spiral electric currents in Tokamaks

    International Nuclear Information System (INIS)

    Fernandes, A.S.; Caldas, I.L.

    1985-01-01

    The influence of the resonant magnetic perturbations, created by electric currents in spirals, on the plasma confinement in a tokamak with circular section and large aspect ratio is investigated. These perturbations create magnetic islands around the rational magnetic surface which has the helicity of the helicoidal currents. The intensities of these currents are calculated in order to the magnetic islands reach the limiter or others rational surfaces, what could provoke the plasma disrupture. The electric current intensities are estimated, in two spiral sets with different helicities, which create a predominantly stocastic region among the rational magnetic surfaces with these helicities. (L.C.) [pt

  20. Tokamak plasma current disruption infrared control system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ulrickson, M.

    1987-01-01

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

  1. Impact of electron trapping on RF current drive in tokamaks

    International Nuclear Information System (INIS)

    Giruzzi, G.; Engelmann, F.

    1987-01-01

    The impact of the presence of trapped electrons on noninductive current drive by RF waves in tokamak plasmas is investigated. The appropriate response function, allowing to express the current drive efficiency J/P by a simple analytical formula, has been derived. The approach displays the reasons for the degradation of the current drive efficiency away from the plasma axis in the case of methods relying on the diffusion of electrons in the velocity component perpendicular to the confining magnetic field. It is shown that this degradation is appreciable even for large resonant parallel velocities. (author) [pt

  2. 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.

  3. 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.

  4. 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)

  5. Hamiltonian analysis of fast wave current drive in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Becoulet, A; Fraboulet, D; Giruzzi, G; Moreau, D; Saoutic, B [Association Euratom-CEA, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Chinardet, J [CISI Ingenierie, Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)

    1993-12-01

    The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs.

  6. Hamiltonian analysis of fast wave current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Becoulet, A.; Fraboulet, D.; Giruzzi, G.; Moreau, D.; Saoutic, B.

    1993-12-01

    The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs

  7. Current drive and profile control in low aspect ratio tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.; Chiu, S.C.; Lin-Liu, Y.R.; Miller, R.L.; Turnbull, A.D.

    1995-07-01

    The key to the theoretically predicted high performance of a low aspect ratio tokamak (LAT) is its ability to operate at very large plasma current*I p . The plasma current at low aspect ratios follows the approximate formula: I p ∼ (5a 2 B t /Rqψ) [(1 + κ 2 )/2] [A/(A - 1)] where A quadruple-bond R/a which was derived from equilibrium studies. For constant qψ and B t , I p can increase by an order of magnitude over the case of tokamaks with A approx-gt 2.5. The large current results in a significantly enhanced β t (quadruple-bond β N I p /aB t ) possibly of order unity. It also compensates for the reduction in A to maintain the same confinement performance assuming the confinement time τ follows the generic form ∼ HI p P -1 / 2 R 3 / 2 κ 1 / 2 . The initiation and maintenance of such a large current is therefore a key issue for LATs

  8. Flux surface shape and current profile optimization in tokamaks

    International Nuclear Information System (INIS)

    Dobrott, D.R.; Miller, R.L.

    1977-01-01

    Axisymmetric tokamak equilibria of noncircular cross section are analyzed numerically to study the effects of flux surface shape and current profile on ideal and resistive interchange stability. Various current profiles are examined for circles, ellipses, dees, and doublets. A numerical code separately analyzes stability in the neighborhood of the magnetic axis and in the remainder of the plasma using the criteria of Mercier and Glasser, Greene, and Johnson. Results are interpreted in terms of flux surface averaged quantities such as magnetic well, shear, and the spatial variation in the magnetic field energy density over the cross section. The maximum stable β is found to vary significantly with shape and current profile. For current profiles varying linearly with poloidal flux, the highest β's found were for doublets. Finally, an algorithm is presented which optimizes the current profile for circles and dees by making the plasma everywhere marginally stable

  9. Eddy current calculations for the tore supra tokamak

    International Nuclear Information System (INIS)

    Blum, J.; Dupas, L.; Leloup, C.; Thooris, B.

    1983-01-01

    This paper deals with the calculation of the eddy currents in the structures of a Tokamak, which can be assimilated to thin conductors, so that the three-dimensional problem can be reduced mathematically to a two-dimensional one, the variables being two orthogonal coordinates of the considered surface. A variational formulation of the problem in terms of the electric vector potential is then given and a finite element method has been used, which enables to treat the complicated geometry of the toroidal field magnet, the mechanical structures and the vacuum vessels of Tore Supra

  10. Neoclassical current effects in neutral-beam-heated tokamak discharges

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1981-01-01

    There is a long-standing prediction from neoclassical theory that strong contributions to the toroidal current should be driven by friction between trapped and passing particles when βsub(pol) exceeds root (R/a) in a tokamak. A number of neutral-beam heating experiments can now produce such parameters, and it is of interest to calculate the behaviour which should occur in this regime to determine the feasibility of using such a 'bootstrap' current as a steady-state tokamak current source. It is found that the neoclassical current should be large enough to reverse the external loop voltage for typical experimental parameters (ISX-B, in particular) in cases where the total current is fixed and to produce a detectable excess of total current above the pre-programmed (demand) value in cases where the loop voltage is regulated. Other manifestations of such a current should be either: a sharp rise in the central q-value (producing a cessation of internal m=1 and m=2 MHD activity), with an enhancement by two orders of magnitude of ion thermal conductivity (due to the formation of a hollow current density profile and a consequent drop in local values of the poloidal magnetic field in the central plasma region), or an enhanced tendency for disruption (arising from magnetic reconnection in hollow-profile equilibria). Since these gross manifestations are absent in a wide range of experiments on the Impurity Study Experiment (ISX-B), as reported earlier, the conclusion is that the neoclassical current, if present, can have a value no larger than 25% of its theoretically calculated value. Since the neoclassical particle (Ware) pinch is strongly related to the neoclassical current in the theory (Onsager reciprocity), the existence of the particle pinch is thus called into question. (author)

  11. Predictions of of fast wave heating, current drive, and current drive antenna arrays for advanced tokamaks

    International Nuclear Information System (INIS)

    Batchelor, D.B.; Baity, F.W.; Carter, M.D.

    1995-01-01

    The objective of the advanced tokamak program is to optimize plasma performance leading to a compact tokamak reactor through active, steady state control of the current profile using non-inductive current drive and profile control. To achieve this objective requires compatibility and flexibility in the use of available heating and current drive systems - ion cyclotron radio frequency (ICRF), neutral beams, and lower hybrid. For any advanced tokamak, the following are important challenges to effective use of fast waves in various role of direct electron heating, minority ion heating, and current drive: (1) to employ the heating and current drive systems to give self-consistent pressure and current profiles leading to the desired advanced tokamak operating modes; (2) to minimize absorption of the fast waves by parasitic resonances, which limit current drive; (3) to optimize and control the spectrum of fast waves launched by the antenna array for the required mix of simultaneous heating and current drive. The paper addresses these issues using theoretical and computational tools developed at a number of institutions by benchmarking the computations against available experimental data and applying them to the specific case of TPX. (author). 6 refs, 3 figs

  12. Predictions of fast wave heating, current drive, and current drive antenna arrays for advanced tokamaks

    International Nuclear Information System (INIS)

    Batchelor, D.B.; Baity, F.W.; Carter, M.D.

    1994-01-01

    The objective of the advanced tokamak program is to optimize plasma performance leading to a compact tokamak reactor through active, steady state control of the current profile using non-inductive current drive and profile control. To achieve these objectives requires compatibility and flexibility in the use of available heating and current drive systems--ion cyclotron radio frequency (ICRF), neutral beams, and lower hybrid. For any advanced tokamak, the following are important challenges to effective use of fast waves in various roles of direct electron heating, minority ion heating, and current drive: (1) to employ the heating and current drive systems to give self-consistent pressure and current profiles leading to the desired advanced tokamak operating modes; (2) to minimize absorption of the fast waves by parasitic resonances, which limit current drive; (3) to optimize and control the spectrum of fast waves launched by the antenna array for the required mix of simultaneous heating and current drive. The authors have addressed these issues using theoretical and computational tools developed at a number of institutions by benchmarking the computations against available experimental data and applying them to the specific case of TPX

  13. Analytic description of tokamak equilibrium sustained by high fraction bootstrap current

    International Nuclear Information System (INIS)

    Shi Bingren

    2002-01-01

    Recently, to save the current drive power and to obtain more favorable confinement merit for tokamak reactor, large faction bootstrap current sustained equilibrium has attracted great interests both theoretically and experimentally. An powerful expanding technique and the tokamak ordering are used to expand the Grad-Shafranov equation to obtain a series of ordinary differential equations which allow for different sets of input parameters. The fully bootstrap current sustained tokamak equilibria are then solved analytically

  14. Neoclassical effects on RF current drive in tokamaks

    International Nuclear Information System (INIS)

    Yoshioka, K.; Antonsen, T.M. Jr.

    1986-01-01

    Neoclassical effects on RF current drive which arise because of the inhomogeneity of the magnetic field in tokamak devices are analysed. A bounce averaged 2-D Fokker-Planck equation is derived from the drift kinetic equation and is solved numerically. The model features current drive due to a strong RF wave field. The efficiency of current drive by electron cyclotron waves is significantly reduced when the waves are absorbed at the low magnetic field side of a given flux surface, whereas the efficiency remains at the same level as in the homogeneous ideal plasma when the waves are absorbed at the high field side. The efficiency of current drive by fast waves (compressional Alfven waves) with low phase velocity (vsub(parallel)/vsub(th)<1) is significantly degraded by neoclassical effects, no matter where the wave is absorbed, and the applicability of this wave seems, therefore, to be doubtful. (author)

  15. Spreading of wave-driven currents in a tokamak

    International Nuclear Information System (INIS)

    Ignat, D.W.; Kaita, R.; Jardin, S.C.; Okabayashi, M.

    1996-01-01

    Lower hybrid current drive (LHCD) in the tokamak Princeton Beta Experiment-Modification (PBX-M) is computed with a dynamic model in order to understand an actual discharge aimed at raising the central q above unity. Such configurations offer advantages for steady-state operation and plasma stability. For the particular parameters of this PBX-M experiment, the calculation found singular profiles of plasma current density J and safety factor q developing soon after LHCD begins. Smoothing the lower hybrid-driven current and power using a diffusion-Eke equation and a velocity-independent diffusivity for fast-electron current brought the model into reasonable agreement with the measurements if D fast ∼ 1.0 m 2 /s. Such a value for D fast is in the range suggested by other work

  16. Vessel eddy current characteristics in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jana, Subrata; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Dhongde, Jasraj; Masand, Harish

    2016-11-15

    Highlights: • Eddy current distribution in the SST-1 vacuum vessel. • Circuit model analysis of eddy current. • A comparison of the field lines with and without the plasma column in identical conditions. • The influence of eddy current in magnetic NULL dynamics. - Abstract: Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285 m and minor axis of 0.81 m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

  17. 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)

  18. 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

  19. Electron-cyclotron current drive in the tokamak physics experiment

    International Nuclear Information System (INIS)

    Smith, G.R.; Kritz, A.H.; Radin, S.H.

    1992-01-01

    Ray-tracking calculations provide estimates of the electron-cyclotron heating (ECH) power required to suppress tearing modes near the q=2 surface in the Tokamak Physics Experiment. Effects of finite beam width and divergence are included, as are the effects of scattering of the ECH power by drift-wave turbulence. A frequency of about 120 GHz allows current drive on the small-R (high-B) portion of q=2, while 80 GHz drives current on the large-R (low-B) portion. The higher frequency has the advantages of less sensitivity to wave and plasma parameters and of no trapped-electron degradation of current-drive efficiency. Less than 1 MW suffices to suppress tearing modes even with high turbulence levels

  20. 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)

  1. Enhanced lower hybrid current drive experiments on HT-7 tokamak

    International Nuclear Information System (INIS)

    Shen Weici; Kuang Guangli; Liu Yuexiu; Ding Bojiang; Shi Yaojiang

    2003-01-01

    Effective Lower Hybrid Current Driving (LHCD) and improved confinement experiments in higher plasma parameters (I p >200 kA, n e >2 x 10 13 cm -3 , T e ≥1 keV) have been curried out in optimized LH wave spectrum and plasma parameters in HT-7 superconducting tokamak. The dependence of current driving efficiency on LH power spectrum, plasma density (anti n e ) and toroidal magnetic field B T has been obtained under optimal conditions. A good CD efficiency was obtained at higher plasma current and higher electron density. The improvement of the energy confinement time is accompanied with the increase in line averaged electron density, and in ion and electron temperatures. The highest current driving efficiency reached η CD =I p (anti n e )R/P RF ≅1.05 x 10 19 Am -2 /W. Wave-plasma coupling was sustained in a good state and the reflective coefficient was less than 5%. The experiments have also demonstrated the ability of LH wave in the start-up and ramp-up of the plasma current. The measurement of the temporal distribution of plasma parameter shows that lower hybrid leads to a broader profile in plasma parameter. The LH power deposition profile and the plasma current density profile were modeled with a 2D Fokker-Planck code corresponding to the evolution process of the hard x-ray detector array

  2. Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

    Science.gov (United States)

    Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

    2017-10-01

    The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

  3. Control of bootstrap current in the pedestal region of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Shaing, K. C. [Institute for Space and Plasma Sciences, National Cheng Kung University, Tainan City 70101, Taiwan (China); Department of Engineering Physics, University of Wisconsin, Madison, Wisconsin 53796 (United States); Lai, A. L. [Institute for Space and Plasma Sciences, National Cheng Kung University, Tainan City 70101, Taiwan (China)

    2013-12-15

    The high confinement mode (H-mode) plasmas in the pedestal region of tokamaks are characterized by steep gradient of the radial electric field, and sonic poloidal U{sub p,m} flow that consists of poloidal components of the E×B flow and the plasma flow velocity that is parallel to the magnetic field B. Here, E is the electric field. The bootstrap current that is important for the equilibrium, and stability of the pedestal of H-mode plasmas is shown to have an expression different from that in the conventional theory. In the limit where ‖U{sub p,m}‖≫ 1, the bootstrap current is driven by the electron temperature gradient and inductive electric field fundamentally different from that in the conventional theory. The bootstrap current in the pedestal region can be controlled through manipulating U{sub p,m} and the gradient of the radial electric. This, in turn, can control plasma stability such as edge-localized modes. Quantitative evaluations of various coefficients are shown to illustrate that the bootstrap current remains finite when ‖U{sub p,m}‖ approaches infinite and to provide indications how to control the bootstrap current. Approximate analytic expressions for viscous coefficients that join results in the banana and plateau-Pfirsch-Schluter regimes are presented to facilitate bootstrap and neoclassical transport simulations in the pedestal region.

  4. Intense relativistic electron beam injector system for tokamak current drive

    International Nuclear Information System (INIS)

    Bailey, V.L.; Creedon, J.M.; Ecker, B.M.; Helava, H.I.

    1983-01-01

    We report experimental and theoretical studies of an intense relativistic electron beam (REB) injection system designed for tokamak current drive experiments. The injection system uses a standard high-voltage pulsed REB generator and a magnetically insulated transmission line (MITL) to drive an REB-accelerating diode in plasma. A series of preliminary experiments has been carried out to test the system by injecting REBs into a test chamber with preformed plasma and applied magnetic field. REBs were accelerated from two types of diodes: a conventional vacuum diode with foil anode, and a plasma diode, i.e., an REB cathode immersed in the plasma. REB current was in the range of 50 to 100 kA and REB particle energy ranged from 0.1 to 1.0 MeV. MITL power density exceeded 10 GW/cm 2 . Performance of the injection system and REB transport properties is documented for plasma densities from 5 x 10 12 to 2 x 10 14 cm -3 . Injection system data are compared with numerical calculations of the performance of the coupled system consisting of the generator, MITL, and diode

  5. Experimental modeling of eddy currents and deflections for tokamak limiters

    International Nuclear Information System (INIS)

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-01-01

    In this study, experiments were performed to investigate deflection, current, and material stress in cantilever beams with the Fusion ELectromagnetic Induction eXperiment (FELIX) at the Argonne National Laboratory. Since structures near the plasma are typically cantilevered, the beams provide a good model for the limiter blades of a tokamak fusion reactor. The test pieces were copper, aluminum, phosphor bronze, and brass cantilever beams, clamped rigidly at one end with a nonconducting support frame inside the FELIX test volume. The primary data recorded as functions of time were the beam deflection measured with a noncontact electro-optical device, the total eddy current measured with a Rogowski coil and linking through a central hole in the beam, and the material stress extracted from strain gauges. Measurements of stress and deflection were taken at selected positions along the beam. The extent of the coupling effect depends on several factors. These include the size, the electrical and mechanical properties of the beam, segmenting of the beam, the decay rate of the dipole field, and the strength of the solenoid field

  6. Variations of current profiles in tokamaks. Formation mechanism and confinement property of current-hole configuration

    International Nuclear Information System (INIS)

    Takizuka, Tomonori

    2003-01-01

    The formation mechanism of the current hole in tokamak plasmas is reviewed. Experimental results of JT-60U are shown. Increase of the off-central noninductive current is a key factor for the current-hole formation. The internal Transport Barrier (ITB), which generates large bootstrap current, plays an important role. The central current density in the hole stays nearly 0. The idea of a new equilibrium for a tokamak plasma with a current hole is introduced. This equilibrium configuration called Axisymmetric Tri-Magnetic-Islands (ATMI) equilibrium', has three islands along the R direction (a central-negative-current island and side-positive-current islands). The equilibrium is stable with the elongation coils when the current in the ATMI region is limited to a small amount. The confinement properties of a current-hole configuration with box-type ITB is described. A scaling of the core poloidal beta inside the ITB, β p,core , is given as ε f β p,core approx. = 1, which suggests the equilibrium limit (ε f : inverse aspect ratio at the ITB foot). Though the core stored energy is little dependent on the heating power, the estimated heat diffusivity in the ITB region moderately correlates with a neoclassical diffusivity. (author)

  7. First results on fast wave current drive in advanced tokamak discharges in DIII-D

    International Nuclear Information System (INIS)

    Prater, R.; Cary, W.P.; Baity, F.W.

    1995-07-01

    Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m 2

  8. Electron Bernstein wave current drive in the start-up phase of a tokamak discharge

    International Nuclear Information System (INIS)

    Montes, A.; Ludwig, G.O.

    1986-04-01

    Current drive by electron Bernstein waves in the start-up phase of tokamak discharges is studied. A general analytical expression is derived for the figure of merit J/Pd associated with these waves. This is coupled with a ray tracing code, allowing the calculation of the total current generated per unit of incident power in realistic tokamak conditions. The resuts show that the electron Bernstein waves can drive substantial currents even at very low electron temperatures. (Author) [pt

  9. Structured Cable for High-Current Coils of Tokamaks

    Science.gov (United States)

    Benson, Christopher; McIntyre, Peter; Sattarov, Akhdiyor; Mann, Thomas

    2011-10-01

    The 45 kA superconducting cable for the ITER central solenoid coil has yielded questionable results in two recent tests. In both cases the cable Tc increased after cycling only a fraction of the design life, indicating degradation due to fatigue and fracture among the superconducting strands. The Accelerator Research Lab at Texas A&M University is developing a design for a Nb3Sn structured cable suitable for such tokamak coils. The superconductor is configured in 6 sub-cables, and each subcable is supported within a channel of a central support structure within a high-strength armor sheath. The structured cable addresses two issues that are thought to compromise opposition at high current. The strands are supported without cross-overs (which produce stress concentration); and armor sheath and core structure bypass stress through the coil and among subcables so that the stress within each subcable is only what is produced directly upon it. Details of the design and plans for development will be presented.

  10. On tokamak equilibria with a zero current or negative current central region

    International Nuclear Information System (INIS)

    Chu, M.S.; Parks, P.B.

    2002-01-01

    Several tokamak experiments have reported the development of a central region with vanishing currents (the current hole). The straightforward application of results from the work of Greene, Johnson and Weimer [Phys. Fluids 14, 671 (1971)] on a tokamak equilibrium to these plasmas leads to the apparent singularities in several physical quantities including the Shafranov shift and casts doubts on the existence of this type of equilibria. In this paper, the above quoted equilibrium theory is re-examined and extended to include equilibria with a current hole. It is shown that singularities can be circumvented and that equilibria with a central current hole do satisfy the magnetohydrodynamic equilibrium condition with regular behavior for all the physical quantities and do not lead to infinitely large Shafranov shifts. Isolated equilibria with negative current in the central region could exist. But equilibria with negative currents in general do not have neighboring equilibria and thus cannot have experimental realization, i.e., no negative currents can be driven in the central region

  11. Current profile evolution during fast wave current drive on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Forest, C.B.; Baity, F.W.

    1995-06-01

    The effect of co and counter fast wave current drive (FWCD) on the plasma current profile has been measured for neutral beam heated plasmas with reversed magnetic shear on the DIII-D tokamak. Although the response of the loop voltage profile was consistent with the application of co and counter FWCD, little difference was observed between the current profiles for the opposite directions of FWCD. The evolution of the current profile was successfully modeled using the ONETWO transport code. The simulation showed that the small difference between the current profiles for co and counter FWCD was mainly due to an offsetting change in the o at sign c current proffie. In addition, the time scale for the loop voltage to reach equilibrium (i.e., flatten) was found to be much longer than the FWCD pulse, which limited the ability of the current profile to fully respond to co or counter FWCD

  12. Numerical Simulation of Neoclassical Currents, Parallel Viscosity, and Radial Current Balance in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Kiviniemi, T.

    2001-01-01

    One of the principal problems en route to a fusion reactor is that of insufficient plasma confinement, which has lead to both theoretical and experimental research into transport processes in the parameter range relevant for fusion energy production. The neoclassical theory of tokamak transport is well-established unlike the theory of turbulence driven anomalous transport in which extensive progress has been made during last few years. So far, anomalous transport has been dominant in experiments, but transport may be reduced to the neoclassical level in advanced tokamak scenarios. This thesis reports a numerical study of neoclassical fluxes, parallel viscosity, and neoclassical radial current balance in tokamaks. Neoclassical parallel viscosity and particle fluxes are simulated over a wide range of collisionalities, using the fully kinetic five-dimensional neoclassical orbit-following Monte Carlo code ASCOT. The qualitative behavior of parallel viscosity derived in earlier analytic models is shown to be incorrect for high poloidal Mach numbers. This is because the poloidal dependence of density was neglected. However, in high Mach number regime, it is the convection and compression terms, rather than the parallel viscosity term, that are shown to dominate the momentum balance. For fluxes, a reasonable agreement between numerical and analytical results is found in the collisional parameter regime. Neoclassical particle fluxes are additionally studied in the banana regime using the three-dimensional Fokker-Planck code DEPORA, which solves the drift-kinetic equation with finite differencing. Limitations of the small inverse aspect ratio approximation adopted in the analytic theory are addressed. Assuming that the anomalous transport is ambipolar, the radial electric field and its shear at the tokamak plasma edge can be solved from the neoclassical radial current balance. This is performed both for JET and ASDEX Upgrade tokamaks using the ASCOT code. It is shown that

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

    International Nuclear Information System (INIS)

    Peters, M.

    1996-01-01

    In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity χ to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Peters, M

    1996-01-16

    In the first experiment the plasma current in the RTP tokamak is varied. Here the underlying idea was to check whether at a low plasma current, transport in the tokamak resembles transport in stellarators more than at higher currents. Secondly, experiments have been done to study the relation of the diffusivity {chi} to the temperature and its gradient in both W7-AS and RTP. In this case the underlying idea was to find the explanation for the phenomenon observed in both tokamaks and stellarators that the quality of the confinement degrades when more heating is applied. A possible explanation is that the diffusivity increases with the temperature or its gradient. Whereas in standard tokamak and stellarator experiments the temperature and its gradient are strongly correlated, a special capability of the plasma heating system of W7-AS and RTP can force them to decouple. (orig.).

  15. 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 ...

  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. Characteristics of current quenches during disruptions in the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zhang, Y; Chen, Z Y; Fang, D; Jin, W; Huang, Y H; Wang, Z J; Yang, Z J; Chen, Z P; Ding, Y H; Zhang, M; Zhuang, G

    2012-01-01

    Characteristics of tokamak current quenches are an important issue for the determination of electro-magnetic forces that act on the in-vessel components and vacuum vessel during major disruptions. The characteristics of current quenches in spontaneous disruptions in the J-TEXT tokamak have been investigated. It is shown that the waveforms for the fastest current quenches are more accurately fitted by linear current decays than exponential, although neither is a good fit in many slower cases. The minimum current quench time is about 2.4 ms for the J-TEXT tokamak. The maximum instantaneous current quench rate is more than seven times the average current quench rate in J-TEXT. (paper)

  18. Numerical simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

    International Nuclear Information System (INIS)

    Jardin, S.C.; Schmidt, J.A.

    1998-01-01

    The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrape-off layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer advantages over the more conventional method of controlling this instability when applied in a reactor environment. (author)

  19. Internal m=1, n=1 helical mode in a tokamak with nonmonotonic current profile

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Mikhajlovskij, A.B.

    1988-01-01

    Internal helical mode in a tokamak with two resonance surfaces, on which storing coefficient reduces to unity is studied theoretically. A general criterion for the investigated perturbations stability is obtained. Dispersion equation, describing both ideal and resistive helical modes, is derived. Analytic calculations for the case of perturbations localized near the tokamak axis are made. It is shown that in the framework of standard ideal hydrodynamics such perturbations are unstable at characteristic nonmonotonous profiles of the current

  20. Comparison between 3D eddy current patterns in tokamak in-vessel components generated by disruptions

    International Nuclear Information System (INIS)

    Sakellaris, J.; Crutzen, Y.

    1996-01-01

    During plasma disruption events in Tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current into eddy currents in the passive structures. In the ITER program two design concepts have been proposed. One approach (ITER CDA design) is based on copper stabilization loops (i.e., twin loops) attached to box-shaped blanket segments, electrically and mechanically separated along the toroidal direction. For another design concept (ITER EDA design) based on lower plasma elongation there is no need for specific stabilization loops. The passive stabilization is obtained by toroidally continuous components (i.e., the plasma facing wall of the blanket segments allows a continuity along the toroidal direction). Consequently, toroidal currents flow, when electromagnetic transients occur. Electromagnetic loads appear in the blanket structures in case of plasma disruptions and/or vertical displacement events either for the ITER CDA design concept or for the ITER EDA design concept. In this paper the influence of the in-vessel design configuration concepts--insulated segments or electrically continuous structures--in terms of magnetic shielding and electric insulation on the magnitude and the flow pattern of the eddy currents is investigated. This investigation will allow a performance evaluation of the two proposed design concepts

  1. Effect of Equilibrium Current Profiles on External Kink Modes in Tokamaks

    International Nuclear Information System (INIS)

    Liu Chao; Liu Yue; Ma Zhaoshuai

    2014-01-01

    Based on a linearized MHD model, the effect of equilibrium current profiles on external kink modes in tokamaks is studied by MARS code. Three types of equilibrium current profiles are adopted in this work. Firstly, a set of parabolic equilibrium current profiles are chosen. In these profiles the maximum current values in the center of the plasma are fixed, and the currents have different gradient and jump at the plasma boundary. The effects of the current gradient and jump on the growth rate of external kink mode are investigated. It is found that the current jump which causes the q profiles to change plays an important role in the external kink modes in tokamaks. Secondly, a set of step equilibrium current profiles with different jump positions are chosen. The effect of jump position on external kink modes is discussed. Thirdly, a set of parabolic equilibrium current profiles with current bumps are chosen for the case of off-axis heating. The effects of height, width and position of the current bumps on external kink modes are analyzed. The flat equilibrium current profiles are disadvantageous for the MHD stabilities of tokamaks, because of the large current jump at the plasma edge. The peaked equilibrium current profiles and a large and localized current bump near the plasma edge benefit the MHD stabilities of tokamaks

  2. Mitigation of current quench by runaway electrons in LHCD discharges in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Lu, H.W.; Hu, L.Q.; Lin, S.Y.; Zhong, G.Q.

    2009-01-01

    Production of runaway electrons during a major disruption has been observed in HT-7 Tokamak. The runaway current plateaus, which can carry part of the pre-disruptive current, are observed in lower-hybrid current drive (LHCD) limiter discharges. It is found that the runaway current can mitigate the disruptions effectively. Detailed observations are presented on the runaway electrons generated following disruptions in the HT-7 tokamak with carbon limited discharges. The results indicate that the magnetic oscillations play an important role in the activity of runaway electrons in disruption. (author)

  3. 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

  4. On a mechanism of switching off low-hybrid run away currents in tokamak devices

    International Nuclear Information System (INIS)

    Budnikov, V.N.; Esipov, L.A.; Irzak, M.A.

    1990-01-01

    The problem of the generation of low-hybrid run-away currents (LR) in tokamak devices is described. The mechanism of switching off LRCs is considered. Qualitative representation of the density limit, the transitions of which stops the generation of currents, is given

  5. Numerical and experimental analysis of eddy currents induced in tokamak machines

    International Nuclear Information System (INIS)

    Takahashi, T.; Takahashi, G.; Kazawa, Y.; Suzuki, Y.

    1977-01-01

    This paper deals with eddy current phenomena in Tokamak machines. A numerical method is presented which will permit eddy currents to be calculated. Examples of numerical results and a discussion of the JT-60 are shown. Calculations are checked by measurements in basic models

  6. 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.)

  7. Fast wave current drive experiment on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R.; Porkolab, M.; Baity, F.W.; Goulding, R.H.; Hoffman, J.D.; James, R.A.; Kawashima, H.

    1992-06-01

    One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition ω - κ parallele υ parallele = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D

  8. 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.

  9. Implications of rf current drive theory for next step steady-state tokamak design

    International Nuclear Information System (INIS)

    Schultz, J.H.

    1985-06-01

    Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

  10. 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.

  11. Modeling of Eddy current distribution and equilibrium reconstruction in the SST-1 Tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Sharma, Deepti; Radhakrishnana, Srinivasan; Daniel, Raju; Shankara Joisa, Y.; Atrey, Parveen Kumar; Pathak, Surya Kumar; Singh, Amit Kumar

    2015-01-01

    Toroidal continuity of the vacuum vessel and the cryostat leads to the generation of large eddy currents in these passive structures during the Ohmic phase of the steady state superconducting tokamak SST-1. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. During the ramping down of the Ohmic transformer current (OT), the resultant eddy currents flowing in the passive conductors play a crucial role in governing the plasma equilibrium. Amount of this eddy current and its distribution has to be accurately determined such that this can be fed to the equilibrium reconstruction code as an input. For the accurate inclusion of the effect of eddy currents in the reconstruction, the toroidally continuous conducting structures like the vacuum vessel and the cryostat with large poloidal cross-section and any other poloidal field (PF) coil sitting idle on the machine are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green's function and the induced currents are evaluated for the OT waveform of each plasma discharge. Consistency of this filament model is cross-checked with the 11 in-vessel and 12 out-vessel toroidal flux loop signals in SST-1. Resistances of the filaments are adjusted to reproduce the experimental measurements of these flux loops in pure OT shots and shots with OT and vertical field (BV). Such shots are taken routinely in SST-1 without the fill gas to cross-check the consistency of the filament model. A Grad-Shafranov (GS) equation solver, named as IPREQ, has been developed in IPR to reconstruct the plasma equilibrium through searching for the best-fit current density profile. Ohmic transformer current (OT), vertical field coil current (BV), currents in the passive filaments along with the plasma pressure (p) and current (I p ) profiles are used as inputs to the IPREQ

  12. Prospects for steady-state tokamak reactor operation through feedback control of the current density profile

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, D

    1994-12-31

    A brief overview of the most relevant experiments on current profile modifications, strong improvements with respect to the usual L-mode scaling laws and Troyon beta limit is presented, as relevant issues for most tokamaks. Practical means and scenarios for producing and maintaining the optimum current profiles in the various phases of the thermonuclear discharge (profile formation, current ramp-up, burn phase) are proposed. (author). 34 refs., 3 figs.

  13. Studies of non-inductive current drive in the CDX-U tokamak

    International Nuclear Information System (INIS)

    Hwang, Y.S.

    1993-01-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges, dc-helicity injection and internally-generated pressure-driven currents, have been developed on the CDX-U tokamak. To study the equilibrium and transport of these plasmas, a full set of magnetic diagnostics was installed. By applying a finite element method and a least squares error fitting technique, internal plasma current distributions are reconstructed from the measurements. Electron density distributions were obtained from 2 mm interferometer measurements by a similar least squares error technique utilizing magnetic flux configurations obtained by the magnetic analysis. Neoclassical pressure-driven currents in ECH plasmas are modeled with the reconstructed magnetic structure, using the electron density distribution and the electron temperature profile measured by a Langmuir probe. In the dc-helicity injection scheme, the need to increase injection current and maintain plasma equilibrium restricts possible arrangements. Several injection configurations were investigated, with the best found to be outside injection with a single divertor configuration, where the cathode is placed at the low field side of the x-point. Both pressure-driven and dc-helicity injected tokamaks show the importance of plasma equilibrium in obtaining high plasma current. Programmed vertical field operation has proven to be very important in achieving high plasma current. These non-inductive current drive techniques show great potential as efficient current drive methods for future steady-state and/or long-pulse fusion reactors

  14. First experimental results with the Current Limit Avoidance System at the JET tokamak

    Energy Technology Data Exchange (ETDEWEB)

    De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Galeani, S. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Jachmich, S. [Association EURATOM-Belgian State, Koninklijke Militaire School - Ecole Royale Militaire, B-1000 Brussels (Belgium); Joffrin, E. [IRFM-CEA, Centre de Cadarache, 13108 Saint-paul-lez-Durance (France); Lennholm, M. [EFDA Close Support Unit, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); European Commission, B-1049 Brussels (Belgium); Lomas, P.J. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Neto, A.C. [Associazione EURATOM-IST, Instituto de Plasmas e Fusao Nuclear, IST, 1049-001 Lisboa (Portugal); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Pironti, A. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Sips, A.C.C. [European Commission, B-1049 Brussels (Belgium); Varano, G.; Vitelli, R. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Zaccarian, L. [CNRS, LAAS, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Universitè de Toulouse, LAAS, F-31400 Toulouse (France)

    2013-06-15

    The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

  15. 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)

  16. Wave form of current quench during disruptions in tokamaks

    International Nuclear Information System (INIS)

    Sugihara, Masayoshi; Gribov, Yuri; Shimada, Michiya; Lukash, Victor; Kawano, Yasunori; Yoshino, Ryuji; Miki, Nobuharu; Ohmori, Junji; Khayrutdinov, Rustam

    2003-01-01

    The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential. (author)

  17. Generation of suprathermal electrons during plasma current startup by lower hybrid waves in a tokamak

    International Nuclear Information System (INIS)

    Ohkubo, K.; Toi, K.; Kawahata, K.

    1984-10-01

    Suprathermal electrons which carry a seed current are generated by non-resonant parametric decay instability during initial phase of lower hybrid current startup in the JIPP T-IIU tokamak. From the numerical analysis, it is found that parametrically excited lower hybrid waves at lower side band can bridge the spectral gap between the thermal velocity and the low velocity end in the pump power spectrum. (author)

  18. High-energy tritium beams as current drivers in tokamak reactors

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams

  19. Sawtooth control by on-axis electron cyclotron current drive on the WT-3 tokamak

    International Nuclear Information System (INIS)

    Asakawa, M.; Tanabe, K.; Nakayama, A.; Watanabe, M.; Nakamura, M.; Tanaka, H.; Maekawa, T.; Terumichi, Y.

    1999-01-01

    The experiments on control of sawtooth oscillations (STO) by electron cyclotron current drive (ECCD) have been performed on the WT-3 tokamak. Stabilization and excitation of STO are observed for counter-ECCD and co-ECCD, respectively, when the position of the power deposition is located inside the inversion radius. These results are due to the modification of the current profile near the magnetic axis. (author)

  20. Comparison between voltage by turn measured on different tokamaks operating in hybrid wave current drive regime

    International Nuclear Information System (INIS)

    Briffod, G.; Hoang, G.T.

    1987-06-01

    On a tokamak in a current drive operation with a hybrid wave, the R.F. current is estimated from the voltage drop by plasma turn generated by R.F. power application. This estimated current is not proportional to the injected power. There still exists in the plasma an electric field corresponding to the current part produced by induction. The role evaluation of this parameter on the current drive efficiency is important. In this report the relation voltage-R.F. current is studied on Petula and results on the voltage evolution by turn on different machines are compared [fr

  1. Neural network evaluation of tokamak current profiles for real time control (abstract)

    Science.gov (United States)

    Wróblewski, Dariusz

    1997-01-01

    Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental data is demonstrated.

  2. Neural network evaluation of tokamak current profiles for real time control

    Science.gov (United States)

    Wróblewski, Dariusz

    1997-02-01

    Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental datais demonstrated.

  3. Non-inductive current drive via helicity injection by Alfven waves in low aspects ratio Tokamak

    International Nuclear Information System (INIS)

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

    1996-01-01

    A theoretical investigation of radio frequency (RF) current drive via helicity injection in low aspect ratio tokamaks was carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell was considered. Toroidal features of low aspect ratio tokamaks were simulated by incorporation of the following effects: (i) arbitrarily small aspect ratio, R o /a ≡ 1/ε (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. The study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {ω Alf (r)} min ≤ω≥{ω Alf (r)} max , where ω Alf (r)≡ω[n(r),B o (o)] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-p, ideal magneto-hydrodynamics, the wave equation with correct boundary (matching) conditions was solved, the RF field components were found and subsequently, current drive , power deposition and efficiency were computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low aspect ratio tokamaks, in the SAW mode. A special algorithm was developed which enables the selection of the antenna parameters providing optimal current drive efficiency. (authors)

  4. Neural network evaluation of tokamak current profiles for real time control

    International Nuclear Information System (INIS)

    Wroblewski, D.

    1997-01-01

    Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q 0 , minimum value of q, q min , and the location of q min . Very good performance of the trained neural network both for simulated test data and for experimental datais demonstrated. copyright 1997 American Institute of Physics

  5. Neural network evaluation of tokamak current profiles for real time control (abstract)

    International Nuclear Information System (INIS)

    Wroblewski, D.

    1997-01-01

    Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q 0 , minimum value of q, q min , and the location of q min . Very good performance of the trained neural network both for simulated test data and for experimental data is demonstrated. copyright 1997 American Institute of Physics

  6. 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.

  7. Mass transport and the bootstrap current from Ohm's law in steady-state tokamaks

    International Nuclear Information System (INIS)

    Kim, J.-S.; Greene, J.M.

    1989-01-01

    The consequences of mass conservation and Ohm's law are examined for steady state Tokamaks. In a Tokamak, magnetofluid-dynamic waves rapidly equilibrate pressure and toroidal field along magnetic surfaces. As a result, the detailed current distribution is determined by the flux surface averaged poloidal and toroidal currents. The electrons that carry the plasma current are impeded in their motion by interactions with ions, which is resistivity and its generalizations, and by interactions with electrons, which is viscosity and its generalizations. The important viscous terms arise from the interaction between trapped and untrapped electrons, and so viscosity acts by impeding poloidal current. properly chosen, the results of neoclassical theory are The neoclassical viscous coefficient is here regarded as less likely than Spitzer conductivity to be experimentally relevant in a turbulent Tokamak. Thus, the toroidal Ohm's law is regarded as being more reliable than the poloidal Ohm's law. A combination of toroidal and poloidal Ohm's law, namely the component parallel to the magnetic field, eliminates the influence of plasma fueling, and directly relates the bootstrap current and the pressure gradient. The latter is the usual relation, but, since i

  8. Simulation of enhanced tokamak performance on DIII-D using fast wave current drive

    International Nuclear Information System (INIS)

    Grassie, J.S. de; Lin-Liu, Y.R.; Petty, C.C.; Pinsker, R.I.; Chan, V.S.; Prater, R.; John, H. St.; Baity, F.W.; Goulding, R.H.; Hoffman, D.H.

    1993-01-01

    The fast magnetosonic wave is now recognized to be a leading candidate for noninductive current drive for the tokamak reactor due to the ability of the wave to penetrate to the hot dense core region. Fast wave current drive (FWCD) experiments on DIII-D have realized up to 120 kA of rf current drive, with up to 40% of the plasma current driven noninductively. The success of these experiments at 60 MHz with a 2 MW transmitter source capability has led to a major upgrade of the FWCD system. Two additional transmitters, 30 to 120 MHz, with a 2 MW source capability each, will be added together with two new four-strap antennas in early 1994. Another major thrust of the DIII-D program is to develop advanced tokamak modes of operation, simultaneously demonstrating improvements in confinement and stability in quasi-steady-state operation. In some of the initial advanced tokamak experiments on DIII-D with neutral beam heated (NBI) discharges it has been demonstrated that energy confinement time can be improved by rapidly elongating the plasma to force the current density profile to be more centrally peaked. However, this high-l i phase of the discharge with the commensurate improvement in confinement is transient as the current density profile relaxes. By applying FWCD to the core of such a κ-ramped discharge it may be possible to sustain the high internal inductance and elevated confinement. Using computational tools validated on the initial DIII-D FWCD experiments we find that such a high-l i advanced tokamak discharge should be capable of sustainment at the 1 MA level with the upgraded capability of the FWCD system. (author) 16 refs., 3 figs., 1 tab

  9. Basic principle of constant q/sub a/ current build-up in tokamaks

    International Nuclear Information System (INIS)

    Kikuchi, M.

    1985-05-01

    An analytic expression is derived such that the current profile shape is kept constant during the current build-up phase in tokamaks. The required conductivity profile is parametrized by two externally controllable parameters, I/sub p/ and a/sub p/ in the case of the Gaussian current profile. It is shown that a Gaussian current profile can be maintained for a realistically broad conductivity profile by using the constant q/sub a/ current build-up method even under the condition of a high I/sub p/

  10. Bootstrap and fast wave current drive for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1991-09-01

    Using the multi-species neoclassical treatment of Hirshman and Sigmar we study steady state bootstrap equilibria with seed currents provided by low frequency (ICRF) fast waves and with additional surface current density driven by lower hybrid waves. This study applies to reactor plasmas of arbitrary aspect ratio. IN one limit the bootstrap component can supply nearly the total equilibrium current with minimal driving power ( o = 18 MA needs P FW = 15 MW, P LH = 75 MW). A computational survey of bootstrap fraction and current drive efficiency is presented. 11 refs., 8 figs

  11. A continuous winding scheme for superconducting tokamak coils with cable-in-conduit conductor

    International Nuclear Information System (INIS)

    Kim, Sang-ho; Chung, Kie-hyung; Lee, Deok Kyo

    2001-01-01

    Superconducting magnet coils are essential for steady-state or long-pulse operation of tokamaks. In an advanced tokamak, the central solenoid (CS) coils are usually divided into several pairs of modules to provide for an extra plasma shaping capability in addition to those available from the shaping (poloidal field) coils. In the conventional pancake winding scheme of superconducting coils, each coil consists of separate superconducting 'double-pancake' coils connected together in series; however, such joints are not superconducting, which is one of the major disadvantages, especially in pulsed operations. A new type of winding was adopted for the ITER CS coil, which consists of cylindrical shell 'layers' joined in series. A disadvantage of this layer winding is its inability to yield modular coils that can provide certain degree of plasma shaping. Joints can be removed in a coil winding pack with the conventional pancake winding scheme, if the conductor is sufficiently long and the winding machine is properly equipped. The compactness, however, cannot be preserved with this scheme. The winding compactness is important since the radial build of the CS coils is one of the major parameters that determine the machine size. In this paper, we present a continuous winding scheme that requires no joints, allows coil fabrication at minimum dimension, and meets the flux swing requirement and other practical aspects

  12. Current drive experiments in the HIT-II spherical tokamak

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Gu, P.; Isso, V.A.; Jewell, P.E.; McCollam, K.J.; Nelson, B.A.; Ramon, R.; Redd, A.J.; Sieck, P.E.; Smith, R.J.; Nagata, M.; Uyama, T.

    2001-01-01

    The Helicity Injected Torus (Hit) program has made progress in understanding relaxation and helicity injection current drive. Helicity-conserving MHD activity during the inductive (Ohmic) current ramp demonstrates the profile flattening needed for coaxial helicity injection (CHI). Results from cathode and anode central column (CC) CHI pulses are consistent with the electron locking model of current drive from a pure n=1 mode. Finally, low density CHI, compatible with Ohmic operation, has been achieved. Some enhancement of CHI discharges with the application of Ohmic is shown. (author)

  13. Tearing modes in tokamaks with lower hybrid current drive

    International Nuclear Information System (INIS)

    Xu, X.Q.

    1990-08-01

    In this paper, the effect of current drive on the tearing modes in the semi-collisional regime is analyzed using the drift-kinetic equation. A collisional operator is developed to model electron parallel conductivity. For the pure tearing modes the linear and quasilinear growth rates in the Rutherford regimes have been found to have roughly the same forms with a modified resistivity as without current drive. One interesting result is the prediction of a new instability. This instability, driven by the current gradient inside the tearing mode layer, is possibly related to MHD behavior observed in these experiments. 9 refs

  14. Monte Carlo simulation of lower hybrid current drive in tokamaks

    International Nuclear Information System (INIS)

    Sipilae, S.K.; Heikkinen, J.A.

    1994-01-01

    In the report a method for noninductive current drive studies based on three-dimensional simulation of test particle orbits is presented. A Monte Carlo momentum diffusion operator is developed to model the wave-particle interaction. The scheme can be utilised in studies of current drive efficiency as well as in examining the current density profiles caused by waves with a finite parallel wave number spectrum and a nonuniform power deposition profile in a toroidal configuration space of arbitrary shape. Calculations performed with a uniform poorer deposition profile of lower hybrid waves for axisymmetric magnetic configurations having different aspect ratios and poloidal cross-section shape confirm the semianalytic estimates for the current drive efficiency based on the solutions of the flux surface averaged Fokker-Planck equation for configurations with circular poloidal cross section. The consequences of the combined effect of radial diffusion, magnetic trapping and radially nonhomogeneous power deposition and background plasma parameter profiles are investigated

  15. A relativistic model of electron cyclotron current drive efficiency in tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Lin-Liu Y.R.

    2012-09-01

    Full Text Available A fully relativistic model of electron cyclotron current drive (ECCD efficiency based on the adjoint function techniques is considered. Numerical calculations of the current drive efficiency in a tokamak by using the variational approach are performed. A fully relativistic extension of the variational principle with the modified basis functions for the Spitzer function with momentum conservation in the electron-electron collision is described in general tokamak geometry. The model developed has generalized that of Marushchenko’s (N.B . Marushchenko, et al. Fusion Sci. & Tech., 2009, which is extended for arbitrary temperatures and covers exactly the asymptotic for u ≫ 1 when Z → ∞, and suitable for ray-tracing calculations.

  16. A Finite Element Versus Analytical Approach to the Solution of the Current Diffusion Equation in Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Šesnic, S.; Dorić, V.; Poljak, D.; Šušnjara, A.; Artaud, J.F.

    2018-01-01

    Roč. 46, č. 4 (2018), s. 1027-1034 ISSN 0093-3813 R&D Projects: GA MŠk(CZ) 8D15001 Institutional support: RVO:61389021 Keywords : Finite element analysis * Tokamaks * current diffusion equation (CDE) * finite-element method (FEM) Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 1.052, year: 2016

  17. Continuous measurement of an atomic current

    Science.gov (United States)

    Laflamme, C.; Yang, D.; Zoller, P.

    2017-04-01

    We are interested in dynamics of quantum many-body systems under continuous observation, and its physical realizations involving cold atoms in lattices. In the present work we focus on continuous measurement of atomic currents in lattice models, including the Hubbard model. We describe a Cavity QED setup, where measurement of a homodyne current provides a faithful representation of the atomic current as a function of time. We employ the quantum optical description in terms of a diffusive stochastic Schrödinger equation to follow the time evolution of the atomic system conditional to observing a given homodyne current trajectory, thus accounting for the competition between the Hamiltonian evolution and measurement back action. As an illustration, we discuss minimal models of atomic dynamics and continuous current measurement on rings with synthetic gauge fields, involving both real space and synthetic dimension lattices (represented by internal atomic states). Finally, by "not reading" the current measurements the time evolution of the atomic system is governed by a master equation, where—depending on the microscopic details of our CQED setups—we effectively engineer a current coupling of our system to a quantum reservoir. This provides interesting scenarios of dissipative dynamics generating "dark" pure quantum many-body states.

  18. Intrinsic non-inductive current driven by ETG turbulence in tokamaks

    Science.gov (United States)

    Singh, Rameswar; Kaw, P. K.; Singh, R.; Gürcan, Ã.-. D.

    2017-10-01

    Motivated by observations and physics understanding of the phenomenon of intrinsic rotation, it is suggested that similar considerations for electron dynamics may result in intrinsic current in tokamaks. We have investigated the possibility of intrinsic non-inductive current in the turbulent plasma of tokamaks. Ohm's law is generalized to include the effect of turbulent fluctuations in the mean field approach. This clearly leads to the identification of sources and the mechanisms of non-inductive current drive by electron temperature gradient turbulence. It is found that a mean parallel electro-motive force and hence a mean parallel current can be generated by (1) the divergence of residual current flux density and (2) a non-flux like turbulent source from the density and parallel electric field correlations. Both residual flux and the non-flux source require parallel wave-number k∥ symmetry breaking for their survival which can be supplied by various means like mean E × B shear, turbulence intensity gradient, etc. Estimates of turbulence driven current are compared with the background bootstrap current in the pedestal region. It is found that turbulence driven current is nearly 10% of the bootstrap current and hence can have a significant influence on the equilibrium current density profiles and current shear driven modes.

  19. 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.

  20. Numerical simulations of the radio-frequency-driven toroidal current in tokamaks

    International Nuclear Information System (INIS)

    Peysson, Y.; Decker, J.

    2014-01-01

    Radio-frequency (rf) waves are a powerful tool for improving the performance and stability of tokamak plasmas through heating and current drive mechanisms, allowing current density profile control and steady-state operation. From first principles, and taking advantage from the ordering between the various time and space scales, fast and powerful numerical tools have been developed to calculate the rf-driven current. The current drive problem in tokamaks is first introduced with the purpose of maintaining a steady-state self-organized toroidal magnetohydrodynamic equilibrium, such that a minimal amount of the fusion power has to be recycled to control the plasma current. The strict criterion that characterizes a steady-state discharge is derived from the response of the tokamak, considered as a transformer, and of the plasma, when an external source of current is applied. The calculation of a rf-driven source of current requires solving self-consistently a set of equations describing the dynamics of wave fields and charged particles in an inhomogeneous magnetized plasma. The range of applicability of these equations is discussed, as well as numerical methods developed to solve them, such as the ray-tracing code C3PO and the three-dimensional linearized relativistic bounce-averaged electron Fokker-Planck solver LUKE. Simulations of current drive by lower-hybrid waves are presented to illustrate the applications of our numerical tools. Current drive modeling includes the effect of electron density fluctuations at the plasma edge, and the case of electron cyclotron waves used for stabilization of the 3/2 neoclassical tearing modes in ITER is studied in detail. Finally, ongoing developments, including cross effects between momentum and configuration spaces, aiming at improving current drive calculations are discussed. (authors)

  1. Analysis of current diffusive ballooning mode in tokamaks

    International Nuclear Information System (INIS)

    Uchida, M.; Fukuyama, A.; Itoh, S.-I.; Yagi, M.

    1999-12-01

    The effect of finite gyroradius on the current diffusive ballooning mode is examined. Starting from the reduced MHD equations including turbulent transports, coupling with drift motion and finite gyroradius effect of ions, we derive a ballooning mode equation with complex transport coefficients. The eigenfrequency, saturation level and thermal diffusivity are evaluated numerically from the marginal stability condition. Preliminary results of their parameter dependence is presented. (author)

  2. Studies on fast wave current drive in the JAERI tokamaks

    International Nuclear Information System (INIS)

    Kimura, H.; Yamamoto, T.; Fujii, T.; Kawashima, H.; Tamai, H.; Saigusa, M.; Imai, T.; Hamamatsu, K.; Fukuyama, A.

    1991-01-01

    Fast wave electron heating experiment (FWEH) on JFT-2M and JT-60 and analysis of fast wave current drive (FWCD) ability on JT-60U are presented. In the JFT-2M, absorption of fast waves have been investigated by using a phased four-loop antenna array. The absorption of the fast waves has been studied for various plasma parameters by using combination of other additional heating methods such as electron cyclotron heating (ECH) and ion cyclotron heating. It is shown that the absorption efficiency estimated from various methods well correlates with one calculated theoretically in single pass damping. Interaction of the fast waves with fast electrons in combination with ECH has been examined through the measurement of non-thermal electron cyclotron emission (ECE). The observed ECE during FWEH is well explained by the theoretical model, which indicates generation of the appreciable energetic fast electrons by the fast waves. New four-loop array antennas have been employed to improve the absorption of unidirectionally-propagating waves. Characteristics of antenna loading resistance can be reproduced by a coupling calculation code. In JT-60, FWEH experiment in combination with lower hybrid current drive was performed. Power absorption efficiency of fast wave is substantially improved in combination with LHCD of relatively low power for both phasing modes. Bulk electron heating is observed with high-k // mode and coupling with fast electron is confirmed in hard X-ray emission with low-k // mode. The results are consistent with theoretical prediction based on 1.D full wave code. Synergetic effects between FWEH and LHCD are found. Coupling calculation indicates that eight-loop antenna is favourable for keeping high directivity in the required N // -range. Current drive efficiency is calculated with 1-D full wave code including trapped particle effects and higher harmonic ion cyclotron damping

  3. Improved plasma confinement by modulated toroidal current on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Zhao Junyu; Shen Biao; Luo Jiarong

    2004-01-01

    The improved confinement phase was observed during modulating toroidal current on the Hefei superconducting Tokamak-7 (HT-7). This improved plasma confinement phase is characterized by suppressing magnetohydrodynamic (MHD) instabilities effectively, thus increased the central line averaged electron density and the central electron temperature about 33%, out-put steeper density profiles, and reduced hydrogen radiation from the edge as well. The global energy confinement time was increased by 27%-45%; The impurity radiation was reduced by modulation of plasma toroidal current; particle confinement time was increased about two times; a stronger radial negative electric field formed inside the limiter. The radial electric field during modulating current was calculated and disscused. (authors)

  4. Current drive by Alfvacute en waves in elongated cross-section tokamak

    International Nuclear Information System (INIS)

    Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; de Azevedo, C.A.; de Assis, A.S.

    1997-01-01

    The general approach to the Alfvacute en wave current drive problem in tokamaks with elongated transverse cross-sections was considered in this paper. Model approximations are used to describe circulating and trapped particle dynamics. This approach gives the accuracy of some percents. The expressions for the time-averaged longitudinal current and the radio-frequency currents have been obtained. They are supposed to be useful for a further analytical and computational solution of this problem. As an example, kinetic Alfvacute en waves are considered in this paper. copyright 1997 American Institute of Physics

  5. MHD simulations of DC helicity injection for current drive in tokamaks

    International Nuclear Information System (INIS)

    Sovinec, C.R.; Prager, S.C.

    1994-12-01

    MHD computations of DC helicity injection in tokamak-like configurations show current drive with no ''loop voltage'' in a resistive, pressureless plasma. The self-consistently generated current profiles are unstable to resistive modes that partially relax the profile through the MHD dynamo mechanism. The current driven by the fluctuations leads to closed contours of average poloidal flux. However, the 1% fluctuation level is large enough to produce a region of stochastic magnetic field. A limited Lundquist number (S) scan from 2.5 x 10 3 to 4 x 10 4 indicates that both the fluctuation level and relaxation increase with S

  6. Soft x-ray camera for internal shape and current density measurements on a noncircular tokamak

    International Nuclear Information System (INIS)

    Fonck, R.J.; Jaehnig, K.P.; Powell, E.T.; Reusch, M.; Roney, P.; Simon, M.P.

    1988-05-01

    Soft x-ray measurements of the internal plasma flux surface shaped in principle allow a determination of the plasma current density distribution, and provide a necessary monitor of the degree of internal elongation of tokamak plasmas with a noncircular cross section. A two-dimensional, tangentially viewing, soft x-ray pinhole camera has been fabricated to provide internal shape measurements on the PBX-M tokamak. It consists of a scintillator at the focal plane of a foil-filtered pinhole camera, which is, in turn, fiber optically coupled to an intensified framing video camera (/DELTA/t />=/ 3 msec). Automated data acquisition is performed on a stand-alone image-processing system, and data archiving and retrieval takes place on an optical disk video recorder. The entire diagnostic is controlled via a PDP-11/73 microcomputer. The derivation of the polodial emission distribution from the measured image is done by fitting to model profiles. 10 refs., 4 figs

  7. Current density and continuity in discretized models

    International Nuclear Information System (INIS)

    Boykin, Timothy B; Luisier, Mathieu; Klimeck, Gerhard

    2010-01-01

    Discrete approaches have long been used in numerical modelling of physical systems in both research and teaching. Discrete versions of the Schroedinger equation employing either one or several basis functions per mesh point are often used by senior undergraduates and beginning graduate students in computational physics projects. In studying discrete models, students can encounter conceptual difficulties with the representation of the current and its divergence because different finite-difference expressions, all of which reduce to the current density in the continuous limit, measure different physical quantities. Understanding these different discrete currents is essential and requires a careful analysis of the current operator, the divergence of the current and the continuity equation. Here we develop point forms of the current and its divergence valid for an arbitrary mesh and basis. We show that in discrete models currents exist only along lines joining atomic sites (or mesh points). Using these results, we derive a discrete analogue of the divergence theorem and demonstrate probability conservation in a purely localized-basis approach.

  8. On ray stochasticity during lower hybrid current drive in tokamaks

    International Nuclear Information System (INIS)

    Bizarro, J.P.; Moreau, D.

    1992-08-01

    A comprehensive and detailed analysis is presented on the importance of toroidally induced ray stochasticity for the modelling of lower hybrid current drive and for the dynamics of the launched power spectrum. A combined ray tracing and Fokker-Planck code is used and the injected lower hybrid power distribution in poloidal angle and in parallel wave index is accurately represented by taking into account the poloidal extent of the antenna ad by efficiently covering the full range of its radiated spectrum. The importance of the balance between the wave damping and the exponential divergence of nearby ray trajectories in determining the shape of the predicted lower hybrid power deposition profiles is emphasized. When a sufficiently large number of rays is used to densely cover the region of the launched power spectrum which is affected by stochastic effects, code predictions are shown to be stable with respect to small changes in initial conditions and plasma parameters and to be consistent with experimental data

  9. High current superconductors for tokamak toroidal field coils

    International Nuclear Information System (INIS)

    Fietz, W.A.

    1976-01-01

    Conductors rated at 10,000 A for 8 T and 4.2 K are being purchased for the first large coil segment tests at ORNL. Requirements for these conductors, in addition to the high current rating, are low pulse losses, cryostatic stability, and acceptable mechanical properties. The conductors are required to have losses less than 0.4 W/m under pulsed fields of 0.5 T with a rise time of 1 sec in an ambient 8-T field. Methods of calculating these losses and techniques for verifying the performance by direct measurement are discussed. Conductors stabilized by two different cooling methods, pool boiling and forced helium flow, have been proposed. Analysis of these conductors is presented and a proposed definition and test of stability is discussed. Mechanical property requirements, tensile and compressive, are defined and test methods are discussed

  10. 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)

  11. 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.)

  12. The effect of non-inductive current drive on tokamak transport

    International Nuclear Information System (INIS)

    Helander, P; Akers, R J; Valovic, M; Peysson, Y

    2005-01-01

    Non-inductive current drive causes cross-field neoclassical transport in a tokamak, in much the same way that the toroidal electric field used to drive the plasma current produces the so-called Ware pinch. This transport can be either inwards or outwards, depending on the current drive mechanism, and can be either larger or smaller than the analogous Ware pinch. A Green's function formalism is used to calculate the transport produced by wave-driven currents, which is found to be inwards for electron-cyclotron and lower-hybrid current drive. Its magnitude is proportional to the collisionality of the current-carrying electrons and therefore smaller than the Ware pinch when the resonant electrons are suprathermal. In contrast, neutral-beam current drive produces outward particle transport when the beams are injected in the same toroidal direction as the plasma current, and inward particle transport otherwise. This transport is somewhat larger than the corresponding Ware pinch. Together, they may explain an observation made on several tokamaks over the years, most recently on MAST, that density profiles tend to be more peaked during counter-injection

  13. Beta and current limits in the Doublet III tokamak

    International Nuclear Information System (INIS)

    Strait, E.J.; Chu, M.S.; Jahns, G.L.

    1986-04-01

    Neutral-beam heated discharges in Doublet III exhibit an operational beta limit, β/sub T/(%) less than or equal to 3.5 I(MA)/a(m)B(T), in good agreement with several theoretical predictions for ideal external kink or ballooning modes. These theories predict that the β limit has no explicit dependence on plasma shape (for nominal dee shapes). This aspect of the theory was confirmed in Doublet III by varying the elongation (kappa) from 1.0 to 1.6 and the triangularity (delta) from -0.1 to 0.9 and finding in all cases the same β limit. The maximum achievable beta thus depends on the minimum achievable value of the safety factor q. In Doublet III, the operational current limit is given by q greater than or equal to 1.7 for limiter-defined discharges and q greater than or equal to 2.7 for separatrix-defined discharges. Operation with q approx.2 was achieved for 1.0 less than or equal to kappa less than or equal to 1.6. Both β and q limits are characterized by major disruptions which usually terminate the discharge. In both cases, the disruptions often have a precursor oscillation with toroidal mode number n = 1, poloidal mode number m = 2 or 3, a frequency of zero to a few kHz, and a growth time on the order of a millisecond. These observations suggest that the proximate cause of these disruptions is a kink or tearing mode, pressure-driven in one case and current-driven in the other. Theoretical analyses of discharges at both limits will be compared. Modes with a high toroidal mode number, 3 less than or equal to n less than or equal to 5, and ballooning character have been observed near the β/sub T/ limit. These modes do not appear to be closely connected with the disruptions. Heating efficiency, ΔW/ΔP, remains constant up to the limiting disruption. Fishbone modes appear to be mainly a feature of high β/sub p/ operation and not connected to the β/sub T/ limit

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

    International Nuclear Information System (INIS)

    Laborde, L.

    2005-12-01

    Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)

  15. Numerical analysis on the synergy between electron cyclotron current drive and lower hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Chen, S Y; Hong, B B; Liu, Y; Lu, W; Huang, J; Tang, C J; Ding, X T; Zhang, X J; Hu, Y J

    2012-01-01

    The synergy between electron cyclotron current drive (ECCD) and lower hybrid current drive (LHCD) is investigated numerically with the parameters of the HL-2A tokamak. Based on the understanding of the synergy mechanisms, a high current driven efficiency or a desired radial current profile can be achieved through properly matching the parameters of ECCD and LHCD due to the flexibility of ECCD. Meanwhile, it is found that the total current driven by the electron cyclotron wave (ECW) and the lower hybrid wave (LHW) simultaneously can be smaller than the sum of the currents driven by the ECW and LHW separately, when the power of the ECW is much larger than the LHW power. One of the reasons leading to this phenomenon (referred to as negative synergy in this context) is that fast current-carrying electrons tend to be trapped, when the perpendicular velocity driven by the ECW is large and the parallel velocity decided by the LHW is correspondingly small. (paper)

  16. Investigation of the LH wave energy conversion and current drive efficiency in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Chen, Z.Y.; Wan, B.N.; Shi, Y.J.; Lin, S.Y.; Hu, L.Q.; Asif, M.

    2005-01-01

    Lower hybrid current drive (LHCD) plasmas in the presence of DC electric filed have been investigated based on Karney-Fisch theory in the HT-7 tokamak. The relatively small scatter in the experimental data with various values of waveguide phasing and lower hybrid power, when plotted in the Karney-Fisch diagram, confirms that a reasonable theoretical interpretation is possible for the HT-7 data. The full non-inductively current drive efficiencies are obtained by fitting the experimental data to the theoretical curve. The efficiency strongly depends on the lower hybrid wave phase velocity

  17. Finite Larmor radius effects on Alfven wave current drive in low-aspect ratio tokamaks

    International Nuclear Information System (INIS)

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

    1998-01-01

    Alfven wave current drive (AWCD) in low-aspect ratio (A≡R/a=1/ε > or approx. 1) tokamaks (LARTs) is studied numerically. For this, the full-wave equation (E parallel ≠0) with a Vlasov-based dielectric tensor is solved by relaxation techniques, subject to appropriate boundary conditions at the plasma centre and at the plasma-vacuum interface, as well as the concentric antenna current sheet and at the external metallic wall. A systematic investigation of the physical characteristics of the AWCD generated in LARTs when kinetic effects are considered is carried out and illustrative results are presented and discussed. (author)

  18. 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

  19. Study of lower hybrid current drive system in tokamak fusion devices

    International Nuclear Information System (INIS)

    Maebara, Sunao

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 μP is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 μsec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10 -6 Pam 3 /sm 2 at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm -2 (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10 -5 Pam 3 /sm 2 which is low enough for an antenna material. (author)

  20. 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)

  1. Design of the power supply system for the plasma current modulation on J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, M.; Shao, J.; Ma, S.X., E-mail: mashaoxiang@hust.edu.cn; Liang, X.; Yu, K.X.; Pan, Y.

    2016-10-15

    Highlights: • A modification scheme of heating field power supply system for plasma current modulation. • High-power fast control power supply with multilevel cascade circuit. • Restraining circulating current with coupled inductors in cyclic symmetric structure. - Abstract: In order to further study the influence of current modulation parameters on suppressing tearing instability, the plasma current should be modulated in a wider range. So a modification scheme is designed to improve the performance of ohmic heating power supply system on J-TEXT tokamak. A multilevel cascade circuit with carrier phase-shifted PWM technique has been proposed. Coupled inductors are connected in the form of cyclic symmetry to restrain the circulating current caused by multiple paralleled branches. The simulation proves this proposed current modulation power supply system matches output requirement and achieves good current sharing effect. Finally, a prototype is designed, and the experiment results can verify the correctness of the simulation model well.

  2. Non-existence of Normal Tokamak Equilibria with Negative Central Current

    International Nuclear Information System (INIS)

    Hammett, G.W.; Jardin, S.C.; Stratton, B.C.

    2003-01-01

    Recent tokamak experiments employing off-axis, non-inductive current drive have found that a large central current hole can be produced. The current density is measured to be approximately zero in this region, though in principle there was sufficient current-drive power for the central current density to have gone significantly negative. Recent papers have used a large aspect-ratio expansion to show that normal MHD equilibria (with axisymmetric nested flux surfaces, non-singular fields, and monotonic peaked pressure profiles) can not exist with negative central current. We extend that proof here to arbitrary aspect ratio, using a variant of the virial theorem to derive a relatively simple integral constraint on the equilibrium. However, this constraint does not, by itself, exclude equilibria with non-nested flux surfaces, or equilibria with singular fields and/or hollow pressure profiles that may be spontaneously generated

  3. Minimizing the magnetohydrodynamic potential energy for the current hole region in tokamaks

    International Nuclear Information System (INIS)

    Chu, M.S.; Parks, P.B.

    2004-01-01

    The current hole region in the tokamak has been observed to arise naturally during the development of internal transport barriers. The magnetohydrodynamic (MHD) potential energy in the current hole region is shown to be determined completely in terms of the displacements at the edge of the current hole. For modes with finite toroidal mode number n≠0, the minimized potential energy is the same as if the current hole region were a vacuum region. For modes with toroidal mode number n=0, the displacement is a superposition of three types of independent displacements: a vertical displacement or displacements that compress only the plasma, or the toroidal field uniformly. Thus for ideal MHD perturbations of plasma with a current hole, the plasma behaves as if it were bordered by an extra ''internal vacuum region.'' The relevance of the present work to computer simulations of plasma with a current hole region is also discussed

  4. MINIMIZING THE MHD POTENTIAL ENERGY FOR THE CURRENT HOLE REGION IN TOKAMAKS

    International Nuclear Information System (INIS)

    CHU, M.S; PARKS, P.B

    2004-01-01

    The current hole region in the tokamak has been observed to arise naturally during the development of internal transport barriers. The magnetohydrodynamic (MHD) potential energy in the current hole region is shown to be determined completely in terms of the displacements at the edge of the current hole. For modes with finite toroidal mode number n ≠ 0, the minimized potential energy is the same as if the current hole region were a vacuum region. For modes with toroidal mode number n = 0, the displacement is a superposition of three types of independent displacements: a vertical displacement or displacements that compress only the plasma or the toroidal field uniformly. Thus for ideal MHD perturbations of plasma with a current hole, the plasma behaves as if it were bordered by an extra ''internal vacuum region''. The relevance of the present work to computer simulations of plasma with a current hole region is also discussed

  5. Magnetic Diagnostics for Equilibrium Reconstructions in the Presence of Nonaxisymmetric Eddy Current Distributions in Tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J.; Zakharov, L.

    2010-01-01

    The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R 0 = 0.4 m and a = 0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

  6. Feedback control of current drive by using hybrid wave in tokamaks

    International Nuclear Information System (INIS)

    Wijnands, T.J.; CEA Centre d'Etudes de Cadarache, 13 - Saint-Paul-lez-Durance

    1997-03-01

    This work is focussed on an important and recent development in present day Controlled Nuclear Fusion Research and Tokamaks. The aim is to optimise the energy confinement for a certain magnetic configuration by adapting the radial distribution of the current. Of particular interest are feedback control scenarios with stationary modifications of the current profile using current, driven by Lower Hybrid waves. A new feedback control system has been developed for Tore Supra and has made a large number of new operation scenarios possible. In one of the experiments described here, there is no energy exchange between the poloidal field system and the plasma, the current is controlled by the power of the Lower Hybrid waves while the launched wave spectrum is used to optimise the current profile shape and the energy confinement. (author)

  7. 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)

  8. Study on paralleled inverters with current-sharing coupled inductors on J-TEXT Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Shao, J.; Rao, B., E-mail: borao@hust.edu.cn; Zhang, M.; Ma, S.X.; Liang, X.; Yu, K.X.; Pan, Y.

    2016-12-15

    Highlights: • A modification scheme of heating field power supply system for plasma current modulation. • High-power fast control power supply with multilevel cascade circuit. • Restraining circulating current with coupled inductors in cyclic symmetric structure. • Analysis on the topology with current-sharing coupled inductors. - Abstract: The coupled inductors in paralleled inverters are applied to restrain the high frequency circulating current on J-TEXT Tokamak. Compared with individual inductor, this method has the benefit of high voltage utilization, less volume and weight of the inductor. In this paper, circuit topology of coupled inductors in cyclic symmetry structure for steady-state operation is analyzed and then the design of the inductor is introduced. The maximum circulating current is related to number of parallel branch, DC side voltage, self-inductance of the inductor and the frequency of carrier wave. The simulation and prototype experiment results verify the design.

  9. Generation of noninductive current by electron-Bernstein waves on the COMPASS-D Tokamak.

    Science.gov (United States)

    Shevchenko, V; Baranov, Y; O'Brien, M; Saveliev, A

    2002-12-23

    Electron-Bernstein waves (EBW) were excited in the plasma by mode converted extraordinary (X) waves launched from the high field side of the COMPASS-D tokamak at different toroidal angles. It has been found experimentally that X-mode injection perpendicular to the magnetic field provides maximum heating efficiency. Noninductive currents of up to 100 kA were found to be driven by the EBW mode with countercurrent drive. These results are consistent with ray tracing and quasilinear Fokker-Planck simulations.

  10. Magnetic ripple and the modeling of lower-hybrid current drive in tokamaks

    International Nuclear Information System (INIS)

    Peysson, Y.; Arslanbekov, R.; Basiuk, V.; Carrasco, J.; Litaudon, X.; Moreau, D.; Bizarro, J.P.

    1996-01-01

    Using ray-tracing, a detailed investigation of the lower hybrid (LH) wave propagation in presence of toroidal magnetic field ripple is presented. By coupling ray tracing with a one-dimensional relativistic Fokker-Planck code, simulations of LH experiments have been performed for the Tore Supra tokamak. Taking into account magnetic ripple in LH simulations, a better agreement is found between numerical predictions and experimental observations, such as non-thermal Bremsstrahlung emission, current profile, ripple-induced power losses in local magnetic mirrors, when plasma conditions correspond to the ' 'few passes' regime. (author)

  11. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    Czech Academy of Sciences Publication Activity Database

    Urban, Jakub; Decker, J.; Peysson, Y.; Preinhaelter, Josef; Shevchenko, V.; Taylor, G.; Vahala, L.; Vahala, G.

    2011-01-01

    Roč. 51, č. 8 (2011), 083050-083050 ISSN 0029-5515 R&D Projects: GA ČR GA202/08/0419; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : spherical tokamak * electron Bernstein wave (EBW) * heating * current drive * electron cyclotron wave Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/8/083050/pdf/0029-5515_51_8_083050.pdf

  12. Lower hybrid current drive experiments with graphite limiters in the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Liu, J.; Gao, X.; Hu, L.Q.; Asif, M.; Chen, Z.Y.; Ding, B.J.; Zhou, Q.; Liu, H.Q.; Jie, Y.X.; Kong, W.; Lin, S.Y.; Ding, Y.H.; Gao, L.; Xu, Q.

    2006-01-01

    Recent progress of lower hybrid (LH) experiments with new graphite limiters configuration in the HT-7 tokamak is presented. The lower hybrid current drive (LHCD) efficiency can be determined by fitting based on experimental data. Improved particle confinement was observed via LHCD (P LHW >300 kW) characterized by the particle confinement time τ p increased about 1.56 times. It is found that runaways are suppressed during loop voltage is decreasing at the flat-top phase of LH discharges. The main limitations of pulse length are presented in long-pulse experiments with new limiter configuration

  13. Inter-ELM evolution of the edge current density profile on the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Dunne, Michael G.

    2014-01-01

    The sudden decrease of plasma stored energy and subsequent power deposition on the first wall of a tokamak device due to edge localised modes (ELMs) is potentially detrimental to the success of a future fusion reactor. Understanding and control of ELMs is critical for the longevity of these devices and also to maximise their performance. The commonly accepted picture of ELMs posits a critical pressure gradient and current density in the plasma edge, above which coupled magnetohydrodynamic (MHD) peeling-ballooning modes are driven unstable. Much analysis has been presented in recent years on the spatial and temporal evolution of the edge pressure gradient. However, the edge current density has typically been overlooked due to the difficulties in measuring this quantity. In this thesis, a novel method of current density recovery is presented, using the equilibrium solver CLISTE to reconstruct a high resolution equilibrium utilising both external magnetic and internal edge kinetic data measured on the ASDEX Upgrade (AUG) tokamak. The evolution of the edge current density relative to an ELM crash is presented, showing that a resistive delay in the buildup of the current density is unlikely. An uncertainty analysis shows that the edge current density can be determined with an accuracy consistent with that of the kinetic data used. A comparison with neoclassical theory demonstrates excellent agreement between the current density determined by CLISTE and the calculated profiles. Three ELM mitigation regimes are investigated: Type-II ELMs, ELMs suppressed by external magnetic perturbations (MPs), and Nitrogen seeded ELMs. In the first two cases, the current density is found to decrease as mitigation onsets, indicating a more ballooning-like plasma behaviour. In the latter case, the flux surface averaged current density can decrease while the local current density increases, thus providing a mechanism to suppress both the peeling and ballooning modes.

  14. Inter-ELM evolution of the edge current density profile on the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Dunne, Michael G.

    2014-02-15

    The sudden decrease of plasma stored energy and subsequent power deposition on the first wall of a tokamak device due to edge localised modes (ELMs) is potentially detrimental to the success of a future fusion reactor. Understanding and control of ELMs is critical for the longevity of these devices and also to maximise their performance. The commonly accepted picture of ELMs posits a critical pressure gradient and current density in the plasma edge, above which coupled magnetohydrodynamic (MHD) peeling-ballooning modes are driven unstable. Much analysis has been presented in recent years on the spatial and temporal evolution of the edge pressure gradient. However, the edge current density has typically been overlooked due to the difficulties in measuring this quantity. In this thesis, a novel method of current density recovery is presented, using the equilibrium solver CLISTE to reconstruct a high resolution equilibrium utilising both external magnetic and internal edge kinetic data measured on the ASDEX Upgrade (AUG) tokamak. The evolution of the edge current density relative to an ELM crash is presented, showing that a resistive delay in the buildup of the current density is unlikely. An uncertainty analysis shows that the edge current density can be determined with an accuracy consistent with that of the kinetic data used. A comparison with neoclassical theory demonstrates excellent agreement between the current density determined by CLISTE and the calculated profiles. Three ELM mitigation regimes are investigated: Type-II ELMs, ELMs suppressed by external magnetic perturbations (MPs), and Nitrogen seeded ELMs. In the first two cases, the current density is found to decrease as mitigation onsets, indicating a more ballooning-like plasma behaviour. In the latter case, the flux surface averaged current density can decrease while the local current density increases, thus providing a mechanism to suppress both the peeling and ballooning modes.

  15. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  16. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  17. Fast computational scheme for feedback control of high current fusion tokamaks

    International Nuclear Information System (INIS)

    Dong, J.Q.; Khayrutdinov, R.; Azizov, E.; Jardin, S.

    1992-01-01

    An accurate and fast numerical model of tokamak plasma evolution is presented. In this code (DINA) the equilibrium problem of plasmas with free boundaries in externally changing magnetic fields is solved simultaneously with the plasma transport equation. The circuit equations are solved for the vacuum vessel and passive and active coils. The code includes pellet injection, neutral beam heating, auxiliary heating, and alpha particle heating. Bootstrap and beam-driven plasma currents are accounted for. An inverse variable technique is utilized to obtain the coordinates of the equilibrium magnetic surfaces. This numerical algorithm permits to determine the flux coordinates very quickly and accurately. The authors show that using the fully resistive MHD analysis the region of stability (to vertical motions) is wider than using the rigid displacement model. Comparing plasma motions with the same gain, it is seen that the plasma oscillates more in the rigid analysis than in the MHD analysis. They study the influence of the pick up coil's location and the possibility of control of the plasma vertical position. They use a simple modification of the standard control law that enables the control of the plasma with pick up coils located at any position. This flexibility becomes critical in the design of future complex high current tokamak systems. The fully resistive MHD model permits to obtain accurate estimates of the plasma response. This approach yields computational time savings of one to two orders of magnitude with respect to other existing MHD models. In this sense, conventional numerical algorithms do not provide suitable models for application of modern control techniques into real time expert systems. The proposed inverse variable technique is rather suitable for incorporation in a comprehensive expert system for feedback control of fusion tokamaks in real time

  18. Measurement of plasma current in Tokamaks using an optical fibre reflectometry technique

    Directory of Open Access Journals (Sweden)

    Wuilpart Marc

    2018-01-01

    Full Text Available An optical time-domain reflectometer sensitive to the polarization of light is proposed for the measurement of plasma current in the Tore Supra fusion reactor. The measurement principle relies on the Faraday effect i.e. on the generation of a circular birefringence along an optical fiber subject to an axial magnetic field. The circular birefringence induces a polarization rotation that can be mapped along the fiber thanks to an opticaltime domain reflectometer followed by an linear polarizer. A proper fitting of the measurement trace then allows determining the applied plasma current. The sensor has been experimentally validated on the Tore Supra tokamak fusion reactor for a plasma current range going from 0.6 to 1.5 MA. A maximum error of 13.50% has been observed for the lowest current.

  19. Axisymmetric disruption dynamics including current profile changes in the ASDEX-Upgrade tokamak

    International Nuclear Information System (INIS)

    Nakamura, Y.; Pautasso, G.; Gruber, O.; Jardin, S.C.

    2002-01-01

    Axisymmetric MHD simulations have revealed a new driving mechanism that governs the vertical displacement event (VDE) dynamics in tokamak disruptions. A rapid flattening of the plasma current profile during the disruption plays a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect, due to an abrupt change in the current profile, is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the plasma current quench, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)

  20. A poloidal non-uniformity of the collisionless parallel current in a tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Romannikov, A.; Fenzi-Bonizec, C

    2005-07-01

    The collisionless distortion of the ion (electron) distribution function at certain points on a magnetic surface is studied in the framework of a simple model of a large aspect ratio tokamak plasma. The flow velocity driven by this distortion is calculated. The possibility of an additional non-uniform collisionless parallel current density on a magnetic surface, other than the known neo-classical non-uniformity is shown. The difference between the parallel current density on the low and high field side of a magnetic surface is close to the neoclassical bootstrap current density. The first Tore-Supra experimental test indicates the possibility of the poloidal non-uniformity of the parallel current density. (authors)

  1. 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

  2. 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.

  3. Fast-wave current drive modelling for large non-circular tokamaks

    International Nuclear Information System (INIS)

    Batchelor, D.B.; Goldfinger, R.C.; Jaeger, E.F.; Carter, M.D.; Swain, D.W.; Ehst, D.; Karney, C.F.F.

    1990-01-01

    It is widely recognized that a key element in the development of an attractive tokamak reactor, and in the successful achievement of the mission of ITER, is the development of an efficient steady-state current drive technique. Fast waves in the ion cyclotron range of frequencies hold the promise to drive steady-state currents with the required efficiency and to effectively heat the plasma to ignition. Advantages over other heating and current drive techniques include low cost per watt and the ability to penetrate to the center of high-density plasmas. The primary issues that must be resolved are: can an antenna array be designed to radiate the required spectrum of waves and have adequate coupling properties? Will the rf power be efficiently absorbed by electrons in the desired velocity range without unacceptable parasitic damping by fuel ions or α particles? What will the efficiency of current drive be when toroidal effects such as trapped particles are included? Can a practical rf system be designed and integrated into the device? We have addressed these issues by performing extensive calculations with ORION, a 2-D code, and the ray tracing code RAYS, which calculate wave propagation, absorption and current drive in tokamak geometry, and with RIP, a 2-D code that self-consistently calculates current drive with MHD equilibrium. An important figure of merit in this context is the integrated, normalized current drive efficiency. The calculations that we present here emphasize the ITER device. We consider a low-frequency scenario such that no ion resonances appear in the machine, and a high-frequency scenario such that the deuterium second harmonic resonance is just outside the plasma and the tritium second harmonic is in the plasma, midway between the magnetic axis and the inside edge. In both cases electron currents are driven by combined TTMP and Landau damping of the fast waves

  4. 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

  5. Non-inductive current drive and RF heating in SST-1 tokamak

    International Nuclear Information System (INIS)

    2000-01-01

    Steady state superconducting tokamak (SST-1) machine is being developed for 1000 sec operation at different operating parameters. Radio Frequency (RF) and neutral beam injection (NBI) methods are planned in SST-1 for noninductive current drive and heating. In this paper, we describe the non-inductive current drive and RF heating methods that are being developed for this purpose. SST-1 is a large aspect ratio tokamak configured to run double-null divertor plasmas with significant elongation (κ = 1.7-1.9) and triangularity (δ = 0.4-0.7). SST-1 has a major radius of 1.1 in and minor radius of 0.2 m. Circular and shaped plasma experiments would be conducted at 1.5 and 3 T toroidal magnetic field in three different phases with I p = 110 kA and 220 kA. Two main factors have been considered during the development of auxiliary systems, namely, high heat flux (1 MW/m 2 ) incident on the plasma facing antennae components and fast feedback for constant power input due to small energy confinement time (∼ 10 ms). (author)

  6. On the generation of Alfven wave current drive in low aspect ratio Tokamaks with neoclassical conductivity

    International Nuclear Information System (INIS)

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

    1998-01-01

    Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high β, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E p arallel=0) MHD and non-ideal, resistive (E p arallel≠0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E p arallel≠0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)

  7. On the generation of Alfven wave current drive in low aspect ratio Tokamaks with neoclassical conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Bruma, C.; Cuperman, S.; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, Tel Aviv (Israel)

    1998-08-01

    Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high {beta}, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E{sub p}arallel=0) MHD and non-ideal, resistive (E{sub p}arallel{ne}0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E{sub p}arallel{ne}0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)

  8. Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

    Science.gov (United States)

    Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

    2017-10-01

    Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

  9. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    Science.gov (United States)

    Urban, Jakub; Decker, Joan; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

    2011-08-01

    The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs) operate generally in high-β regimes, in which the usual EC O- and X-modes are cut off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves—controllable localized H&CD that can be used for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions, which are the fundamental EBW parameters that can be chosen and controlled. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

  10. Relative merits of size, field, and current on ignited tokamak performance

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1988-01-01

    A simple global analysis is developed to examine the relative merits of size (L = a or R/sub 0 /), field (B/sub 0 /), and current (I) on ignition regimes of tokamaks under various confinement scaling laws. Scalings of key parameters with L, B/sub 0 /, and I are presented at several operating points, including (a) optimal path to ignition (saddle point), (b) ignition at minimum beta, (c) ignition at 10 keV, and (d) maximum performance at the limits of density and beta. Expressions for the saddle point and the minimum conditions needed for ohmic ignition are derived analytically for any confinement model of the form tau/sub E/ ∼ n/sup x/T/sup y/. For a wide range of confinement models, the ''figure of merit'' parameters and I are found to give a good indication of the relative performance of the devices where q* is the cylindrical safety factor. As an illustration, the results are applied to representative ''CIT'' (as a class of compact, high-field ignition tokamaks) and ''Super-JETs'' [a class of large-size (few x JET), low-field, high-current (≥20-MA) devices.

  11. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  12. A study on current density distribution reproduction by bounded-eigenfunction expansion for a tokamak plasma

    International Nuclear Information System (INIS)

    Kurihara, Kenichi

    1997-11-01

    Plasma current density distribution is one of the most important controlled variables to determine plasma performance of energy confinement and stability in a tokamak. However, its reproduction by using magnetic measurements solely is recognized to yield an ill-posed problem. A method to presume the formulas giving profiles of plasma pressure and current has been adopted to regularize the ill-posedness, and hence it has been reported the current density distribution can be reproduced as a solution of Grad-Shafranov equation within a certain accuracy. In order to investigate its strict reproducibility from magnetic measurements in this inverse problem, a new method of 'bounded-eigenfunction expansion' is introduced, and it was found that the reproducibility directly corresponds to the independence of a series of the special function. The results from various investigations in an aspect of applied mathematics concerning this inverse problem are presented in detail. (author)

  13. Modeling of the sawtooth instability in tokamaks using a current viscosity term

    International Nuclear Information System (INIS)

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

    1988-08-01

    We propose a new method for modeling the sawtooth instability and other MHD activity in axisymmetric tokamak transport simulations. A hyper-resistivity (or current viscosity) term is included in the mean field Ohm's law to describe the effects of the three-dimensional fluctuating fields on the evolution of the inverse transform, q, characterizing the mean fields. This term has the effect of flattening the current profile, while dissipating energy and conserving helicity. A fully implicit MHD transport and 2-D toroidal equilibrium code has been developed to calculate the evolution in time of the q-profile and the current profile using this new term. The results of this code are compared to the Kadomtsev reconnection model in the circular cylindrical limit. 17 refs., 8 figs

  14. Cross effects on electron-cyclotron and lower-hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Fidone, I.; Giruzzi, G.; Krivenski, V.; Mazzucato, E.; Ziebell, L.F.

    1986-11-01

    Electron cyclotron resonance current drive in a tokamak plasma in the presence of a lower hybrid tail is investigated using a 2D Fokker-Planck code. For an extraordinary mode at oblique propagation and down-shifted frequency it is shown that the efficiency of electron cyclotron current drive becomes, i) substantially greater than the corresponding efficiency of a Maxwellian plasma at the same bulk temperature, ii) equal or greater than that of the lower hybrid waves, iii) comparable with the efficiency of a Maxwellian plasma at much higher temperature. This enhancement results from a beneficial cross-effect of the two waves on the formation of the current carrying electron tail. (5 fig; 17 refs)

  15. Modelling the effects of the sawtooth instability in tokamaks using a current viscosity term

    International Nuclear Information System (INIS)

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

    1989-01-01

    A new method for modelling the sawtooth instability and other MHD activity in axisymmetric tokamak transport simulations is proposed. A hyper-resistivity (or current viscosity) term is included in the mean field Ohm's law to describe the effects of the three-dimensional fluctuating fields on the evolution of the inverse transform q characterizing the mean fields. This term has the effect of flattening the current profile while dissipating energy and conserving helicity. A fully implicit MHD transport and two-dimensional toroidal equilibrium code has been developed to calculate the evolution in time of the q-profile and the current profile using this new term. The results of this code are compared with the Kadomtsev reconnection model in the circular cylindrical limit. (author). 26 refs, 10 figs

  16. Numerical calculation of high frequency fast wave current drive in a reactor grade tokamak

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi; Hamamatsu, Kiyotaka

    1988-02-01

    A fast wave current drive with a high frequency is estimated for a reactor grade tokamak by the ray tracing and the quasi-linear Fokker-Planck calculations with an assumption of single path absorption. The fast wave can drive RF current with the drive efficiency of η CD = n-bar e (10 19 m -3 )I RC (A)R(m)/P RF (W) ∼ 3.0 when the wave frequency is selected to be f/f ci > 7. A sharp wave spectrum and a ph|| >/υ Te ∼ 3.0 are required to obtain a good efficiency. A center peaked RF current profile can be formed with an appropriate wave spectrum even in the high temperature plasma. (author)

  17. Study of fast wave current drive in a KT-2 tokamak plasma

    International Nuclear Information System (INIS)

    Hong, B.G.; Hamamatsu, Kiyotaka

    1996-02-01

    Global analysis of fast wave current drive in a KT-2 tokamak plasma is performed by using the code, TASKW1, developed by JAERI and Okayama University (Dr. Fukuyama), which solves the kinetic wave equation in a one dimensional slab geometry. A phase-shifted antenna array is used to inject toroidal momentum to electrons. To find guidelines of optimum antenna design for efficient current drive, accessibility conditions are derived. The dependence of the current drive efficiency on launching conditions such as the total number of antennas, phase and spacing is investigated for two cases of wave frequency; f=30 MHz ( cH ) and f=225 MHz (=5f cH ). (author)

  18. Efficiency of LH current drive in tokamaks featuring an internal transport barrier

    International Nuclear Information System (INIS)

    Oliveira, C I de; Ziebell, L F; Rosa, P R da S

    2005-01-01

    In this paper, we study the effects of the occurrence of radial transport of particles in a tokamak on the efficiency of the current drive by lower hybrid (LH) waves, in the presence of an internal transport barrier. The results are obtained by numerical solution of the Fokker-Planck equation which rules the evolution of the electron distribution function. We assume that the radial transport of particles can be due to magnetic or electrostatic fluctuations. In both cases the efficiency of the current drive is shown to increase with the increase of the fluctuations that originate the transport. The dependence of the current drive efficiency on the depth and position of the barrier is also investigated

  19. Experiments on electron temperature profile resilience in FTU tokamak with continuous and modulated ECRH

    International Nuclear Information System (INIS)

    Cirant, S.

    2002-01-01

    Experiments performed on FTU tokamak, aiming at validation of physics-based transport models of the electron temperature profile resilience, are presented. ECRH is used to probe transport features, both in steady-state and in response to perturbations, while ECCD and LHCD are used for current density profile shaping. Observed confinement behaviour shows agreement with a critical temperature gradient length modelling. Central, low gradient plasma is characterized by low stiffness and low electron thermal diffusivity. Strong stiffness and high conduction are found in the confinement region. Resilience is experimentally characterized by an index of the resistance of the profile to adapt its shape to localized ECRH, while the diffusivity and its low-high transition are measured both by power balance and heat pulse propagation analysis. A particular attention is given to the investigation of the transition layer between low-high diffusivity and low-high stiffness regions. A dependence of LTc on magnetic shear, similar to what found in Tore Supra, and consistent with ETG based anomalous transport, is found. (author)

  20. Continuous, edge localized ion heating during non-solenoidal plasma startup and sustainment in a low aspect ratio tokamak

    Science.gov (United States)

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

    2017-07-01

    Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV  ⩽  650 eV, which is in contrast to T i,OV  ⩽  70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while {{T}\\text{i,\\parallel}} experiences little change, in agreement with two-fluid reconnection theory. This ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.

  1. Study of lower hybrid current drive system in tokamak fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Maebara, Sunao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 {mu}P is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 {mu}sec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10{sup -6} Pam{sup 3}/sm{sup 2} at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm{sup -2} (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10{sup -5} Pam{sup 3}/sm{sup 2} which is low enough for an antenna material. (author)

  2. 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)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Overview of steady-state tokamak operation and current drive experiments in TRIAM-1M

    International Nuclear Information System (INIS)

    Zushi, H.; Nakamura, K.; Hanada, K.

    2005-01-01

    Experiments aiming at 'day long operation at high performance' have been carried out. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation (SSTO) is studied under the localized PWI conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. Formation and sustainment of an internal transport barrier ITB in enhanced current drive mode (ECD) has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave ECW injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied. (author)

  5. Resistive effects on helicity-wave current drive generated by Alfven waves in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1997-01-01

    This work is concerned with the investigation of non-ideal (resistive) MHD effects on the excitation of Alfven waves by externally launched fast-mode waves, in simulated tokamak plasmas; both continuum range, CR ({ω Alf (r)} min Alf (r)} max ) and discrete range, DR, where global Alfven eigenmodes, GAEs (ω Alf (r)} min ) exist, are considered. (Here, ω Alf (r) ≡ ω Alf [n(r), B 0 (r)] is an eigenfrequency of the shear Alfven wave). For this, a cylindrical current carrying plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is used. Toroidicity effects are simulated by adopting for the axial equilibrium magnetic field component a suitable radial profile; shear and finite relative poloidal magnetic field are properly accounted for. A dielectric tensor appropriate to the physical conditions considered in this paper is derived and presented. (author)

  6. 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)

  7. Review of experiments on current drive in tokamaks by means of RF waves

    International Nuclear Information System (INIS)

    Hooke, W.

    1984-01-01

    Experimental results on lower hybrid current drive in tokamak plasmas are reviewed. Pulse lengths of 3.5 seconds and currents above 400 kA have been generated at plasma densities such that the wave frequency is greater than about twice the lower hybrid frequency. Current drive ceases above a critical density, nsub(c). However, nsub(c) increases with wave frequency. So that for f = 4.6 GHz current drive has been seen at n-barsub(e) approx.= 10 14 cm -3 and a density limit has yet to be established. Evidence for a collisional scaling law for current-drive efficiency is summarized. Detailed measurements of bremsstrahlung x-rays show a distribution which is qualitatively similar to that predicted by quasilinear theory. Microwave emission at frequencies less than the plasma frequency may shed light on the current-drive mechanism. Applications of current drive including plasma and current start-up and transformer recharging are discussed. (author)

  8. Evidence for Anomalous Effects on the Current Evolution in Tokamak Operating Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Casper, T; Jayakumar, R; Allen, S; Holcomb, C; Makowski, M; Pearlstein, L; Berk, H; Greenfield, C; Luce, T; Petty, C; Politzer, P; Wade, M; Murakami, M; Kessel, C

    2006-10-03

    Alternatives to the usual picture of advanced tokamak (AT) discharges are those that form when anomalous effects alter the plasma current and pressure profiles and those that achieve stationary characteristics through mechanisms so that a measure of desired AT features is maintained without external current-profile control. Regimes exhibiting these characteristics are those where the safety factor (q) evolves to a stationary profile with the on-axis and minimum q {approx} 1 and those with a deeply hollow current channel and high values of q. Operating scenarios with high fusion performance at low current and where the inductively driven current density achieves a stationary configuration with either small or non-existing sawteeth may enhance the neutron fluence per pulse on ITER and future burning plasmas. Hollow current profile discharges exhibit high confinement and a strong ''box-like'' internal transport barrier (ITB). We present results providing evidence for current profile formation and evolution exhibiting features consistent with anomalous effects or with self-organizing mechanisms. Determination of the underlying physical processes leading to these anomalous effects is important for scaling of current experiments for application in future burning plasmas.

  9. Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

    Science.gov (United States)

    Ren, Q.

    2015-11-01

    Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

  10. Development of high thermal flux components for continuous operation in Tokamaks

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Coston, J.F.; Deschamps, P.; Lipa, M.

    1991-01-01

    High heat flux plasma facing components are under development and appropriate experimental evaluations have been carried out in order to operate during cycles of several hundred seconds. In Tore Supra, a large tokamak with a plasma nominal duration in excess of 30 seconds, solutions are tested that could be later applied to the NET/ITER tokamak, where peaked heat flux values of 15 MW/m 2 on the divertor plates are foreseen. The proposed concept is a swirl square tube design protected with brazed CFC flat tiles. Development programs and validation tests are presented. The tests results are compared with calculations

  11. Traveling wave antenna for fast wave heating and current drive in tokamaks

    International Nuclear Information System (INIS)

    Ikezi, H.; Phelps, D.A.

    1995-07-01

    The traveling wave antenna for heating and current drive in the ion cyclotron range of frequencies is shown theoretically to have loading and wavenumber spectrum which are largely independent of plasma conditions. These characteristics have been demonstrated in low power experiments on the DIII-D tokamak, in which a standard four-strap antenna was converted to a traveling wave antenna through use of external coupling elements. The experiments indicate that the array maintains good impedance matching without dynamic tuning during abrupt changes in the plasma, such as during L- to H-mode transitions, edge localized mode activity, and disruptions. An analytic model was developed which exhibits the features observed in the experiments. Guidelines for the design of traveling wave antennas are derived from the validated model

  12. Edge fluctuations and global confinement with lower hybrid current drive in the ASDEX tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Stoeckel, J; Soeldner, F X; Giannone, L.; Leuterer, F; Steuer, K H [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Team

    1992-03-01

    Electrostatic edge fluctuations were investigated by means of Langmuir probes on the ASDEX tokamak in lower hybrid current drive regimes, simultaneously with the global particle and energy balances. It was found that the edge fluctuations are reduced and the global particle/energy confinement improves when the LH power is below the initial ohmic power. The maximum reduction of the fluctuations and the best confinement occur when the total power input (OH + LH) is minimum. With a LH power higher than the initial OH value, the fluctuation level increases noticeably, while no improvement of the global confinement is observed. The increase of the edge fluctuations seems to be poloidally localized and caused by local power deposition in front of the grill antenna. Therefore, the relative positions of the probe and antenna structure have to be taken account for correct interpretation of the fluctuation data. (orig.).

  13. Edge fluctuations and global confinement with lower hybrid current drive in the ASDEX tokamak

    International Nuclear Information System (INIS)

    Stoeckel, J.; Soeldner, F.X.; Giannone, L.; Leuterer, F.; Steuer, K.H.

    1992-03-01

    Electrostatic edge fluctuations were investigated by means of Langmuir probes on the ASDEX tokamak in lower hybrid current drive regimes, simultaneously with the global particle and energy balances. It was found that the edge fluctuations are reduced and the global particle/energy confinement improves when the LH power is below the initial ohmic power. The maximum reduction of the fluctuations and the best confinement occur when the total power input (OH + LH) is minimum. With a LH power higher than the initial OH value, the fluctuation level increases noticeably, while no improvement of the global confinement is observed. The increase of the edge fluctuations seems to be poloidally localized and caused by local power deposition in front of the grill antenna. Therefore, the relative positions of the probe and antenna structure have to be taken account for correct interpretation of the fluctuation data. (orig.)

  14. Internal magnetic probe measurements of MHD activity and current profiles in a tokamak

    International Nuclear Information System (INIS)

    Giannone, L.; Cross, R.C.

    1987-01-01

    Mirnov oscillations and plasma disruptions in the TORTUS tokamak have been studied by using both internal and external magnetic probe arrays. The internal probe was also used to measure the plasma current distribution so that results could be compared with resistive tearing mode calculations. The growth of m = 3, 4 and 5 modes was found to be consistent with linear tearing mode theory before a disruption but not after it. The observed mode amplitudes, typically b θ /B θ ∼ 5%, were much larger than theoretical estimates based on the magnetic energy available to drive the modes. Despite the presence of large islands near the limiter, most of the disruptions observed were associated with rapid growth of internal modes. (author). 23 refs, 14 figs

  15. Internal magnetic probe measurements of MHD activity and current profiles in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Giannone, L.; Cross, R C; Hutchinson, I H

    1987-01-01

    Mirnov oscillations and plasma disruptions in the TORTUS tokamak have been studied by using both internal and external magnetic probe arrays. The internal probe was also used to measure the plasma current distribution so that results could be compared with resistive tearing mode calculations. The growth of m = 3, 4 and 5 modes was found to be consistent with linear tearing mode theory before a disruption but not after it. The observed mode amplitudes, typically b/sub theta//B/sub theta/ approx. 5%, were much larger than theoretical estimates based on the magnetic energy available to drive the modes. Despite the presence of large islands near the limiter, most of the disruptions observed were associated with rapid growth of internal modes. (author). 23 refs, 14 figs.

  16. Statistical theory of wave propagation and multipass absorption for current drive in Tokamaks

    International Nuclear Information System (INIS)

    Moreau, D.; Litaudon, X.

    1993-07-01

    The effect of ray stochasticity on the multipass absorption of lower-hybrid waves, used to drive current in tokamaks, is considered. In toroidal geometry, stochasticity arises as an intrinsic property of the Hamiltonian ray trajectories for lower-hybrid waves. Based on the wave kinetic equation, a diffusion equation is derived, with damping and sources, for the wave energy density in the stochastic layer. This equation is solved simultaneously with the electron Fokker-Planck equation to describe the quasilinear flattening of the electron distribution function and the subsequent modification of the wave damping. It is shown that the spectral gap is filled in a self-regulating manner, so that the boundaries of the diffused wave spectrum are independent of the level of ray stochastic diffusion. A simple model for the self-consistent wave spectrum and the radial profile of absorbed power is proposed

  17. Traveling-wave antenna for fast-wave heating and current drive in tokamaks

    International Nuclear Information System (INIS)

    Ikezi, H.; Phelps, D.A.

    1997-01-01

    The travelling-wave antenna for heating and current drive in the ion cyclotron range of frequencies is shown theoretically to have loading and wavenumber spectra that are largely independent of plasma conditions. These characteristics have been demonstrated in low-power experiments on the DIII-D tokamak, in which a standard four-strap antenna was converted to a traveling-wave antenna through use of external coupling elements. The experiments indicate that the array maintains good impedance matching without dynamic tuning during abrupt changes in the plasma, such as during L- to H-mode transitions, edge-localized mode activity, and disruptions. An analytic model was developed that exhibits the features observed in the experiments. Guidelines for the design of travelling-wave antennas are derived from the validated model. 11 refs., 14 figs

  18. On the dynamics of the power spectrum during lower hybrid current drive in Tokamaks

    International Nuclear Information System (INIS)

    Bizarro, J.P.

    1993-01-01

    An investigation is provided on the propagation and absorption of the power spectrum during lower hybrid current drive in Tokamaks. A combined ray tracing and Fokker-Planck code is utilized and stochastic effects induced by toroidicity are correctly taken into account by using a large number of rays. It is shown that when strong wave damping prevails the absorbed spectrum is very similar in shape to the launched one, although some broadening and shifting in parallel wave index generally occur, and power deposition is localized. If the wave damping is weak and stochastic effects are important, rays end up sweeping the entire plasma cross-section, power deposition turns out to be extended, and the absorbed spectrum is much broader than the launched one

  19. Fast wave current drive in H mode plasmas on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Grassie, J.S. de; Baity, F.W.

    1999-01-01

    Current driven by fast Alfven waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma. (author)

  20. Fast wave and electron cyclotron current drive in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Pinsker, R.I.; Austin, M.E.

    1995-01-01

    The non-inductive current drive from directional fast Alfven and electron cyclotron waves was measured in the DIII-D tokamak in order to demonstrate these forms of radiofrequency (RF) current drive and to compare the measured efficiencies with theoretical expectations. The fast wave frequency was 8 times the deuterium cyclotron frequency at the plasma centre, while the electron cyclotron wave was at twice the electron cyclotron frequency. Complete non-inductive current drive was achieved using a combination of fast wave current drive (FWCD) and electron cyclotron current drive (ECCD) in discharges for which the total plasma current was inductively ramped down from 400 to 170 kA. For steady current discharges, an analysis of the loop voltage revealed up to 195 kA of a non-inductive current (out of 310 kA) during combined electron cyclotron and fast wave injection, with a maximum of 110 kA of FWCD and 80 kA of ECCD achieved (not simultaneously). The peakedness of the current profile increased with RF current drive, indicating that the driven current was centrally localized. The FWCD efficiency increased linearly with the central electron temperature as expected; however, the FWCD was severely degraded in low current discharges owing to incomplete fast wave absorption. The measured FWCD agreed with the predictions of a ray tracing code only when a parasitic loss of 4% per pass was included in the modelling along with multiple pass absorption. Enhancement of the second harmonic ECCD efficiency by the toroidal electric field was observed experimentally. The measured ECCD was in good agreement with Fokker-Planck code predictions. (author). 41 refs, 13 figs, 1 tab

  1. A simulation study on burning profile tailoring of steady state, high bootstrap current tokamaks

    International Nuclear Information System (INIS)

    Nakamura, Y.; Takei, N.; Tobita, K.; Sakamoto, Y.; Fujita, T.; Fukuyama, A.; Jardin, S.C.

    2007-01-01

    From the aspect of fusion burn control in steady state DEMO plant, the significant challenges are to maintain its high power burning state of ∝3-5 GW without burning instability, hitherto well-known as ''thermal stability'', and also to keep its desired burning profile relevant with internal transport barrier (ITB) that generates high bootstrap current. The paper presents a simulation modeling of the burning stability coupled with the self-ignited fusion burn and the structure-formation of the ITB. A self-consistent simulation, including a model for improved core energy confinement, has pointed out that in the high power fusion DEMO plant there is a close, nonlinear interplay between the fusion burnup and the current source of non-inductive, ITB-generated bootstrap current. Consequently, as much distinct from usual plasma controls under simulated burning conditions with lower power (<<1 GW), the selfignited fusion burn at a high power burning state of ∝3-5 GW becomes so strongly selforganized that any of external means except fuelling can not provide the effective control of the stable fusion burn.It is also demonstrated that externally applied, inductive current perturbations can be used to control both the location and strength of ITB in a fully noninductive tokamak discharge. We find that ITB structures formed with broad noninductive current sources such as LHCD are more readily controlled than those formed by localized sources such as ECCD. The physics of the inductive current is well known. Consequently, we believe that the controllability of the ITB is generic, and does not depend on the details of the transport model (as long as they can form an ITB for sufficiently reversed magnetic shear q-profile). Through this external control of the magnetic shear profile, we can maintain the ITB strength that is otherwise prone to deteriorate when the bootstrap current increases. These distinguishing capabilities of inductive current perturbation provide steady

  2. Design of the RF system for Alfven wave heating and current drive in a TCA/BR tokamak

    International Nuclear Information System (INIS)

    Ruchko, L.; Andrade, M.L.; Ozono, E.; Galvao, R.M.O.; Degaspari, F.T.; Nascimento, I.C.

    1995-01-01

    The advanced RF system for Alfven wave plasma heating and current drive in TCA/BR tokamak is presented. The antenna system is capable of exciting the standing and travelling wave M = -1,N = 1,N =-4,-6 with single helicity and thus provides the possibility to improve Alfven wave plasma heating efficiency in TCA/BR tokamak and to increase input power level up to P ≅ 1 MW, without the uncontrolled density rise which was encountered in previous TCA (Switzerland) experiments. (author). 4 refs., 3 figs

  3. Conversion of magnetic energy to runaway kinetic energy during the termination of runaway current on the J-TEXT tokamak

    Science.gov (United States)

    Dai, A. J.; Chen, Z. Y.; Huang, D. W.; Tong, R. H.; Zhang, J.; Wei, Y. N.; Ma, T. K.; Wang, X. L.; Yang, H. Y.; Gao, H. L.; Pan, Y.; the J-TEXT Team

    2018-05-01

    A large number of runaway electrons (REs) with energies as high as several tens of mega-electron volt (MeV) may be generated during disruptions on a large-scale tokamak. The kinetic energy carried by REs is eventually deposited on the plasma-facing components, causing damage and posing a threat on the operation of the tokamak. The remaining magnetic energy following a thermal quench is significant on a large-scale tokamak. The conversion of magnetic energy to runaway kinetic energy will increase the threat of runaway electrons on the first wall. The magnetic energy dissipated inside the vacuum vessel (VV) equals the decrease of initial magnetic energy inside the VV plus the magnetic energy flowing into the VV during a disruption. Based on the estimated magnetic energy, the evolution of magnetic-kinetic energy conversion are analyzed through three periods in disruptions with a runaway current plateau.

  4. Efficiency of the generation of impulsion by cyclotron waves currents of the electrons in an Axisymmetric Tokamak

    International Nuclear Information System (INIS)

    Gutierrez T, C.; Beltran P, M.

    2004-01-01

    The neoclassical theory of transport is used to calculate the current efficiency of electronic cyclotron impulsion (ECCD) in an axisymmetric tokamak in the few collisions regime. The standard parameter of the tokamak is used to obtain a system of equations that describe the hydrodynamic of the plasma, where the ponderomotive force (PM) due to high power radio frequency waves is taken in account. The PM force is produced in the proximity of electron cyclotron resonance surface in a specific poloidal localization. The efficiency ECCD is analyzed in the cases of first and second harmonic (for different angles of injection of radio frequency waves) and it is validated using the experimental values of the TCV and T-10 tokamaks. The results are according to those obtained by means of the techniques of the Green functions. (Author)

  5. Structure and parameters dependences of Alfven wave current drive generated in the low-field side of simulated spherical tokamaks

    International Nuclear Information System (INIS)

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

    1999-01-01

    Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)

  6. Structure and parameters dependences of Alfven wave current drive generated in the low-field side of simulated spherical tokamaks

    International Nuclear Information System (INIS)

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

    2001-01-01

    Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)

  7. Feedback control of current drive by using hybrid wave in tokamaks; Asservissement de la generation de courant par l`onde hybride dans un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wijnands, T.J. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Sciences de la Matiere

    1997-03-01

    This work is focussed on an important and recent development in present day Controlled Nuclear Fusion Research and Tokamaks. The aim is to optimise the energy confinement for a certain magnetic configuration by adapting the radial distribution of the current. Of particular interest are feedback control scenarios with stationary modifications of the current profile using current, driven by Lower Hybrid waves. A new feedback control system has been developed for Tore Supra and has made a large number of new operation scenarios possible. In one of the experiments described here, there is no energy exchange between the poloidal field system and the plasma, the current is controlled by the power of the Lower Hybrid waves while the launched wave spectrum is used to optimise the current profile shape and the energy confinement. (author) 151 refs.

  8. Inductive current startup in large tokamaks with expanding minor radius and rf assist

    International Nuclear Information System (INIS)

    Borowski, S.K.

    1984-02-01

    Auxiliary rf heating of electrons before and during the current-rise phase of a large tokamak, such as the Fusion Engineering Device (R = 4.8 m, a = 1.3 m, sigma = 1.6, B/sub T/ = 3.62 T), is examined as a means of reducing both the initiation loop voltage and resistive flux expenditure during startup. Prior to current initiation, 1 to 2 MW of electron cyclotron resonance heating power at approx. 90 GHz is used to create a small volume of high conductivity plasma (T/sub e/ approx. = 100 eV, n/sub e/ approx. = 10 19 m -3 ) near the upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a 0 approx. = 0.2 to 0.4 m) current channel to be established with a relatively low initial loop voltage (less than or equal to 25 V as opposed to approx. 100 V without rf assist). During the subsequent plasma expansion and current ramp phase, a combination of rf heating (up to 5 MW) and current profile control leads to a substantial savings in volt-seconds by: (1) minimizing the resistive flux consumption; and (2) maintaining the internal flux at or near the flat profile limit

  9. Stable equilibria for bootstrap-current-driven low aspect ratio tokamaks

    International Nuclear Information System (INIS)

    Miller, R.L.; Lin-Liu, Y.R.; Turnbull, A.D.; Chan, V.S.; Pearlstein, L.D.; Sauter, O.; Villard, L.

    1997-01-01

    Low aspect ratio tokamaks (LATs) can potentially provide a high ratio of plasma pressure to magnetic pressure β and high plasma current I at a modest size. This opens up the possibility of a high-power density compact fusion power plant. For the concept to be economically feasible, bootstrap current must be a major component of the plasma current, which requires operating at high β p . A high value of the Troyon factor β N and strong shaping is required to allow simultaneous operation at a high-β and high bootstrap fraction. Ideal magnetohydrodynamic stability of a range of equilibria at aspect ratio 1.4 is systematically explored by varying the pressure profile and shape. The pressure and current profiles are constrained in such a way as to assure complete bootstrap current alignment. Both β N and β are defined in terms of the vacuum toroidal field. Equilibria with β N ≥8 and β∼35%endash 55% exist that are stable to n=∞ ballooning modes. The highest β case is shown to be stable to n=0,1,2,3 kink modes with a conducting wall. copyright 1997 American Institute of Physics

  10. 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

  11. Enhancing current density profile control in tokamak experiments using iterative learning control

    NARCIS (Netherlands)

    Felici, F.A.A.; Oomen, T.A.E.

    2015-01-01

    Tokamaks are toroidal devices to create and confine high-temperature plasmas, and are presently at the forefront of nuclear fusion research. Many parameters in a tokamak are feedback controlled, but some quantities that are either difficult to measure or difficult to control are still controlled by

  12. Prevention of the current-quench phase of a major disruption in a tokamak reactor

    International Nuclear Information System (INIS)

    Miller, J.B.

    1987-01-01

    The 2-D Tokamak Simulation Code written by the Princeton Plasma Physics Laboratory was joined to a 3-D eddy-current code, which models periodic torus sectors. The combined system was found to be an efficient and accurate method for modeling the plasma/eddy current interaction during a major disruption. For modeling large highly compartmentalized structures, artificially increasing the self-inductance and limiting the mutual inductance of current elements were necessary to enhance numerical stability. Even with these modifications, a slowly growing instability made the results unreliable after 58 ms. This model was used to demonstrate prevention of the current quench phase of a major disruption in INTOR. The average plasma temperature was reduced to 150 eV over 3 ms. The (outboard) breeding blanket structure was constructed of CuBeNi and was electrically connected between torus sectors. Disruption recovery coils were provided inboard of the inboard shield (linking the toroidal field coils). It was necessary to supply to these coils a total of 500 MW for 0.6 s and to reheat the plasma to full beta in 6 s. The calculation shows a method of recovery from the most severe disruption probable. Determining the severity of the disruption from which recovery would be cost effective is beyond the scope of this study

  13. Preliminary oscillating fluxes current drive experiment in DIII-D tokamak

    International Nuclear Information System (INIS)

    Yamaguchi, S.; Schaffer, M.; Kondoh, Y.

    1995-01-01

    A preliminary oscillating flux helicity injection experiment was done on DIII-D tokamak. The toroidal flux was modulated by programming the plasma elongation. Instead of programming the surface voltage directly, the plasma current was programmed with a periodic modulation at some phase shift. The theoretical basis of this modulation is discussed in terms of the helicity injection and also introduced by cross-field motion of the modulated plasma. Because the primary winding is well coupled with the plasma current and the power supply is strong, the plasma current behaves as programmed. However, as the plasma shape is not coupled strongly with the shaping and equilibrium coils, the elongation amplitude and phase are affected by the change of plasma current and do not behave as programmed. Because of this, the voltage induced by the helicity injection is low, and the experiment did not test the principle of helicity injection. The injection powers of helicity and energy, and the electric field intensity of the helicity injection model and the cross-field motion of plasma are compared with each other experimentally. The improvement necessary to do the experiment is also proposed. ((orig.))

  14. Inductive current startup in large tokamaks with expanding minor radius and RF assist

    International Nuclear Information System (INIS)

    Borowski, S.K.

    1983-01-01

    Auxiliary RF heating of electrons before and during the current rise phase of a large tokamak, such as the Fusion Engineering Device, is examined as a means of reducing both the initiation loop voltage and resistive flux expenditure during startup. Prior to current initiation, 1 to 2 MW of electron cyclotron resonance heating power at approx.90 GHz is used to create a small volume of high conductivity plasma (T/sub e/ approx. = 100 eV, n/sub e/ approx. = 10 19 m -3 ) near the upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a 0 approx.< 0.4 m) current channel to be established with a relatively low initial loop voltage (approx.< 25 V as opposed to approx.100 V without RF assist). During the subsequent plasma expansion and current ramp phase, additional RF power is introduced to reduce volt-second consumption due to plasma resistance. To study the preheating phase, a near classical particle and energy transport model is developed to estimate the electron heating efficiency in a currentless toroidal plasma. The model assumes that preferential electron heating at the UHR leads to the formation of an ambipolar sheath potential between the neutral plasma and the conducting vacuum vessel and limiter

  15. Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

    Science.gov (United States)

    Pace, D. C.; Austin, M. E.; Bardoczi, L.; Collins, C. S.; Crowley, B.; Davis, E.; Du, X.; Ferron, J.; Grierson, B. A.; Heidbrink, W. W.; Holcomb, C. T.; McKee, G. R.; Pawley, C.; Petty, C. C.; Podestà, M.; Rauch, J.; Scoville, J. T.; Spong, D. A.; Thome, K. E.; Van Zeeland, M. A.; Varela, J.; Victor, B.

    2018-05-01

    An engineering upgrade to the neutral beam system at the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic ( E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2 MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities and results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.

  16. Monitoring of the current profile by using cyclotronic electron waves in tokamaks; Controle du profil de courant par ondes cyclotroniques electroniques dans les tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Dumont, R

    2001-08-01

    The subject of this thesis is the study of the cyclotronic electron wave as a monitoring tool of the current profile. The first chapter is dedicated to basic notions concerning tokamak plasmas and current generation. The second chapter is centered on the use of fast electrons to generate current and on its modelling. The propagation and absorption of the cyclotronic electron wave require a specific polarization state whose characteristics must be carefully chosen according to some parameters of the discharge, the chapter 3 deals with this topic. The absorption of a wave in a plasma depends greatly on the velocity distribution of the particles that make up the plasma and this distribution is constantly modified by the energy of the wave, so this phenomenon is non-linear and its physical description is difficult. In a case of a fusion plasma, a sophisticated approximation called quasi-linear theory can be applied with some restrictions that are presented in chapter 4. Chapters 5 and 6 are dedicated to kinetics scenarios involving the low hybrid wave and the cyclotronic electron wave inside the plasma. Some experiments dedicated to the study of the cyclotronic electron wave have been performed in Tore-supra (France) and FTU (Italy) tokamaks, they are presented in the last chapter. (A.C.)

  17. On the evaluation of currents in a tokamak plasma during combined Ohmic and RF current drive

    International Nuclear Information System (INIS)

    Eckhartt, D.

    1986-09-01

    By taking into account the rf-generated enhancement of the plasma electric conductivity (as formulated by Fisch in the limit of weak dc electric fields) a relation is derived between the ratio of rf to Ohmically driven currents and other plasma parameters to be measured before and after the rf onset under the condition of constant net plasma current. (author)

  18. Advanced antenna system for Alfven wave plasma heating and current drive in TCABR tokamak

    International Nuclear Information System (INIS)

    Ruchko, L.F.; Ozono, E.; Galvao, R.M.O.; Nascimento, I.C.; Degasperi, F.T.; Lerche, E.

    1998-01-01

    An advanced antenna system that has been developed for investigation of Alfven wave plasma heating and current drive in the TCABR tokamak is described. The main goal was the development of such a system that could insure the excitation of travelling single helicity modes with predefined wave mode numbers M and N. The system consists of four similar modules with poloidal windings. The required spatial spectrum is formed by proper phasing of the RF feeding currents. The impedance matching of the antenna with the four-phase oscillator is accomplished by resonant circuits which form one assembly unit with the RF feeders. The characteristics of the antenna system design with respect to the antenna-plasma coupling and plasma wave excitation, for different phasing of the feeding currents, are summarised. The antenna complex impedance Z=Z R +Z I is calculated taking into account both the plasma response to resonant excitation of fast Alfven waves and the nonresonant excitation of vacuum magnetic fields in conducting shell. The matching of the RF generator with the antenna system during plasma heating is simulated numerically, modelling the plasma response with mutually coupled effective inductances with corresponding active Z R and reactive Z I impedances. The results of the numerical simulation of the RF system performance, including both the RF magnetic field spectrum analysis and the modeling of the RF generator operation with plasma load, are presented. (orig.)

  19. Confinement and transport properties during current ramps in the ASDEX Upgrade tokamak

    Science.gov (United States)

    Fable, E.; Angioni, C.; Hobirk, J.; Pereverzev, G.; Fietz, S.; Hein, T.; ASDEX Upgrade Team

    2011-04-01

    A detailed analysis of experimental data from the ASDEX Upgrade tokamak is carried out to shed light on the properties of confinement and transport in the current ramp-up and ramp-down phases of the plasma discharge. The experimental database is used to identify the relevant ranges of parameters explored during the ramp-up and the ramp-down. The energy confinement time observed in the two ramps displays interesting evolution, in many cases attaining different values at the same current level between ramp-up and ramp-down. The possible reasons for this behaviour are investigated. Interpretative transport simulations are used as a tool to clarify the interplay between different parameters, which are coupled in a non-linear way. In addition, a theory-based transport model is used to understand the behaviour of confinement as observed in the experiment, evidencing the role of both turbulent and neoclassical transport. Linear gyrokinetic calculations are performed to identify the relevant turbulence regime, showing that a broad range of frequencies, in the trapped electron modes (TEMs) and in the ion temperature gradient modes (ITGs) regimes, is explored during both the ramp-up and ramp-down. In the same framework, a quasi-linear model is applied to calculate the value of the local logarithmic density gradient and compare it with the experimental value. Finally, first non-linear simulations of heat transport during the current ramps are presented.

  20. Efficient ion heating of tokamak plasma by application of positive and negative current pulse in TRIAM-1

    International Nuclear Information System (INIS)

    Toi, Kazuo; Hiraki, Naoji; Nakamura, Kazuo; Mitarai, Osamu; Kawai, Yoshinobu

    1980-01-01

    The efficient heating of bulk ions of tokamak plasma is observed by application of the pulsed toroidal electric field much higher than the Dreicer field with the positive and negative polarities for the ohmic heating field. No deleterious effect on the confinement properties of tokamak plasma appears by the heating. The decay time of ion temperature raised by the heating pulse agrees well with the prediction by the neoclassical transport theory. The magnitude of the current induced by the pulsed electric field with the positive polarity is limited by the violent current disruption. In the case of the negative polarity, this is limited by lack of the MHD equilibrium due to vanishing the total plasma current. The ratio of drift velocity to electron thermal one / attains around 0.5, which suggests that the efficient ion heating may be due to the current-driven turbulence. (author)

  1. Efficient ion heating of tokamak plasma by application of positive and negative current pulse in TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-02-01

    The efficient heating of bulk ions of tokamak plasma is observed by application of the pulsed toroidal electric field much higher than the Dreicer field with the positive and negative polarities for the ohmic heating field. No deleterious effect on the confinement properties of tokamak plasma appears by the heating. The decay time of ion temperature raised by the heating pulse agrees well with the prediction by the neoclassical transport theory. The magnitude of the current induced by the pulsed electric field with the positive polarity is limited by the violent current disruption. In the case of the negative polarity, this is limited by lack of the MHD equilibrium due to vanishing the total plasma current. The ratio of drift velocity to electron thermal one / attains around 0.5, which suggests that the efficient ion heating may be due to the current-driven turbulence.

  2. Moment approach to neoclassical flows, currents and transport in auxiliary heated tokamaks

    International Nuclear Information System (INIS)

    Kim, Yil Bong.

    1988-02-01

    The moment approach is utilized to derive the full complement of neoclassical transport processes in auxiliary heated tokamaks. The effects of auxiliary heating [neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH)] considered arise from the collisional interaction between the background plasma species and the fast-ion-tail species. From a known fast ion distribution function we evaluate the parallel (to the magnetic field) momentum and heat flow inputs to the background plasma. Then, through the momentum and heat flow balance equations, we can determine the induced parallel flows (and current) and radial transpot fluxes in ''equilibrium'' (on the time scale much longer than the collisional relaxation time, i.e., t >> 1ν/sub ii/). In addition to the fast-ion-induced current, the total neoclassical current includes the boostap current, which is driven by the pressure and temperature gradients, the Pfirsch-Schlueter current which is required for charge neutrality, and the neoclassical (including trapped particle effects) Spitzer current due to the parallel electric field. The radial transport fluxes also include off-diagonal compnents in the transport matrix which correspond to the Ware (neoclassical) pinch due to the inductive applied electric field an the fast-ion-induced radial fluxes, in addition to the usual pressure- and temperature-gradient-driven fluxes (particle diffusion and heat conduction). Once the tranport coefficient are completely determined, the radial fluxes and the heat fluxes can be substituted into the density and energy evolution equations to provide a complete description of ''equilibrium'' (δδt << ν/sub ii/) neoclassical transport processes in a plasma. 47 refs., 14 figs

  3. Test of the quasi-optical grill for lower hybrid current drive on the CASTOR tokamak

    International Nuclear Information System (INIS)

    Klima, R.; Pavlo, P.; Preinhaelter, J.; Stoeckel, J.; Zacek, F.; Jakubka, K.; Kletecka, P.; Kryska, L.

    1994-03-01

    Feasibility studies of a new diffraction structure for launching lower hybrid waves into a tokamak plasma - of the microwave quasi-optical grill - are reported. The main parameters of the grill designed for the CASTOR tokamak are summarized, and results of preliminary radiation pattern measurements of a non-optimized model antenna are presented. The influence of a relatively great curvature of the plasma surface in the CASTOR tokamak is discussed and the ray tracing of the launched lower-hybrid wave in the actual CASTOR plasma is shown. Finally, the results of probe measurements of the CASTOR plasma core are given. (J.U.) 17 figs., 9 refs

  4. Non-inductive current drive via helicity injection by Alfven waves in low-aspect-ratio tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cuperman, S.; Bruma, C.; Komoshvili, K. [Tel Aviv Univ. (Israel). Sackler Faculty of Exact Sciences

    1996-08-01

    A theoretical investigation of radio-frequency (RF) current drive via helicity injection in low aspect ratio tokamaks is carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is considered. Toroidal features of low-aspect-ratio tokamaks are simulated by incorporating the following effects: (i) arbitrarily small aspect ratio, R{sub O}/a ``identical to`` 1/{epsilon}; (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. This study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub min}{r_brace} {<=} {omega} {<=} {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub max}, where {omega}{sub Alf}({tau}) ``identical to`` {omega}{sub Alf}[n({tau}), B{sub O}({tau})] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-{beta} magnetohydrodynamics, the wave equation with correct boundary (matching) conditions is solved, the RF field components are found, and subsequently current drive, power deposition and efficiency are computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low-aspect-ratio tokamaks, in the SAW mode. A special algorithm is developed that enables one to select the antenna parameters providing optimal current drive efficiency. (Author).

  5. Resistive effects on helicity-wave current drive generated by Alfven waves in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Bruma, C.; Cuperman, S.; Komoshvili, K. [Tel Aviv Univ. (Israel). Faculty of Exact Sciences

    1997-05-01

    This work is concerned with the investigation of non-ideal (resistive) MHD effects on the excitation of Alfven waves by externally launched fast-mode waves, in simulated tokamak plasmas; both continuum range, CR ({l_brace}{omega}{sub Alf}(r){r_brace}{sub min} < {omega} < {l_brace}{omega}{sub Alf}(r){r_brace}{sub max}) and discrete range, DR, where global Alfven eigenmodes, GAEs ({omega} < {l_brace}{sub Alf}(r){r_brace}{sub min}) exist, are considered. (Here, {omega}{sub Alf}(r) {identical_to} {omega}{sub Alf}[n(r), B{sub 0}(r)] is an eigenfrequency of the shear Alfven wave). For this, a cylindrical current carrying plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is used. Toroidicity effects are simulated by adopting for the axial equilibrium magnetic field component a suitable radial profile; shear and finite relative poloidal magnetic field are properly accounted for. A dielectric tensor appropriate to the physical conditions considered in this paper is derived and presented. (author).

  6. Study of matrix converter as a current-controlled power supply in QUEST tokamak

    International Nuclear Information System (INIS)

    Liu, Xiaolong; Jiang, Yi; Nakamura, Kazuo

    2011-01-01

    Because QUEST tokamak has a divertor configuration with a higher κ and a negative n-index, a precise power supply with a rapid response is needed to control the vertical position of the plasma. A matrix converter is a direct power conversion device that uses an array of controlled bidirectional switches as the main power elements for creating a variable-output current system. This paper presents a novel three-phase to two-phase topological matrix converter as a proposed power supply that stabilizes the plasma vertical position and achieves unity input power factor. An indirect control strategy in which the matrix converter is split into a virtual rectifier stage and a virtual inverter stage is adopted. In the virtual rectifier stage, the instantaneous active power and reactive power are decoupled on the basis of system equations derived from the DQ transformation; hence, unity power factor is achieved. Space vector pulse width modulation is adopted to determine the switching time of each switch in the virtual rectifier; the output voltage of the virtual rectifier is adjusted by the virtual inverter stage to obtain the desired load current. Theoretical analyses and simulation results are provided to verify its feasibility. (author)

  7. Continuous and real-time data acquisition system for superconducting tokamaks HT-7 and TRIAM-1M

    International Nuclear Information System (INIS)

    Wang, F.; Luo, J.R.; Nakamura, K.; Sato, K.N.; Hanada, K.; Sakamoto, M.; Idei, H.; Kawasaki, S.; Nakashima, H.

    2006-01-01

    Conventional data acquisition systems cannot deal with data acquisition for a long-time discharge of a nuclear fusion reactor. Thus, continuous data acquisition with a real-time data presentation during discharge must be developed. Two data acquisition systems, which include alternating CAMAC data acquisition and long-time PCI data acquisition, are designed for the long-time operation of HT-7 tokamak. Since an effective alternating mode is adopted, the alternating CAMAC data acquisition can accurately and continuously acquire data at a rate of 10 kHz. The acquired data is immediately transmitted to a data server and real-time results can be presented during the plasma discharge. As for the long-time PCI data acquisition, a special kind of PCI A/D card, which has a hard disk on board, is designed to collect data at a max speed of 200 kHz. Thus, the total sampling duration is only related to the capacity of the hard disk on board. These two types of data acquisitions were applied to HT-7 tokamak and a 250 s discharge was acquired. These data acquisition systems were also successfully demonstrated on a 2500 s plasma discharge on TRIAM-1M. This paper describes the two data acquisitions in detail

  8. Modeling and control of the current density profile in Tokamaks and its relation to electron transport

    International Nuclear Information System (INIS)

    Zucca, C.

    2009-04-01

    The current density in tokamak plasmas strongly affects transport phenomena, therefore its understanding and control represent a crucial challenge for controlled thermonuclear fusion. Within the vast framework of tokamak studies, three topics have been tackled in the course of the present thesis: first, the modelling of the current density evolution in electron Internal Transport Barrier (eITB) discharges in the Tokamak à Configuration Variable (TCV); second, the study of current diffusion and inversion of electron transport properties observed during Swing Electron Cyclotron Current Drive (Swing ECCD) discharges in TCV; third, the analysis of the current density tailoring obtained by local ECCD driven by the improved EC system for sawtooth control and reverse shear scenarios in the International Thermonuclear Experimental Reactor (ITER). The work dedicated to the study of eITBs in TCV has been undertaken to identify which of the main parameters, directly related to the current density, played a relevant role in the confinement improvement created during these advanced scenarios. In this context, the current density has to be modeled, there being no measurement currently available on TCV. Since the Rebut-Lallia-Watkins (RLW) model has been validated on TCV ohmic heated plasmas, the corresponding scaling factor has often been used as a measure of improved confinement on TCV. The many interpretative simulations carried on different TCV discharges have shown that the thermal confinement improvement factor, H RLW , linearly increases with the absolute value of the minimum shear outside ρ > 0.3, ρ indicating a normalized radial coordinate. These investigations, performed with the transport code ASTRA, therefore confirmed a general observation, formulated through previous studies, that the formation of the transport barrier is correlated with the magnetic shear reversal. This was, indeed, found to be true in all cases studied, regardless of the different heating and

  9. MHD stability analysis of axisymmetric surface current model tokamaks close to the spheromak regime

    International Nuclear Information System (INIS)

    Honma, Toshihisa; Kaji, Ikuo; Fukai, Ichiro; Kito, Masafumi.

    1984-01-01

    In the toroidal coordinates, a stability analysis is presented for very low-aspect-ratio tokamaks with circular cross section which is described by a surface current model (SCM) of axisymmetric equilibria. The energy principle determining the stability of plasma is treated without any expansion of aspect ratio. Numerical results show that, owing to the occurrence of the non-axisymmetric (n=1) unstable modes, there exists no MHD-stable ideal SCM spheromak characterized by zero external toroidal vacuum field. Instead, a stable spheromak-type plasma which comes to the ideal SCM spheromak is provided by the configuration with a very weak external toroidal field. Close to the spheromak regime (1.0 1 aspect ratio< = 1.1), the minimum safety factor and the critical β-values increase mo notonically with aspect ratio decreasing from a large value, and curves of βsub(p) versus β in the marginal stability approach to an ideal SCM spheromak line βsub(p)=β. (author)

  10. Basic toroidal Effects on Alfven Wave Current in Small Aspect Ratio Tokamaks

    International Nuclear Information System (INIS)

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

    1998-01-01

    The Alfven wave current drive (AWCD) in small aspect ratio Tokamaks is properly calculated, with consideration of the basic toroidicity effects present in (i) the dielectric tensor-operator (involving the strongly toroidal equilibrium profiles), (ii) the structure of the r.f. fields obtained as a solution of the wave equation (through Maxwell's equations' toroidal operators as well as the conversion rate and conversion layer location, depending also on the equilibrium profiles) and (iii) the formulation of the AWCD (which, besides its dependence on the r.f. fields - affected by toroidicity as mentioned at points (i) and (ii) - also requires the equilibrium-magnetic-surface averaging of non-resonant forces involved). Thus, we consider consistent equilibrium profiles with neo-classical conductivity corresponding to an ohmic START-like discharge; use a resistive (anisotropic) MHD dielectric tensor-operator Edith practically no limitations, adequate to describe the plasma response in the pre-heated stage ; solve numerically the 2(1/2)D full- wave equation by the aid of an advanced finite element code developed in; and evaluate the AWCD by the aid of the recently proposed, quite general formulation holding in the case of strongly toroidal fusion devices and including contributions due to helicity injection, momentum transfer and plasma Bow. A general discussion of the results obtained in this work is presented

  11. Very fast feedback control of coil-current in JT-60 tokamak

    International Nuclear Information System (INIS)

    Aoyagi, T.; Terakado, T.; Takahashi, M.; Nobusaka, H.; Yagyu, J.; Matsuzaki, Y.

    1992-01-01

    A direct digital control (DDC) system is adopted for controlling thyristor converters of power supplies in the JT-60 tokamak built in 1984. Microcomputers of the DDC were 5 MHz i8086 microprocessor and programs were written by assembler language and the processing time was under 1ms. They were, however, too old in hardware and too complicated in software. New DDC system has been made in the JT-60 Upgrade (JT-60U) to control the power supplies more quickly under 0.25 and 0.5 ms of the processing time and also to write the programs used by high-level language. The new system consists of a host computer and five microcomputers with microprocessor on VME bus system. The host computer AS3260 performs on-line processing such as setting the DDC under the discharge conditions and so on. Functions of the microcomputers with a 32-bit, 20 MHz microprocessor MC68030, whose OS are VxWorks and programs are written by C language, are real-time processing such as taking in instructions from a ZENKEI computer and in feedback control of currents and voltages of coils every 0.25 and 0.5 ms. The system is now operating very smoothly. (author)

  12. Behavior of hard X-ray emission in discharges with current disruptions in the DAMAVAND and TVD tokamaks

    International Nuclear Information System (INIS)

    Farshi, E.; Amrollahy, R.; Bortnikov, A.V.; Brevnov, N.N.; Gott, Yu.V.; Shurygin, V.A.

    2001-01-01

    Results are presented from studies of the behavior of hard X-ray emission in discharges with current disruptions in the DAMAVAND and TVD tokamaks. The current disruptions are caused by either an MHD instability or the instability related to the vertical displacement of the plasma column. Experiments were conducted at a fixed value of the safety factor at the plasma boundary (q a ≅ 2.3). Experimental data show that, during a disruption caused by an MHD instability, hard X-ray emission is suppressed by this instability if the amplitude of the magnetic field fluctuations exceeds a certain level. If the disruption is caused by the instability related to the vertical displacement of the plasma column, then hard X-ray emission is observed at the instant of disruption. The experimental results show that the physical processes resulting in the generation and suppression of runaway electron beams are almost identical in large and small tokamaks

  13. Effect of eddy currents in the toroidal field coils of a tokamak with an air-core transformer

    International Nuclear Information System (INIS)

    Tani, Keiji; Kobayashi, Tomofumi; Tamura, Sanae

    1975-02-01

    The effect of eddy currents in the copper parts of the toroidal field coils is evaluated for a tokamak with the air-core transformer windings located inside the bore of the toroidal field coils. By introducing appropriate weights to the solutions obtained for a simplified cylindrical model, calculation is made of the induction toroidal electric field on the plasma axis in the presence of the eddy currents. The result shows that, to reduce the influence of the eddy currents on the induction one-turn voltage to the permissible level, it is necessary to choose the optimal number of turns and shape of the single conductor of the toroidal field coil. (auth.)

  14. Resistivity effects in non-inductive RF current drive via helicity injection by Alfven waves: the case of conventional and small aspect ratio Tokamaks

    International Nuclear Information System (INIS)

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

    1996-01-01

    Supplementary non-inductive current drive and heating are necessary to bring Tokamak plasmas into the ignition regime. The resonant excitation of shear Alfven waves (SAW) - in the continuum range (CR) or/and in the discrete global Alfven eigenmode spectrum (GAE's) - represents one potential, suitable method for this purpose. Within the framework of ideal MHD, the current drive (CD) via helicity injection in Tokamak plasmas has been considered by Cuperman et al (1996) and Komoshvili et al. (1996). This work is concerned with the investigation of the non-ideal resistive MHD effects on both the excitation of SAW's (CR's and GAE's) and the generation of non-inductive current drive via helicity injection in Tokamak plasmas. The research covers Tokamak aspect ratios ranging between large value cases (R/a = 10) and the very tight value case (R/ a = 1.2). (authors)

  15. 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

  16. Real Time Hybrid Model Predictive Control for the Current Profile of the Tokamak à Configuration Variable (TCV

    Directory of Open Access Journals (Sweden)

    Izaskun Garrido

    2016-08-01

    Full Text Available Plasma stability is one of the obstacles in the path to the successful operation of fusion devices. Numerical control-oriented codes as it is the case of the widely accepted RZIp may be used within Tokamak simulations. The novelty of this article relies in the hierarchical development of a dynamic control loop. It is based on a current profile Model Predictive Control (MPC algorithm within a multiloop structure, where a MPC is developed at each step so as to improve the Proportional Integral Derivative (PID global scheme. The inner control loop is composed of a PID-based controller that acts over the Multiple Input Multiple Output (MIMO system resulting from the RZIp plasma model of the Tokamak à Configuration Variable (TCV. The coefficients of this PID controller are initially tuned using an eigenmode reduction over the passive structure model. The control action corresponding to the state of interest is then optimized in the outer MPC loop. For the sake of comparison, both the traditionally used PID global controller as well as the multiloop enhanced MPC are applied to the same TCV shot. The results show that the proposed control algorithm presents a superior performance over the conventional PID algorithm in terms of convergence. Furthermore, this enhanced MPC algorithm contributes to extend the discharge length and to overcome the limited power availability restrictions that hinder the performance of advanced tokamaks.

  17. Analysis of Electron Thermal Diffusivity and Bootstrap Current in Ohmically Heated Discharges after Boronization in the HT-7 Tokamak

    International Nuclear Information System (INIS)

    Zhang, X.M.; Wan, B.N.

    2005-01-01

    Significant improvements of plasma performance after ICRF boronization have been achieved in the full range of HT-7 operation parameters. Electron power balance is analyzed in the steady state ohmic discharges of the HT-7 tokamak. The ratio of the total radiation power to ohmic input power increases with increasing the central line-averaged electron density, but decreases with plasma current. It is obviously decreased after wall conditioning. Electron heat diffusivity χ e deduced from the power balance analysis is reduced throughout the main plasma after boronization. χ e decreases with increasing central line-averaged electron density in the parameter range of our study. After boronization, the plasma current profile is broadened and a higher current can be easily obtained on the HT-7 tokamak experiment. It is expected that the fact that the bootstrap current increases after boronization will explain these phenomena. After boronization, the plasma pressure gradient and the electron temperature near the boundary are larger than before, these factors influencing that the ratio of bootstrap current to total plasma current increases from several percent to above 10%

  18. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  19. Axisymmetric MHD simulation of ITB crash and following disruption dynamics of Tokamak plasmas with high bootstrap current

    International Nuclear Information System (INIS)

    Takei, Nahoko; Tsutsui, Hiroaki; Tsuji-Iio, Shunji; Shimada, Ryuichi; Nakamura, Yukiharu; Kawano, Yasunori; Ozeki, Takahisa; Tobita, Kenji; Sugihara, Masayoshi

    2004-01-01

    Axisymmetric MHD simulation using the Tokamak Simulation Code demonstrated detailed disruption dynamics triggered by a crash of internal transport barrier in high bootstrap current, high β, reversed shear plasmas. Self-consistent time-evolutions of ohmic current bootstrap current and induced loop voltage profiles inside the disrupting plasma were shown from a view point of disruption characterization and mitigation. In contrast with positive shear plasmas, a particular feature of high bootstrap current reversed shear plasma disruption was computed to be a significant change of plasma current profile, which is normally caused due to resistive diffusion of the electric field induced by the crash of internal transport barrier in a region wider than the internal transport barrier. Discussion based on the simulation results was made on the fastest record of the plasma current quench observed in JT-60U reversed shear plasma disruptions. (author)

  20. Magnetic analysis including the field due to vacuum vessel eddy currents in the Hitachi Tokamak (HT-2)

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Takeuchi, Kazuhiro; Fukumoto, Hideshi; Otsuka, Michio

    1989-01-01

    A magnetic analysis to determine plasma surface position is applied to the magnetic data of the Hitachi Tokamak (HT-2). The analysis takes account of toroidal eddy currents on the vacuum vessel wall. Magnetic probes in HT-2 are placed on both sides of the wall (plasma side and outside), making it possible to determine magnitudes of eddy currents which flow in the toroidal direction. The magnitudes of the coil currents and eddy currents are determined so as to reproduce the measured magnetic fields, and to reconstruct flux surfaces and plasma surface are reconstructed. Taking into account the eddy currents, the determination errors of the plasma surface position are reduced by up to 1/2.3 during start-up and terminating phases, compared with the case without eddy currents. (author)

  1. A current drive by using the fast wave in frequency range higher than two timeslower hybrid resonance frequency on tokamaks

    Directory of Open Access Journals (Sweden)

    Kim Sun Ho

    2017-01-01

    Full Text Available An efficient current drive scheme in central or off-axis region is required for the steady state operation of tokamak fusion reactors. The current drive by using the fast wave in frequency range higher than two times lower hybrid resonance (w>2wlh could be such a scheme in high density, high temperature reactor-grade tokamak plasmas. First, it has relatively higher parallel electric field to the magnetic field favorable to the current generation, compared to fast waves in other frequency range. Second, it can deeply penetrate into high density plasmas compared to the slow wave in the same frequency range. Third, parasitic coupling to the slow wave can contribute also to the current drive avoiding parametric instability, thermal mode conversion and ion heating occured in the frequency range w<2wlh. In this study, the propagation boundary, accessibility, and the energy flow of the fast wave are given via cold dispersion relation and group velocity. The power absorption and current drive efficiency are discussed qualitatively through the hot dispersion relation and the polarization. Finally, those characteristics are confirmed with ray tracing code GENRAY for the KSTAR plasmas.

  2. Dependence of CIT [Compact Ignition Tokamak] PF [poloidal field] coil currents on profile and shape parameters using the Control Matrix

    International Nuclear Information System (INIS)

    Strickler, D.J.; Peng, Y-K.M.; Jardin, S.C.; Pomphrey, N.

    1990-01-01

    The plasma shaping flexibility of the Compact Ignition Tokamak (CIT) poloidal field (PF) coil set is demonstrated through MHD equilibrium calculations of optimal PF coil current distributions and their variation with poloidal beta, internal inductance, plasma 95% elongation, and 95% triangularity. Calculations of the magnetic stored energy are used to compare solutions associated with various plasma parameters. The Control Matrix (CM) equilibrium code, together with the nonlinear equation and numerical optimization software packages HYBRD, and VMCON, respectively, are used to find equilibrium coil current distributions for fixed divertor geometry, volt-seconds, and plasma profiles in order to isolate the dependence on individual parameters. A reference equilibrium and coil current distribution are chosen, and correction currents dI are determined using the CM equilibrium method to obtain other specified plasma shapes. The reference equilibrium is the κ = 2 divertor at beginning of flattop (BOFT) with a minimum stored energy solution for the coil current distribution. The pressure profile function is fixed

  3. Efficiency of the generation of impulsion by cyclotron waves currents of the electrons in an Axisymmetric Tokamak; Eficiencia de la generacion de corrientes de impulsion por ondas ciclotronicas de los electrones en un Tokamak axisimetrico

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez T, C.; Beltran P, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2004-07-01

    The neoclassical theory of transport is used to calculate the current efficiency of electronic cyclotron impulsion (ECCD) in an axisymmetric tokamak in the few collisions regime. The standard parameter of the tokamak is used to obtain a system of equations that describe the hydrodynamic of the plasma, where the ponderomotive force (PM) due to high power radio frequency waves is taken in account. The PM force is produced in the proximity of electron cyclotron resonance surface in a specific poloidal localization. The efficiency ECCD is analyzed in the cases of first and second harmonic (for different angles of injection of radio frequency waves) and it is validated using the experimental values of the TCV and T-10 tokamaks. The results are according to those obtained by means of the techniques of the Green functions. (Author)

  4. First time observation of local current shrinkage during the MARFE behavior on the J-TEXT tokamak

    Science.gov (United States)

    Shi, Peng; Zhuang, G.; Gentle, K.; Hu, Qiming; Chen, Jie; Li, Qiang; Liu, Yang; Gao, Li; Zhang, Xiaolong; Liu, Hai; Chen, Zhipeng; Zhu, Lizhi; Li, Fuming; Zhou, Yinan; Zeng, Zhong; Liu, Linzi; He, Jiyang

    2017-11-01

    Multifaceted asymmetric radiation as well as strong poloidal asymmetry of the electron density from the edge, dubbed as ‘MARFE’, has been observed in high electron density Ohmically heated plasmas on J-TEXT tokamak. Equilibrium reconstruction based on the measured data from the 17-channel FIR polarimeter-interferometer indicates that an asymmetric plasma current density distribution forms at the edge region and the plasma current shrinkage locates at the MARFE affected region. Furthermore, associated with the localized plasma current shrinkage, a locked mode MHD activity is excited, which then terminate the discharge with a major disruption. Localized plasma current shrinkage at the MARFE region is considered to be the direct cause for the density limit disruptions, and the proposed interpretation is consistent with the experimental observations.

  5. The effect of plasma minor-radius expansion in the current build-up phase of a large tokamak

    International Nuclear Information System (INIS)

    Kobayashi, Tomofumi; Tazima, Teruhiko; Tani, Keiji; Tamura, Sanae

    1977-03-01

    A plasma simulation code has been developed to study the plasma current build-up process in JT-60. Plasma simulation is made with a model which represents well overall plasma behavior of the present-day tokamaks. The external electric circuit is taken into consideration in simulation calculation. An emphasis is placed on the simulation of minor-radius expansion of the plasma and behavior of neutral particles in the plasma during current build-up. A calculation with typical parameters of JT-60 shows a week skin distribution in the current density and the electron temperature, if the minor radius of the plasma expands with build-up of the plasma current. (auth.)

  6. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

    1993-01-01

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  7. Observation of magnetohydrodynamics instabilities in ion Bernstein wave and lower-hybrid-current driving synergetic discharges on HT-7 tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Luo Jiarong; Shen Biao; Zhao Junyu; Hu Liqun; Zhu Yubao; Xu Guosheng; Asif, M.; Gao Xiang; Wan Baonian

    2004-01-01

    The normalized performance indicated by the product of β N H 89 >2 was achieved by a combination of the lower hybrid current driving (LHCD) and the ion Bernstein wave (IBW) heating in the HT-7 tokamak. More than 80% of the plasma current was sustained by the LHCD and the bootstrap current. Large edge pressure gradients were observed. The magnetohydrodynamic (MHD) instabilities were often driven to terminate the discharge or reduce the discharge performance, when the IBW resonant layer was near the rational surface. The resonant layer of the safety factor q=2 is located at 0.6 a with a=27 cm being the minor radius. The width of magnetic island (the poloidal mode number m=2) was about 2 cm. The plasma energy was reduced quickly by 30% by MHD instabilities. The behaviour of MHD instabilities is reported. A large sawtooth activity (m=1) was observed before inducing MHD (m=2)

  8. Observation of Cocurrent Toroidal Rotation in the EAST Tokamak with Lower-Hybrid Current Drive

    International Nuclear Information System (INIS)

    Shi Yuejiang; Xu Guosheng; Wang Fudi; Wang Mao; Fu Jia; Li Yingying; Zhang Wei; Zhang Wei; Chang Jiafeng; Lv Bo; Qian Jinping; Shan Jiafang; Liu Fukun; Ding Siye; Wan Baonian; Lee, Sang-Gon; Bitter, Manfred; Hill, Kenneth

    2011-01-01

    Lower-hybrid waves have been shown to induce a cocurrent change in toroidal rotation of up to 40 km/s in the L-mode plasma core region and 20 km/s in the edge of the EAST tokamak. This modification of toroidal rotation develops on different time scales. For the edge, the time scale is no more than 100 ms, but for the core the time scale is around 1 s. A simple model based on turbulent equipartition and thermoelectric pinch predicts the experimental results.

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

    International Nuclear Information System (INIS)

    Ibrahim, E.; Fowler, T.K.

    1987-06-01

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

  10. Investigation of runaway electrons in the current ramp-up by a fully non-inductive lower hybrid current drive on the EAST tokamak

    International Nuclear Information System (INIS)

    Lu, H W; Zha, X J; Zhong, F C; Hu, L Q; Zhou, R J

    2013-01-01

    The possibility of using a lower hybrid wave (LHW) to ramp up the plasma current (I p ) from a low level to a high enough level required for fusion burn in the EAST (experimental advanced superconducting tokamak) tokamak is examined experimentally. The focus in this paper is on investigating how the relevant plasma parameters evolve during the current ramp-up (CRU) phase driving by a lower hybrid current drive (LHCD) with poloidal field (PF) coil cut-off, especially the behaviors of runaway electrons generated during the CRU phase. It is found that the intensity of runaway electron emission increases first, and then decreases gradually as the discharge goes on under conditions of PF coil cut-off before LHW was launched into plasma, PF coil cut-off at the same time as LHW was launched into plasma, as well as PF coil cut-off after LHW was launched into plasma. The relevant plasma parameters, including H α line emission (Ha), impurity line emission (UV), soft x-ray emission and electron density n e , increase to a high level. The loop voltage decreases from positive to negative, and then becomes zero because of the cut-off of PF coils. Also, the magnetohydrodynamic activity takes place during the CRU driving by LHCD. (paper)

  11. Study of critical beta non-circular tokamak equilibria sustained in steady state by beam driven currents

    International Nuclear Information System (INIS)

    Okano, K.; Ogawa, Y.; Naitou, H.

    1988-07-01

    A new MHD-equilibrium/current-drive analysis code was developed to analyse the high beta tokamak equilibria consistent with the beam driven current profiles. In this new code, the critical beta equilibrium, which is stable against the ballooning mode, the kink mode and the Mercier mode, is determined first using MHD equilibrium and stability analysis codes (EQLAUS/ERATO). Then, the current drive parameters and the plasma parameters, required to sustain this critical beta equilibrium, are determined by iterative calculations. The beam driven current profiles are evaluated by the Fokker-Planck calculations on individual flux surfaces, where the toroidal effects on the beam ion and plasma electron trajectories are considered. The pressure calculation takes into account the beam ion and fast alpha components. A peculiarity of our new method is that the obtained solution is not only consistent with the MHD equilibrium but also consistent with the critical beta limit conditions, in the current profile and the pressure profile. Using this new method, β ∼ 21 % bean and β ∼ 6 % D-type critical beta equilibria were scanned for various parameters; the major radius, magnetic field, temperature, injection energy, etc. It was found that the achievable Q value for the bean type was always about 30 % larger than for the D-type cases, where Q = fusion power/beam power. With strong beanness, Q ∼ 6 for DEMO type tokamaks (∼500 MWth) and Q ∼ 20 for power reactor size (4.5 GWth) are achievable. On the other hand, the Q value would not exceed sixteen for the D-type machines. (author)

  12. A distributed control system for the lower-hybrid current drive system on the Tokamak de Varennes

    International Nuclear Information System (INIS)

    Bagdoo, J.; Guay, J.M.; Chaudron, G.A.; Decoste, R.; Demers, Y.; Hubbard, A.

    1990-01-01

    An rf current drive system with an output power of 1 MW at 3.7 GHz is under development for the Tokamak de Varennes. The control system is based on an Ethernet local-area network of programmable logic controllers as front end, personal computers as consoles, and CAMAC-based DSP processors. The DSP processors ensure the PID control of the phase and rf power of each klystron, and the fast protection of high-power rf hardware, all within a 40 μs loop. Slower control and protection, event sequencing and the run-time database are provided by the programmable logic controllers, which communicate, via the LAN, with the consoles. The latter run a commercial process-control console software. The LAN protocol respects the first four layers of the ISO/OSI 802.3 standard. Synchronization with the tokamak control system is provided by commercially available CAMAC timing modules which trigger shot-related events and reference waveform generators. A detailed description of each subsystem and a performance evaluation of the system will be presented. (orig.)

  13. A distributed control system for the lower-hybrid current drive system on the Tokamak de Varennes

    Science.gov (United States)

    Bagdoo, J.; Guay, J. M.; Chaudron, G.-A.; Decoste, R.; Demers, Y.; Hubbard, A.

    1990-08-01

    An rf current drive system with an output power of 1 MW at 3.7 GHz is under development for the Tokamak de Varennes. The control system is based on an Ethernet local-area network of programmable logic controllers as front end, personal computers as consoles, and CAMAC-based DSP processors. The DSP processors ensure the PID control of the phase and rf power of each klystron, and the fast protection of high-power rf hardware, all within a 40 μs loop. Slower control and protection, event sequencing and the run-time database are provided by the programmable logic controllers, which communicate, via the LAN, with the consoles. The latter run a commercial process-control console software. The LAN protocol respects the first four layers of the ISO/OSI 802.3 standard. Synchronization with the tokamak control system is provided by commercially available CAMAC timing modules which trigger shot-related events and reference waveform generators. A detailed description of each subsystem and a performance evaluation of the system will be presented.

  14. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  15. Measurement of anisotropic soft X-ray emission during radio-frequency current drive in the JFT-2M tokamak

    International Nuclear Information System (INIS)

    Kawashima, Hisato; Matoba, Tohru; Hoshino, Katsumichi; Kawakami, Tomohide; Yamamoto, Takumi; Hasegawa, Mitsuru; Fuchs, Gerhard; Uesugi, Yoshihiko.

    1994-01-01

    A new vertical soft X-ray pulse height analyzer (PHA) system and a tangential PHA system were used to measure the anisotropy of soft X-ray emission during lower-hybrid current drive (LHCD) and also during current drive by the combination of LHCD and electron cyclotron resonance heating (ECRH) in the JFT-2M tokamak. The strong soft X-ray emission was measured in the parallel forward direction during LHCD. When ECRH was applied during LHCD, the perpendicular emission was enhanced. The high-energy electron velocity distribution was evaluated by comparing the measured and calculated X-ray spectra. The distribution form was consistent with the theoretical prediction based on the electron Landau damping of lower-hybrid waves and the electron cyclotron damping of electron cyclotron waves for reasonable energy ranges. (author)

  16. Experimental Study of Reversed Shear Alfven Eigenmodes During The Current Ramp In The Alcator C-Mod Tokamak

    International Nuclear Information System (INIS)

    Edlund, E.M.; Porkolab, M.; Kramer, G.J.; Lin, L.; Lin, Y.; Tsuji, N.; Wukitch, S.J.

    2010-01-01

    Experiments conducted in the Alcator C-Mod tokamak at MIT have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs throughout the current ramp provides a constraint on the evolution of q min , a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive a measure of the adiabatic index, a measure of the plasma compressibility. This scaling bounds the adiabatic index at 1.40 ± 0.15 used in MHD models and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7/4.

  17. Realizing steady-state tokamak operation for fusion energy

    International Nuclear Information System (INIS)

    Luce, T. C.

    2011-01-01

    Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

  18. A current-pulsed power supply with rapid rising and falling edges for magnetic perturbation coils on the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Yan, M.X.; Rao, B.; Ding, Y.H.; Hu, Q.M.; Hu, F.R.; Li, D.; Li, M.; Ji, X.K.; Xu, G.; Zheng, W.; Jiang, Z.H.

    2017-01-01

    Highlights: • The power supply is required to have rapid rising and falling edges. • A modified topology based on the buck chopper of current-pulsed power supply is presented and analyzed. • An entity meeting the electrical requirements has been constructed. • The spike voltage of IGBT is qualitatively analyzed. - Abstract: This study presents the design and principle of a current-pulsed power supply (CPPS) for the tearing mode (TM) feedback control of the J-TEXT tokamak. CPPS is a new method of stabilizing large magnetic islands and accelerating mode rotation through the use of modulated magnetic perturbation. In this application, continuous magnetic perturbation pulse trains with frequency of 1 kHz to kHz, amplitude of 0.25 G, and duty ratio of 20%–50% are required generating via in-vessel magnetic coils. A modified topology based on buck chopper is raised to satisfy the demands of inductive load. This modified topology is characterized by high frequency, rapid rising and falling edges, and large amplitude of current pulses. Appropriate RCD snubber circuit is applied to protect the Insulated Gate Bipolar Transistor (IGBT) switch device. Equipment with peak current that reaches 1 kA, frequency that ranges from 1 kHz to 3 kHz, and rising and falling time within 100 μs was constructed and applied to physical experiment.

  19. A current-pulsed power supply with rapid rising and falling edges for magnetic perturbation coils on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yan, M.X. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Rao, B., E-mail: borao@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Ding, Y.H.; Hu, Q.M.; Hu, F.R.; Li, D.; Li, M.; Ji, X.K.; Xu, G.; Zheng, W.; Jiang, Z.H. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2017-02-15

    Highlights: • The power supply is required to have rapid rising and falling edges. • A modified topology based on the buck chopper of current-pulsed power supply is presented and analyzed. • An entity meeting the electrical requirements has been constructed. • The spike voltage of IGBT is qualitatively analyzed. - Abstract: This study presents the design and principle of a current-pulsed power supply (CPPS) for the tearing mode (TM) feedback control of the J-TEXT tokamak. CPPS is a new method of stabilizing large magnetic islands and accelerating mode rotation through the use of modulated magnetic perturbation. In this application, continuous magnetic perturbation pulse trains with frequency of 1 kHz to kHz, amplitude of 0.25 G, and duty ratio of 20%–50% are required generating via in-vessel magnetic coils. A modified topology based on buck chopper is raised to satisfy the demands of inductive load. This modified topology is characterized by high frequency, rapid rising and falling edges, and large amplitude of current pulses. Appropriate RCD snubber circuit is applied to protect the Insulated Gate Bipolar Transistor (IGBT) switch device. Equipment with peak current that reaches 1 kA, frequency that ranges from 1 kHz to 3 kHz, and rising and falling time within 100 μs was constructed and applied to physical experiment.

  20. Current drive with fast waves, electron cyclotron waves, and neutral injection in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Prater, R.; Petty, C.C.; Pinsker, R.I.

    1993-01-01

    Current drive experiments have been performed on the DIII-D tokamak using fast waves, electron cyclotron waves, and neutral injection. Fast wave experiments were performed using a 4-strap antenna with 1 MW of power at 60 MHz. These experiments showed effective heating of electrons, with a global heating efficiency equivalent to that of neutral injection even when the single pass damping was calculated to be as small as 5%. The damping was probably due to the effect of multiple passes of the wave through the plasma. Fast wave current drive experiments were performed with a toroidally directional phasing of the antenna straps. Currents driven by fast wave current drive (FWCD) in the direction of the main plasma current of up to 100 kA were found, not including a calculated 40 kA of bootstrap current. Experiments with FWCD in the counter current direction showed little current drive. In both cases, changes in the sawtooth behavior and the internal inductance qualitatively support the measurement of FWCD. Experiments on electron cyclotron current drive have shown that 100 kA of current can be driven by 1 MW of power at 60 GHz. Calculations with a Fokker-Planck code show that electron cyclotron current drive (ECCD) can be well predicted when the effects of electron trapping and of the residual electric field are included. Experiments on driving current with neutral injection showed that effective current drive could be obtained and discharges with full current drive were demonstrated. Interestingly, all of these methods of current drive had about the same efficiency. (Author)

  1. Current drive with fast waves, electron cyclotron waves, and neutral injection in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Prater, R.; Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Ikel, H.; Lin-Liu, Y.R.; Luce, T.C.; James, R.A.; Porkolab, M.; Baity, F.W.; Goulding, R.H.; Hoffmann, D.J.; Kawashima, H.; Trukhin, V.

    1992-09-01

    Current drive experiments have been performed on the DIII-D tokamak using fast waves, electron cyclotron waves, and neutral injection. Fast wave experiments were performed using a 4-strap antenna with 1 MW of power at 60 MHz. These experiments showed effective heating of electrons, with a global heating efficiency equivalent to that of neutral injection even when the single pass damping was calculated to be as small as 5%. The damping was probably due to the effect of multiple passes of the wave through the plasma. Fast wave current drive experiments were performed with a toroidally directional phasing of the antenna straps. Currents driven by fast wave current drive (FWCD) in the direction of the main plasma current of up to 100 kA were found, not including a calculated 40 kA of bootstrap current. Experiments with FWCD in the counter current direction showed little current drive. In both cases, changes in the sawtooth behavior and the internal inductance qualitatively support the measurement of FWCD. Experiments on electron cyclotron current drive have shown that 100 kA of current can be driven by 1 MW of power at 60 GHz. Calculations with a Fokker-Planck code show that electron cyclotron current drive (ECCD) can be well predicted when the effects of electron trapping and of the residual electric field are included. Experiments on driving current with neutral injection showed that effective current drive could be obtained and discharges with full current drive were demonstrated. Interestingly, all of these methods of current drive had about the same efficiency, 0.015 x 10 20 MA/MW/m 2

  2. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  3. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  4. Continuous infusion in haemophilia: current practice in Europe

    NARCIS (Netherlands)

    Batorova, A.; Holme, P.; Gringeri, A.; Richards, M.; Hermans, C.; Altisent, C.; Lopez-Fernández, M.; Fijnvandraat, K.

    2012-01-01

    . Continuous infusion (CI) of factor VIII (FVIII) is an effective method for replacement therapy in haemophilia. Recently, concerns have been raised regarding association of CI with the development of inhibitors. The aim of this study was to gain information on the current practices in Europe

  5. CONTROL SYSTEM FOR THE LITHIUM BEAM EDGE PLASMA CURRENT DENSITY DIAGNOSTIC ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PEAVY, J.J.; CARY, W.P; THOMAS, D.M; KELLMAN, D.H.; HOYT, D.M; DELAWARE, S.W.; PRONKO, S.G.E.; HARRIS, T.E.

    2004-03-01

    OAK-B135 An edge plasma current density diagnostic employing a neutralized lithium ion beam system has been installed on the DIII-D tokamak. The lithium beam control system is designed around a GE Fanuc 90-30 series PLC and Cimplicity(reg s ign) HMI (Human Machine Interface) software. The control system operates and supervises a collection of commercial and in-house designed high voltage power supplies for beam acceleration and focusing, filament and bias power supplies for ion creation, neutralization, vacuum, triggering, and safety interlocks. This paper provides an overview of the control system, while highlighting innovative aspects including its remote operation, pulsed source heating and pulsed neutralizer heating, optimizing beam regulation, and beam ramping, ending with a discussion of its performance

  6. Numerical study of the electron heating and current drive by the fast waves in the JFT-2M tokamak plasma

    International Nuclear Information System (INIS)

    Yamamoto, Takumi; Uesugi, Yoshihiko; Hoshino, Katsumichi; Kawashima, Hisato; Ohtsuka, Hideo

    1986-08-01

    A 200 MHz fast wave experiment for the JET-2M tokamak is examined. Noticeable single-path electron Landau damping of the fast waves with the parallel refractive index of N // = 4 is expected in the plasma with electron temperature more than 2.5 keV at the electron density of n e = 1.5 x 10 19 m -3 . Furthermore, it is shown that 8 kA of the plasma current is driven by the fast waves with N //≅ 2 at n e = 3 x 10 19 m -3 in the single-path damping when 100 kW of the rf power radiates into the plasma in the presence of the hot electrons with the temperature of 19 keV and the fraction of the density of 2 %. (author)

  7. 3D eddy-current distribution in a tokamak first wall during a plasma disruption using 'TRIFOU'

    International Nuclear Information System (INIS)

    Chaussecourte, P.; Bossavit, A.; Verite, J.C.; Crutzen, Y.R.

    1989-01-01

    In fusion reactor studies there is a lack of knowledge concerning the electromagnetic-type of phenomena generated by a plasma disruption event (rapid quenching of the plasma current). The induced eddy current distribution in space and time in the passive conducting structural components surrounding the plasma ring needs to be accurately investigated. TRIFOU is a full 3D eddy-current computer program based on a mixed FEM and BIEM technique, using the magnetic field, h, as a state variable, It has already been used in various areas of interest including static or rotating machines, non-destructive testing, induction heating, and research devices such as tokamaks. It can take into account various geometries and a wide range of physical situations (time dependency, physical properties, etc.). The present application is related to the eddy-current situation arising from a strong electromagnetic transient generated in the NET (Next European Torus) first wall segment. With respect to previous numerical simulations, the general 3D approach for the current density shows different eddy current circulations in the front/side shells and in the stiff back plate. The results obtained by TRIFOU are illustrated by means of advanced computer graphic displays and an animation movie. (orig.)

  8. Observation of SOL Current Correlated with MHD Activity in NBI-heated DIII-D Tokamak Discharges

    International Nuclear Information System (INIS)

    Takahashi, H.; Fredrickson, E.D.; Schaffer, M.J.; Austin, M.E.; Evans, T.E.; Lao, L.L.; Watkins, J.G.

    2004-01-01

    This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found during MHD activity that are: (1) slowly growing after a mode-locking-like event, (2) oscillating in the several kHz range and phase-locked with magnetic and electron temperature oscillations, (3) rapidly growing with a sub-ms time scale during a thermal collapse and a current quench, and (4) spiky in temporal behavior and correlated with spiky features in Da signals commonly identified with the edge localized mode (ELM). These SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a toroidally non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) toroidally non-axisymmetric thermoelectric potential, (b) electromotive force (EMF) from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. An effect is found, stemming from the shear in the field line pitch angle, that mitigates the efficacy of a toroidally non-axisymmetric SOLC to generate a toroidally non-axisymmetric error field. Other potential magnetic consequences of the SOLC are identified: (i) its error field can introduce complications in feedback control schemes for stabilizing MHD activity and (ii) its toroidally non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of a SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance

  9. User's guide for SLWDN9, a code for calculating flux-surfaced-averaging of alpha densities, currents, and heating in non-circular tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Miley, G.M.

    1980-03-01

    The code calculates flux-surfaced-averaged values of alpha density, current, and electron/ion heating profiles in realistic, non-circular tokamak plasmas. The code is written in FORTRAN and execute on the CRAY-1 machine at the Magnetic Fusion Energy Computer Center

  10. Effect of Wave Accessibility on Lower Hybrid Wave Current Drive in Experimental Advanced Superconductor Tokamak with H-Mode Operation

    International Nuclear Information System (INIS)

    Li Xin-Xia; Xiang Nong; Gan Chun-Yun

    2015-01-01

    The effect of the wave accessibility condition on the lower hybrid current drive in the experimental advanced superconductor Tokamak (EAST) plasma with H-mode operation is studied. Based on a simplified model, a mode conversion layer of the lower hybrid wave between the fast wave branch and the slow wave branch is proved to exist in the plasma periphery for typical EAST H-mode parameters. Under the framework of the lower hybrid wave simulation code (LSC), the wave ray trajectory and the associated current drive are calculated numerically. The results show that the wave accessibility condition plays an important role on the lower hybrid current drive in EAST plasma. For wave rays with parallel refractive index n ‖ = 2.1 or n ‖ = 2.5 launched from the outside midplane, the wave rays may penetrate the core plasma due to the toroidal geometry effect, while numerous reflections of the wave ray trajectories in the plasma periphery occur. However, low current drive efficiency is obtained. Meanwhile, the wave accessibility condition is improved if a higher confined magnetic field is applied. The simulation results show that for plasma parameters under present EAST H-mode operation, a significant lower hybrid wave current drive could be obtained for the wave spectrum with peak value n ‖ = 2.1 if a toroidal magnetic field B T = 2.5 T is applied. (paper)

  11. Optimization and control of plasma shape and current profile in non-circular cross-section tokamaks

    International Nuclear Information System (INIS)

    Moore, R.W.; Bernard, L.C.; Chan, V.S.

    1981-01-01

    Tokamaks with elongated, non-circular cross-sections are under consideration as fusion reactors because they have the potential for stable operation at high β. Ideal MHD theory, however, predicts that careful current profile control will be required to achieve the potential high-β advantages of non-circular cross-sections. In this paper, high-β equilibria which are stable to all ideal MHD modes are found by optimizing the plasma shape and current profile for doublets, up-down asymmetric dees, and symmetric dees. The ideal MHD stability of these equilibria for low toroidal mode number n is analysed with a global MHD stability code, GATO. The stability to high-n modes is analysed with a localized ballooning code, BLOON. The attainment of high β is facilitated by an automated optimization search on shape and current parameters. The equilibria are calculated with a free-boundary equilibrium code using coils appropriate for the Doublet III experimental device. The optimal equilibria are characterized by broad current profiles with values of βsub(poloidal) approximately equal to 1. Experimental realization of the shapes and current profiles giving the highest β limits is explored with a 1 1/2-D transport code, which simulates the time evolution of the 2-D MHD equilibrium while calculating consistent current profiles from a 1-D transport model. Transport simulations indicate that nearly optimal shapes may be obtained provided that the currents in the field-shaping coils are appropriately programmed and the plasma current profile is sufficiently broad. Obtaining broad current profiles is possible by current ramping, neutral-beam heating, and electron-cyclotron heating. With combinations of these techniques it is possible to approach the optimum β predicted by the MHD theory. (author)

  12. 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

  13. 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.

  14. Calibration of power systems and measurements of discharge currents generated for different coils in the EGYPTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Hegazy, H.; Žáček, František

    2006-01-01

    Roč. 25, 1-2 (2006), s. 73-86 ISSN 0164-0313 Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * EGYPTOR tokamak * Rogowski coil Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.381, year: 2006

  15. Radial profiles of hard X-ray emission during steady state current drive in the TRIAM-1M tokamak

    International Nuclear Information System (INIS)

    Nakamura, Y.; Takabatake, Y.; Jotaki, E.; Moriyama, S.; Nagao, A.; Nakamura, K.; Hiraki, N.; Itoh, S.

    1990-01-01

    The hard X-ray emission from the TRIAM-1M tokamak plasma during steady state lower hybrid current drive with a discharge duration of a few minutes was measured with sodium iodide scintillation spectrometers. The radial profiles of the X-ray emission were also measured and indicate that, in the low density regime (n e =(1-3)x10 12 cm -3 ), the current carrying high energy electrons are mainly in the inner region of the plasma column and their radial profile remains unchanged during current drive. On the other hand, high density discharges (n e =(3-6)x10 12 cm -3 ) are always accompanied by an abrupt drop of the plasma current, and the X-ray emission profile changes from peaked to broad. This change can be attributed to the conditions of wave accessibility. As the electron density increases, the accessibility of the plasma to lower hybrid waves with low values of the parallel wave number n parallel is significantly reduced and high energy electrons resonating with the waves are produced at the plasma periphery. Interaction of these electrons with the limiters causes an increase of the electron density in this region; waves with low n parallel then become completely excluded from the inner part of the plasma column. This interpretation is supported by measurements of the density profile and impurity radiation, and has been confirmed in an investigation of discharges with additional gas puffing. (author). 17 refs, 21 figs

  16. Continuous-waveform constant-current isolated physiological stimulator

    Science.gov (United States)

    Holcomb, Mark R.; Devine, Jack M.; Harder, Rene; Sidorov, Veniamin Y.

    2012-04-01

    We have developed an isolated continuous-waveform constant-current physiological stimulator that is powered and controlled by universal serial bus (USB) interface. The stimulator is composed of a custom printed circuit board (PCB), 16-MHz MSP430F2618 microcontroller with two integrated 12-bit digital to analog converters (DAC0, DAC1), high-speed H-Bridge, voltage-controlled current source (VCCS), isolated USB communication and power circuitry, two isolated transistor-transistor logic (TTL) inputs, and a serial 16 × 2 character liquid crystal display. The stimulators are designed to produce current stimuli in the range of ±15 mA indefinitely using a 20V source and to be used in ex vivo cardiac experiments, but they are suitable for use in a wide variety of research or student experiments that require precision control of continuous waveforms or synchronization with external events. The device was designed with customization in mind and has features that allow it to be integrated into current and future experimental setups. Dual TTL inputs allow replacement by two or more traditional stimulators in common experimental configurations. The MSP430 software is written in C++ and compiled with IAR Embedded Workbench 5.20.2. A control program written in C++ runs on a Windows personal computer and has a graphical user interface that allows the user to control all aspects of the device.

  17. Automated Studies of Continuing Current in Lightning Flashes

    Science.gov (United States)

    Martinez-Claros, Jose

    Continuing current (CC) is a continuous luminosity in the lightning channel that lasts longer than 10 ms following a lightning return stroke to ground. Lightning flashes following CC are associated with direct damage to power lines and are thought to be responsible for causing lightning-induced forest fires. The development of an algorithm that automates continuing current detection by combining NLDN (National Lightning Detection Network) and LEFA (Langmuir Electric Field Array) datasets for CG flashes will be discussed. The algorithm was applied to thousands of cloud-to-ground (CG) flashes within 40 km of Langmuir Lab, New Mexico measured during the 2013 monsoon season. It counts the number of flashes in a single minute of data and the number of return strokes of an individual lightning flash; records the time and location of each return stroke; performs peak analysis on E-field data, and uses the slope of interstroke interval (ISI) E-field data fits to recognize whether continuing current (CC) exists within the interval. Following CC detection, duration and magnitude are measured. The longest observed C in 5588 flashes was 631 ms. The performance of the algorithm (vs. human judgement) was checked on 100 flashes. At best, the reported algorithm is "correct" 80% of the time, where correct means that multiple stations agree with each other and with a human on both the presence and duration of CC. Of the 100 flashes that were validated against human judgement, 62% were hybrid. Automated analysis detects the first but misses the second return stroke in many cases where the second return stroke is followed by long CC. This problem is also present in human interpretation of field change records.

  18. Local magnetic shear control in a tokamak via fast wave minority ion current drive: Theory and experiments in JET

    International Nuclear Information System (INIS)

    Bhatnagar, V.P.; Start, D.F.H.; Jacquinot, J.; Chaland, F.; Cherubini, A.; Porcelli, F.

    1994-01-01

    When an ion cyclotron resonance heating (ICRH) antenna array is phased (Δ Φ ≠ 0 or π), the excited asymmetric k parallel spectrum can drive non-inductive currents by interaction of fast waves both with electrons (transit time magnetic pumping (e-TTMP) and Landau damping (e-LD)) and with ions at minority (fundamental) or harmonic cyclotron resonances, depending upon the scenario. On the basis of earlier theories, a simplified description is presented that includes the minority ion and electron current drive effects simultaneously in a 3-D ray tracing calculation in the tokamak geometry. The experimental results of sawtooth stabilization or destabilization in JET using the minority ion current drive scheme are presented. This scheme allows a modification of the local current density gradient (or the magnetic shear) at the q = 1 surface resulting in a control of a sawteeth. The predictions of the above model of current drive and its effects on sawtooth period calculated in conjunction with a model of stability of internal resistive kink modes, that encompasses the effects of both the fast particle pressure and the local (q = 1) magnetic shear, are found to be qualitatively in good agreement with the experimental results. Further, the results are discussed of our model of fast wave current drive scenarios of magnetic shear reversal with a view to achieving long duration high confinement regimes in the forthcoming experimental campaign on JET. Finally, the results are presented of minority current drive for sawtooth control in next step devices such as the International Thermonuclear Experimental Reactor (ITER). (author). 44 refs, 23 figs, 3 tabs

  19. Stable existence of central current hole in the JT-60U tokamak

    International Nuclear Information System (INIS)

    Miura, Y.; Fujita, T.; Oikawa, T.

    2003-01-01

    In an extreme state of a reversed magnetic shear configuration, it was found in JT-60U that there is almost no plasma current in the central region (called Current Hole). The Current Hole region extends to 40% of the plasma minor radius and it exists stably for several seconds. The Current Hole is formed by the growth of the bootstrap current and it is impossible to drive current in either positive or negative direction by ECH or N-NB inside the Current Hole. In that region, there is almost no gradient of density, temperature and toroidal rotation velocity. It means that there is almost no confinement in the Current Hole and the large energy in that region is sustained only by an internal transport barrier (ITB). The effects of the Current Hole on particle orbits and the effects on an error field on the Current Hole are also discussed. (author)

  20. Mechanisms of the negative synergy effect between electron cyclotron current drive and lower hybrid current drive in tokamak

    International Nuclear Information System (INIS)

    Chen Shaoyong; Hong Binbin; Tang Changjian; Yang Wen; Zhang Xinjun

    2013-01-01

    The synergy current drive by combining electron cyclotron wave (ECW) with lower hybrid wave (LHW) can be used to either increase the noninductive current drive efficiency or shape the plasma current profile. In this paper, the synergy current drive by ECW and LHW is studied with numerical simulation. The nonlinear relationship between the wave powers and the synergy current of ECW and LHW is revealed. When the LHW power is small, the synergy current reduces as the ECW power increases, and the synergy current is even reduced to lower than zero, which is referred as negative synergy in the this context. Research shows that the mechanism of the negative synergy is the peaking effect of LHW power profile and the trapped electrons effect. The present research is helpful for understanding the physics of synergy between electron cyclotron current drive and lower hybrid current drive, it can also instruct the design of experiments. (authors)

  1. Surface currents associated with external kink modes in tokamak plasmas during a major disruption

    Science.gov (United States)

    Ng, C. S.; Bhattacharjee, A.

    2017-10-01

    The surface current on the plasma-vacuum interface during a disruption event involving kink instability can play an important role in driving current into the vacuum vessel. However, there have been disagreements over the nature or even the sign of the surface current in recent theoretical calculations based on idealized step-function background plasma profiles. We revisit such calculations by replacing step-function profiles with more realistic profiles characterized by a strong but finite gradient along the radial direction. It is shown that the resulting surface current is no longer a delta-function current density, but a finite and smooth current density profile with an internal structure, concentrated within the region with a strong plasma pressure gradient. Moreover, this current density profile has peaks of both signs, unlike the delta-function case with a sign opposite to, or the same as the plasma current. We show analytically and numerically that such current density can be separated into two parts, with one of them, called the convective current density, describing the transport of the background plasma density by the displacement, and the other part that remains, called the residual current density. It is argued that consideration of both types of current density is important and can resolve past controversies.

  2. Influence of helical external driven current on nonlinear resistive tearing mode evolution and saturation in tokamaks

    Science.gov (United States)

    Zhang, W.; Wang, S.; Ma, Z. W.

    2017-06-01

    The influences of helical driven currents on nonlinear resistive tearing mode evolution and saturation are studied by using a three-dimensional toroidal resistive magnetohydrodynamic code (CLT). We carried out three types of helical driven currents: stationary, time-dependent amplitude, and thickness. It is found that the helical driven current is much more efficient than the Gaussian driven current used in our previous study [S. Wang et al., Phys. Plasmas 23(5), 052503 (2016)]. The stationary helical driven current cannot persistently control tearing mode instabilities. For the time-dependent helical driven current with f c d = 0.01 and δ c d < 0.04 , the island size can be reduced to its saturated level that is about one third of the initial island size. However, if the total driven current increases to about 7% of the total plasma current, tearing mode instabilities will rebound again due to the excitation of the triple tearing mode. For the helical driven current with time dependent strength and thickness, the reduction speed of the radial perturbation component of the magnetic field increases with an increase in the driven current and then saturates at a quite low level. The tearing mode is always controlled even for a large driven current.

  3. Current-drive on the Versator-II tokamak with a slotted-waveguide fast-wave coupler

    International Nuclear Information System (INIS)

    Colborn, J.A.

    1987-11-01

    A slotted-waveguide fast-wave coupler has been constructed, without dielectric, and used to drive current on the Versator-II tokamak. Up to 35 kW of net microwave power at 2.45 GHz has been radiated into plasmas with 2 x 10 12 cm -3 ≤ mean of n/sub e/ ≤ 1.2 x 10 13 cm -3 and B/sub tor/ ≅ 1.0 T. The launched spectrum had a peak near N/sub parallel/ = -2.0 and a larger peak near N/sub parallel/ = 0.7. Radiating efficiency of the antenna was roughly independent of antenna position except when the antenna was at least 0.2 cm outside the limiter, in which case the radiating efficiency slightly improved as the antenna was moved farther outside. When the coupler was inside the limiter, radiating efficiency improved moderately with increased mean of n/sub e/. Current-drive efficiency was comparable to that of the slow wave and was not affected when the antenna spectrum was reversed; however, no current was driven for mean of n/sub e/ ≤ 2 x 10 12 cm -3 . These results indicate the fast wave was launched, but a substantial part of the power may have been mode-converted to the slow wave, possibly via a downshift in N/sub parallel/, and these slow waves may have been responsible for most of the driven current. Relevant theory for waves in plasma, current-drive efficiency, and coupling of the slotted-waveguide is discussed, the antenna design method is explained, and future work, including the construction of a much-improved probe-fed antenna, is described. 42 refs., 45 figs

  4. Optimum launching of electron-cyclotron power for localized current drive in a hot tokamak

    International Nuclear Information System (INIS)

    Smith, G.R.

    1989-05-01

    Optimum launch parameters are determined for localized electron-cyclotron current drive near the magnetic axis and the q=2 surface by solving several minimization problems. For central current drive, equatorial and bottom launch are compared. Localized current drive near q=2 is studied for equatorial launch and for an alternative outside launch geometry that may be better for suppressing tearing modes and controlling disruptions. 6 refs., 2 figs

  5. Optimization and control of the plasma shape and current profile in noncircular cross-section tokamaks

    International Nuclear Information System (INIS)

    Moore, R.W.; Bernard, L.C.; Chan, V.S.; Davidson, R.H.; Dobrott, D.R.; Helton, F.J.; Miller, R.L.; Pfeiffer, W.; Waltz, R.E.; Wang, T.S.

    1980-06-01

    High-β equilibria which are stable to all ideal MHD modes are found by optimizing the plasma shape and current profile for doublets, up-down asymmetric dees, and symmetric dees. The ideal MHD stability of these equilibria for low toroidal mode number n is analyzed with a global MHD stability code, GATO. The stability to high-n modes is analyzed with a localized ballooning code, BLOON. The attainment of high β is facilitated by an automated optimization search on shape and current parameters. The equilibria are calculated with a free-boundary equilibrium code using coils appropriate for the Doublet III experimental device. The optimal equilibria are characterized by broad current profiles with values of β/sub poloidal/ approx. =1. Experimental realization of the shapes and current profiles giving the highest β limits is explored with a 1 1/2-D transport code, which simulates the time evolution of the 2-D MHD equilibrium while calculating consistent current profiles from a 1-D transport model. Transport simulations indicate that nearly optimal shapes may be obtained provided that the currents in the field-shaping coils are appropriately programmed and the plasma current profile is sufficiently broad. Obtaining broad current profiles is possible by current ramping, neutral beam heating, and electron cyclotron heating. With combinations of these techniques it is possible to approach the optimum β predicted by the MHD theory

  6. Non-inductive plasma initiation and plasma current ramp-up on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Oosako, T.; Shinya, T.; Ambo, T.; Furui, H.; Kato, K.; Nakanishi, A.; Sakamoto, T.; Kakuda, H.; Wakatsuki, T.; Hashimoto, T.; Hiratsuka, J.; Kasahara, H.; Kumazawa, R.; Mutoh, T.; Saito, K.; Seki, T.; Moeller, C.P.; Nagashima, Y.

    2013-01-01

    Plasma current (I p ) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (∼1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further I p ramp-up (to ∼10 kA) is most efficient with waves in the LH frequency range. I p ramp-up to 15 kA was achieved with 60 kW of net RF power P RF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to I p , consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of I p ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array (‘grill’) antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013. (paper)

  7. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

    Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

  8. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  9. Modeling of LH current drive in self-consistent elongated tokamak MHD equilibria

    International Nuclear Information System (INIS)

    Blackfield, D.T.; Devoto, R.S.; Fenstermacher, M.E.; Bonoli, P.T.; Porkolab, M.; Yugo, J.

    1989-01-01

    Calculations of non-inductive current drive typically have been used with model MHD equilibria which are independently generated from an assumed toroidal current profile or from a fit to an experiment. Such a method can lead to serious errors since the driven current can dramatically alter the equilibrium and changes in the equilibrium B-fields can dramatically alter the current drive. The latter effect is quite pronounced in LH current drive where the ray trajectories are sensitive to the local values of the magnetic shear and the density gradient. In order to overcome these problems, we have modified a LH simulation code to accommodate elongated plasmas with numerically generated equilibria. The new LH module has been added to the ACCOME code which solves for current drive by neutral beams, electric fields, and bootstrap effects in a self-consistent 2-D equilibrium. We briefly describe the model in the next section and then present results of a study of LH current drive in ITER. 2 refs., 6 figs., 2 tabs

  10. Plasma current startup by lower hybrid waves in the JIPP T-IIU tokamak

    International Nuclear Information System (INIS)

    Toi, K.; Ohkubo, K.; Kawahata, K.

    1987-04-01

    This paper describes the characteristic behaviours of lower hybrid current startup in JIPP T-IIU. The current startup is carried out by injection of 800 MHz lower hybrid waves into cold and low density plasmas (Te = 10 - 20 eV, n-bar e = 1 - 2 x 10 12 cm -3 ) produced by electron cyclotron resonance or lower hybrid waves only. The plasma current rises up with a characteristic rise time τ r (> approx 30 - 50 ms) and approaches a quasi-steady state value I pm (= 5 - 20 kA), when LHW power of 10 - 50 kW is injected into a torus, controlling the vertical field. The rise time is inversely proportional to the bulk electron density n-bar e , and is comparable to the collision time of current-carrying high energy electrons with bulk plasmas. On the other hand, the current drive efficiency in the quasi-steady state is almost independent of n-bar e , i.e., I pm /P LH = 0.4 - 0.7 A/W for n-bar e = 0.8 - 4 x 10 12 cm -3 . The conversion efficiency of rf energy injected into the torus is typically 5 % during current rise phase, and 10 % at the most efficient case. The effects of the initial injection of ECH power and the observed parametric instabilities on the current startup are investigated from a viewpoint of seed current generation. During rapid current rise when appreciably negative loop voltage is observed the bulk electrons are heated up to 150 eV. Various heating mechanisms responsible for the bulk electron heating are discussed. (author)

  11. Effect of the X-point on the stability of the edge-current-driven MHD mode in Tokamaks

    International Nuclear Information System (INIS)

    Kwon, Ohjin

    2010-01-01

    Quasi-periodic bursts of edge magnetohydrodynamic (MHD) activities, called edge localized modes (ELMs), have been observed in many tokamaks during the H-mode. The high level of heat and particle transport associated with ELMs may cause serious damage to divertors or plasma facing components. It is therefore important to understand the underlying physics of ELMs. We have numerically investigated the effect of the X-point on the stability of the peeling mode, which is thought to be one of the MHD instabilities responsible for small ELMs. Equilibria with pressure and current profiles, which are unstable to the pure peeling mode for moderately elongated plasma, have been used. The X-point in a diverted plasma has been simulated by introducing of a hump in the plasma boundary. The position, depth and width of the X-point have been varied, and their effect on the stability of the peeling mode has been investigated. We have shown that the peeling mode growth rate decreases as the depth increases. This effect is greater for smaller widths for all positions of the X-point considered. Therefore, a sharper X-point is more efficient in stabilizing the peeling mode. Increasing the depth acts to increase the magnetic shear, the stabilizing effect of which has been shown to have very little dependence on the position or the width of the X-point.

  12. Shear flow generation and transport barrier formation on rational surface current sheets in tokamaks

    International Nuclear Information System (INIS)

    Wang Xiaogang; Xiao Chijie; Wang Jiaqi

    2009-01-01

    Full text: A thin current sheet with a magnetic field component in the same direction can form the electrical field perpendicularly pointing to the sheet, therefore an ExB flow with a strong shear across the current sheet. An electrical potential well is also found on the rational surface of RFP as well as the neutral sheet of the magnetotail with the E-field pointing to the rational (neutral) surface. Theoretically, a current singularity is found to be formed on the rational surface in ideal MHD. It is then very likely that the sheet current on the rational surfaces will generate the electrical potential well in its vicinity so the electrical field pointing to the sheet. It results in an ExB flow with a strong shear in the immediate neighborhood of the rational surface. It may be the cause of the transport barrier often seen near the low (m, n) rational surfaces with MHD signals. (author)

  13. Generation of stationary current in a tokamak by electron cyclotron waves

    International Nuclear Information System (INIS)

    Parail, V.V.; Pereverzev, G.V.

    1982-01-01

    Analytical expression for stationary longitudinal current generated in plasma with electron-cyclotron (EC) waves has been derived on the basis of a kinetic equation for electrons with provision for electron-electron and electron- ion collisions. Comparative analysis of efficiency of current excitation with EC and low hybrid (LH) waves has been carried out. It is shown that under similar conditions (for the same introduced powers and the same intervals of interaction of LH waves and electrons) the current value generated with LH waves turns out to be functionally (Vsub(o)/Vsub(e))sup(2) times higher as compared with the current generated with EC waves (vsub(o)-initial velocity of electrons, Vsub(e)-√2Tsub(e)/m) [ru

  14. Separation of Evans and Hiro currents in VDE of tokamak plasma

    Science.gov (United States)

    Galkin, Sergei A.; Svidzinski, V. A.; Zakharov, L. E.

    2014-10-01

    Progress on the Disruption Simulation Code (DSC-3D) development and benchmarking will be presented. The DSC-3D is one-fluid nonlinear time-dependent MHD code, which utilizes fully 3D toroidal geometry for the first wall, pure vacuum and plasma itself, with adaptation to the moving plasma boundary and accurate resolution of the plasma surface current. Suppression of fast magnetosonic scale by the plasma inertia neglecting will be demonstrated. Due to code adaptive nature, self-consistent plasma surface current modeling during non-linear dynamics of the Vertical Displacement Event (VDE) is accurately provided. Separation of the plasma surface current on Evans and Hiro currents during simulation of fully developed VDE, then the plasma touches in-vessel tiles, will be discussed. Work is supported by the US DOE SBIR Grant # DE-SC0004487.

  15. A study on the fusion reactor - A study on wave physics of fast wave heating and the current drive in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Su Won; Yeom, Hyun Ju [Kyonggi University, Suwon (Korea, Republic of); Hong, Sang Hee; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of)

    1996-09-01

    A full 3-dimensional code for fast wave heating and the current drive has been developed ant its results are compared with those of FASTWA for Phaedrus-T tokamak. The finite Larmour radius expansion and the order reduction method have been used to derive the wave equation in the toroidal coordinate from the Maxwell-Vlasov equations. By expanding the fields in poloidal Fourier series, the wave equations are reduced to the system of ordinary differential equations in the radial axis, which are then numerically integrated via the shooting method. In addition, the convergence of the solutions and energy conservation are discussed. Finally, and example calculation of the current drive is presented for the advanced superconducting tokamak which is in its conceptual design phase. 17 refs., 10 tabs., 31 figs. (author)

  16. APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Drayer, R.

    2013-06-09

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each

  17. Applications of current technology for continuous monitoring of spent fuel

    International Nuclear Information System (INIS)

    Drayer, R.

    2013-01-01

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each

  18. Electron cyclotron current drive experiments in LHCD plasmas using a remote steering antenna on the TRIAM-1M tokamak

    International Nuclear Information System (INIS)

    Idei, H.; Hanada, K.; Zushi, H.; Ohkubo, K.; Hasegawa, M.; Kubo, S.; Nishi, S.; Fukuyama, A.; Sato, K.N.; Nakamura, K.; Sakamoto, M.; Iyomasa, A.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Notake, T.; Shimozuma, T.; Ito, S.; Hoshika, H.; Maezono, N.; Nakashima, K.; Ogawa, M.

    2006-01-01

    A remote steering antenna was recently developed for electron cyclotron heating and current drive (ECH/ECCD) experiments on the TRIAM-1M tokamak. This is the first application of the remote steering antenna concept for ECH/ECCD experiments, which have conditions relevant to the International Thermonuclear Experimental Reactor (ITER). Fundamental ECH and ECCD experiments were conducted in the ITER frequency from the low field using this antenna system. In addition to the angles near 0 0 , the launcher was a symmetric direction antenna with an extended steering-angle capability of ±(8 0 -19 0 ). The output beam from the antenna was a well-defined Gaussian with a proper steering angle. The Gaussian content and the steering-angle accuracy were 0.85 and -0.5 0 , respectively. The high power tests measured the antenna transmission efficiency at 0.90-0.94. The efficiencies obtained in the low and high power tests were consistent with the calculations using higher-order modes. In order to excite the pure O/X-modes in the oblique injection, two polarizers were used to control the elliptical polarization of the incident beam for the ECCD experiments. The fundamental O/X-mode ECH/ECCD was applied to lower hyrid current drive plasmas at the optimized incident polarization. In the X-mode experiment, at medium density (∼1 x 10 19 m -3 ), clear differences in the plasma current and the hard x-ray intensity were observed between the co- and counter-steering injections due to the ECCD effect on the coupling of forward fast electrons

  19. Gyrokinetic neoclassical study of the bootstrap current in the tokamak edge pedestal with fully non-linear Coulomb collisions

    Energy Technology Data Exchange (ETDEWEB)

    Hager, Robert, E-mail: rhager@pppl.gov; Chang, C. S., E-mail: cschang@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States)

    2016-04-15

    As a follow-up on the drift-kinetic study of the non-local bootstrap current in the steep edge pedestal of tokamak plasma by Koh et al. [Phys. Plasmas 19, 072505 (2012)], a gyrokinetic neoclassical study is performed with gyrokinetic ions and drift-kinetic electrons. Besides the gyrokinetic improvement of ion physics from the drift-kinetic treatment, a fully non-linear Fokker-Planck collision operator—that conserves mass, momentum, and energy—is used instead of Koh et al.'s linearized collision operator in consideration of the possibility that the ion distribution function is non-Maxwellian in the steep pedestal. An inaccuracy in Koh et al.'s result is found in the steep edge pedestal that originated from a small error in the collisional momentum conservation. The present study concludes that (1) the bootstrap current in the steep edge pedestal is generally smaller than what has been predicted from the small banana-width (local) approximation [e.g., Sauter et al., Phys. Plasmas 6, 2834 (1999) and Belli et al., Plasma Phys. Controlled Fusion 50, 095010 (2008)], (2) the plasma flow evaluated from the local approximation can significantly deviate from the non-local results, and (3) the bootstrap current in the edge pedestal, where the passing particle region is small, can be dominantly carried by the trapped particles in a broad trapped boundary layer. A new analytic formula based on numerous gyrokinetic simulations using various magnetic equilibria and plasma profiles with self-consistent Grad-Shafranov solutions is constructed.

  20. Current Continuing Professional Education Practice among Malaysian Nurses

    Directory of Open Access Journals (Sweden)

    Mei Chan Chong

    2014-01-01

    Full Text Available Nurses need to participate in CPE to update their knowledge and increase their competencies. This research was carried out to explore their current practice and the future general needs for CPE. This cross-sectional descriptive study involved registered nurses from government hospitals and health clinics from Peninsular Malaysia. Multistage cluster sampling was used to recruit 1000 nurses from four states of Malaysia. Self-explanatory questionnaires were used to collect the data, which were analyzed using SPSS version 16. Seven hundred and ninety-two nurses participated in this survey. Only 80% (562 of the nurses had engaged in CPE activities during the past 12 months. All attendance for the various activities was below 50%. Workshops were the most popular CPE activity (345, 43.6% and tertiary education was the most unpopular activity (10, 1.3%. The respondents did perceive the importance of future CPE activities for career development. Mandatory continuing professional education (MCPE is a key measure to ensure that nurses upgrade their knowledge and skills; however, it is recommended that policy makers and nurse leaders in the continuing professional development unit of health service facilities plan CPE activities to meet registered nurses’ (RNs needs and not simply organizational requirements.

  1. Current drive studies for the ARIES steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Ehst, D.A.; Mandrekas, J.

    1994-01-01

    Steady-state plasma operating scenarios are designed for three versions of the ARIES reactor, using non-inductive current drive techniques that have an established database. R.f. waves, including fast and lower hybrid waves, are the reference drivers for the D-T burning ARIES-I and ARIES-II/IV, while neutral beam injection is employed for ARIES-III which burns D- 3 He. Plasma equilibria with a high bootstrap-current component have been used, in order to minimize the recirculating power fraction and cost of electricity. To maintain plasma stability, the driven current profile has been aligned with that of equilibrium by proper choices of the plasma profiles and power launch parameters. Except for ARIES-III, the current-drive power requirements and the relevant technology developments are found to be quite reasonable. The wave-power spectrum and launch requirements are also considered achievable with a modest development effort. Issues such as an improved database for fast-wave current drive, lower-hybrid power coupling to the plasma edge, profile control in the plasma core, and access to the design point of operation remain to be addressed. ((orig.))

  2. External kink mode stability of tokamaks with finite edge current density in plasma outside separatrix

    International Nuclear Information System (INIS)

    Degtyarev, L.; Martynov, A.; Medvedev, S.; Troyon, F.; Villard, L.

    1996-01-01

    Large pressure gradients and current density at the plasma edge and accompanying edge-localized MHD instabilities are typical for H-mode discharges. Low-n external kink modes are a possible cause of the instabilities. The paper mostly deals with external kink modes driven by a finite current density at the plasma boundary (so called peeling modes). It was shown earlier that for a single axis plasma embedded into vacuum the peeling modes are stabilized when separatrix is approaching the plasma boundary. For doublet configurations a finite current density at the internal separatrix does not necessarily lead to external kink instability when the current density vanishes at the boundary. However, a finite current density at the plasma boundary outside the separatrix can drive outer peeling modes. The stability properties and structure of these modes depend on the plasma equilibrium outside the separatrix. The influence of plasma shear and pressure gradient at the boundary on the stability of the outer peeling modes in doublets is studied. The stability of kink modes in divertor configurations with plasma outside the separatrix is very sensitive to the boundary conditions set at open field lines. The choice of the boundary conditions and kink mode stability calculations for the divertor configurations are discussed. (author) 4 figs., 5 refs

  3. Kinetic effects on the currents determining the stability of a magnetic island in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Poli, E., E-mail: emanuele.poli@ipp.mpg.de; Bergmann, A.; Casson, F. J.; Hornsby, W. A. [Max-Planck-Institut für Plasmaphysik (Germany); Peeters, A. G. [University of Bayreuth, Department of Physics (Germany); Siccinio, M.; Zarzoso, D. [Max-Planck-Institut für Plasmaphysik (Germany)

    2016-05-15

    The role of the bootstrap and polarization currents for the stability of neoclassical tearing modes is investigated employing both a drift kinetic and a gyrokinetic approach. The adiabatic response of the ions around the island separatrix implies, for island widths below or around the ion thermal banana width, density flattening for islands rotating at the ion diamagnetic frequency, while for islands rotating at the electron diamagnetic frequency the density is unperturbed and the only contribution to the neoclassical drive arises from electron temperature flattening. As for the polarization current, the full inclusion of finite orbit width effects in the calculation of the potential developing in a rotating island leads to a smoothing of the discontinuous derivatives exhibited by the analytic potential on which the polarization term used in the modeling is based. This leads to a reduction of the polarization-current contribution with respect to the analytic estimate, in line with other studies. Other contributions to the perpendicular ion current, related to the response of the particles around the island separatrix, are found to compete or even dominate the polarization-current term for realistic island rotation frequencies.

  4. U.S.-Japan workshop on 'RF heating and current drive in confinement systems tokamaks'

    International Nuclear Information System (INIS)

    1992-01-01

    The workshop was attended by 8 US scientists and 30 Japanese scientists. The agenda was divided into 2 1/2 days of presentation, 1/2 day group discussions and 1/2 day summary session. There were 10 papers on rf physics, technologies and applications; 6 papers on new concepts, helicity injection and transport; and 6 papers on heating/current drive and scrape-off-layer/divertor conditions. The wide range of topics discussed is an indication of the impressive growth, both in depth and breadth, of the US-Japan workshop in RF Heating and Current Drive. It also benefitted by being combined with the new current drive concepts workshops and the active participation of JAERI scientists. (J.P.N.)

  5. Observation of Self-Generated Flows in Tokamak Plasmas with Lower-Hybrid-Driven Current

    International Nuclear Information System (INIS)

    Ince-Cushman, A.; Rice, J. E.; Reinke, M.; Greenwald, M.; Wallace, G.; Parker, R.; Fiore, C.; Hughes, J. W.; Bonoli, P.; Shiraiwa, S.; Hubbard, A.; Wolfe, S.; Hutchinson, I. H.; Marmar, E.; Bitter, M.; Wilson, J.; Hill, K.

    2009-01-01

    In Alcator C-Mod discharges lower hybrid waves have been shown to induce a countercurrent change in toroidal rotation of up to 60 km/s in the central region of the plasma (r/a∼<0.4). This modification of the toroidal rotation profile develops on a time scale comparable to the current redistribution time (∼100 ms) but longer than the energy and momentum confinement times (∼20 ms). A comparison of the co- and countercurrent injected waves indicates that current drive (as opposed to heating) is responsible for the rotation profile modifications. Furthermore, the changes in central rotation velocity induced by lower hybrid current drive (LHCD) are well correlated with changes in normalized internal inductance. The application of LHCD has been shown to generate sheared rotation profiles and a negative increment in the radial electric field profile consistent with a fast electron pinch

  6. Investigation of lower hybrid current drive during H-mode in EAST tokamak

    International Nuclear Information System (INIS)

    Li Miao-Hui; Ding Bo-Jiang; Kong Er-Hua; Zhang Lei; Zhang Xin-Jun; Qian Jin-Ping; Yan Ning; Han Xiao-Feng; Shan Jia-Fang; Liu Fu-Kun; Wang Mao; Xu Han-Dong; Wan Bao-Nian

    2011-01-01

    H-mode discharges with lower hybrid current drive (LHCD) alone are achieved in EAST divertor plasma over a wide parameter range. These H-mode discharges are characterized by a sudden drop in D α emission and a spontaneous rise in main plasma density. Good lower hybrid (LH) coupling during H-mode is obtained by putting the plasma close to the antenna and by injecting D 2 gas from a pipe near the grill mouse. The analysis of lower hybrid current drive properties shows that the LH deposition profile shifts off axis during H-mode, and current drive (CD) efficiency decreases due to the increase in density. Modeling results of H-mode discharges with a general ray tracing code GENRAY are reported. (physics of gases, plasmas, and electric discharges)

  7. The production of high poloidal tokamak equilibria in Versator II by means of RF current drive

    International Nuclear Information System (INIS)

    Luckhardt, S.C.; Chen, K.-I.; Kesner, J.; Kirkwood, R.; Lane, B.; Porkolab, M.; Squire, J.

    1989-01-01

    Experiments on the Versator II device have been carried out in a regime of low plasma current with the aim of reaching high poloidal beta, β p . Lower-Hybrid RF current drive is used to produce an energetic electron population which carries the plasma current and pressure. In this mode of operation, plasmas with εβ p approaching unity appear attainable. Data from equilibrium magnetic analysis, hard x-ray, and density profiles display an outward magnetic axis shift in agreement with equilibrium theory, and further indicate that q(O) is in the range of 4-6. PEST code modeling of these experiments suggests that some of these plasmas may be near or beyond the transition to the second stability region for ballooning modes. (author)

  8. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  9. Confinement bifurcation by current density profile perturbation in TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.

    2001-01-01

    In the recent experiments performed on TUMAN-3M the possibility to switch on/off the H-mode by current density profile perturbations has been shown. The j(r) perturbations were created by fast Current Ramp Up/Down or by Magnetic Compression produced by a fast increase of the toroidal magnetic field. It was found that the Current Ramp Up (CRU) and Magnetic Compression (MC) are useful means for H-mode triggering. The Current Ramp Down (CRD) triggers H-L transition. The difference in the j(r) behavior in these experiments suggests the peripheral current density may not be the critical parameter controlling L-H and H-L transitions. Confinement bifurcation in the above experiments could be explained by the unified mechanism: variation of a turbulent transport resulting from radial electric field emerging near the edge in the conditions of alternating toroidal electric field Ej and different electron and ion collisionalities. According to the suggested model the toroidal field E φ arising in the periphery during the CRU and MC processes amplifies Ware drift, which mainly influences electron component. As a result the favorable for the transition negative (inward directed) E r emerges. In the CRD scenario, when E φ is opposite to the total plasma current direction, the mechanism should generate positive E r , which is thought to be unfavorable for the H-mode. The experimental data on L-H and H-L transitions in various scenarios and the results of the modeling of E r emerging in the CRU experiment are presented in the paper. (author)

  10. Current generation by alpha particles interacting with lower hybrid waves in TOKAMAKS

    International Nuclear Information System (INIS)

    Belikov, V.S.; Kolesnichenko, Ya.I.; Lisak, M.; Anderson, D.

    1990-01-01

    The problem of the influence of fusion generated alpha particles on lower-hybrid-wave current drive is examined. Analysis is based on a new equation for the LH-wave-fast ion interaction which is derived by taking into consideration the non-zero value of the longitudinal wave number. The steady-state velocity distribution function for high energy alpha particles is found. The alpha current driven by LH-waves as well as the RF-power absorbed by alpha particle are calculated. (authors)

  11. 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

  12. Identification of anomalous Doppler resonance effect during current ramp down in HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li Erzhong, E-mail: rzhonglee@ipp.ac.c [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Hu Liqun; Ling Bili; Liu Yong; Ti Ang; Zhou Reijie; Lu Hongwei; Gao Xiang [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China)

    2010-09-21

    The abrupt steep jump of electron cyclotron emission (ECE) signals during current ramp-down has been observed and explained by an anomalous Doppler resonance effect (ADR). The identifying process of ADR was presented based on the fast Fourier transform (FFT) technique. The threshold value for triggering a steep jump on ECE signals has been identified under different discharge conditions.

  13. Study on lower hybrid current drive efficiency at high density towards long-pulse regimes in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Li, M. H.; Ding, B. J.; Zhang, J. Z.; Gan, K. F.; Wang, H. Q.; Zhang, L.; Wei, W.; Li, Y. C.; Wu, Z. G.; Ma, W. D.; Jia, H.; Chen, M.; Yang, Y.; Feng, J. Q.; Wang, M.; Xu, H. D.; Shan, J. F.; Liu, F. K.; Peysson, Y.

    2014-01-01

    Significant progress on both L- and H-mode long-pulse discharges has been made recently in Experimental Advanced Superconducting Tokamak (EAST) with lower hybrid current drive (LHCD) [J. Li et al., Nature Phys. 9, 817 (2013) And B. N. Wan et al., Nucl. Fusion 53, 104006 (2013).]. In this paper, LHCD experiments at high density in L-mode plasmas have been investigated in order to explore possible methods of improving current drive (CD) efficiency, thus to extend the operational space in long-pulse and high performance plasma regime. It is observed that the normalized bremsstrahlung emission falls much more steeply than 1/n e-av (line-averaged density) above n e-av  = 2.2 × 10 19  m −3 indicating anomalous loss of CD efficiency. A large broadening of the operating line frequency (f = 2.45 GHz), measured by a radio frequency (RF) probe located outside the EAST vacuum vessel, is generally observed during high density cases, which is found to be one of the physical mechanisms resulting in the unfavorable CD efficiency. Collisional absorption of lower hybrid wave in the scrape off layer (SOL) may be another cause, but this assertion needs more experimental evidence and numerical analysis. It is found that plasmas with strong lithiation can improve CD efficiency largely, which should be benefited from the changes of edge parameters. In addition, several possible methods are proposed to recover good efficiency in future experiments for EAST

  14. Fluctuation measurements by Langmuir probes during LHCD on ASDEX tokamak. [LHCD (Lower Hybrid Current Drive)

    Energy Technology Data Exchange (ETDEWEB)

    Stoeckel, J [Ceskoslovenska Akademie Ved, Prague (Czech Republic). Ustav Fyziky Plazmatu; Soeldner, F; Giannone, L.; Leuterer, F [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1991-01-01

    The level of edge electrostatic fluctuations decreases and the global particle/energy confinement improves during lower hybrid current drive (LHCD) regimes on ASDEX, when the total power remains below the initial OH power level. For higher powers, the fluctuations increase noticeably, whereas the global confinement is returning to its OH value. The observed increase of fluctuations is poloidally asymmetric and is caused by local power deposition in front of the grill antenna. (author) 5 refs., 4 figs.

  15. Plasma current start-up experiments without the central solenoid in the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Ejiri, A.; Shiraiwa, S.; Adachi, Y.; Ishii, N.; Kasahara, H.; Nuga, H.; Ono, Y.; Oosako, T.; Sasaki, M.; Shimada, Y.; Sumitomo, N.; Taguchi, I.; Tojo, H.; Tsujimura, J.; Ushigome, M.; Yamada, T.; Hanada, K.; Hasegawa, M.; Idei, H.; Nakamura, K.; Sakamoto, M.; Sasaki, K.; Sato, K.N.; Zushi, H.; Nishino, N.; Mitarai, O.

    2006-01-01

    Several techniques for initiating the plasma current without the use of the central solenoid are being developed in TST-2. While TST-2 was temporarily located at Kyushu University, two types of start-up scenarios were demonstrated. (1) A plasma current of 4 kA was generated and sustained for 0.28 s by either electron cyclotron wave or electron Bernstein wave, without induction. (2) A plasma current of 10 kA was obtained transiently by induction using only outboard poloidal field coils. In the second scenario, it is important to supply sufficient power for ionization (100 kW of EC power was sufficient in this case), since the vertical field during start-up is not adequate to maintain plasma equilibrium. In addition, electron heating experiments using the X-B mode conversion scenario were performed, and a heating efficiency of 60% was observed at a 100 kW RF power level. TST-2 is now located at the Kashiwa Campus of the University of Tokyo. Significant upgrades were made in both magnetic coil power supplies and RF systems, and plasma experiments have restarted. RF power of up to 400 kW is available in the high-harmonic fast wave frequency range around 20 MHz. Four 200 MHz transmitters are now being prepared for plasma current start-up experiments using RF power in the lower-hybrid frequency range. Preparations are in progress for a new plasma merging experiment (UTST) aimed at the formation and sustainment of ultra-high β ST plasmas

  16. Assessment of eddy current effects on compression experiments in the TFTR tokamak

    International Nuclear Information System (INIS)

    Wong, K.L.; Park, W.

    1986-05-01

    The eddy current induced on the TFTR vacuum vessel during compression experiments is estimated based on a cylindrical model. It produces an error magnetic field that generates magnetic islands at the rational magnetic surfaces. The widths of these islands are calculated and found to have some effect on electron energy confinement. However, resistive MHD simulation results indicate that the island formation process can be slowed down by plasma rotation

  17. Adjoint optimization scheme for lower hybrid current rampup and profile control in Tokamak

    International Nuclear Information System (INIS)

    Litaudon, X.; Moreau, D.; Bizarro, J.P.; Hoang, G.T.; Kupfer, K.; Peysson, Y.; Shkarofsky, I.P.; Bonoli, P.

    1992-12-01

    The purpose of this work is to take into account and study the effect of the electric field profiles on the Lower Hybrid (LH) current drive efficiency during transient phases such as rampup. As a complement to the full ray-tracing / Fokker Planck studies, and for the purpose of optimization studies, we developed a simplified 1-D model based on the adjoint Karney-Fisch numerical results. This approach allows us to estimate the LH power deposition profile which would be required for ramping the current with prescribed rate, total current density profile (q-profile) and surface loop voltage. For rampup optimization studies, we can therefore scan the whole parameter space and eliminate a posteriori those scenarios which correspond to unrealistic deposition profiles. We thus obtain the time evolution of the LH power, minor radius of the plasma, volt-second consumption and total energy dissipated. Optimization can thus be performed with respect to any of those criteria. This scheme is illustrated by some numerical simulations performed with TORE-SUPRA and NET/ITER parameters. We conclude with a derivation of a simple and general scaling law for the flux consumption during the rampup phase

  18. Current state of continuous ambulatory peritoneal dialysis in Egypt

    Directory of Open Access Journals (Sweden)

    Khaled Mohamed Amin Elzorkany

    2017-01-01

    Full Text Available Patients with end-stage renal disease (ESRD continue to increase in number worldwide, especially in developing countries. Although continuous ambulatory peritoneal dialysis (CAPD has comparable survival advantages as hemodialysis (HD, it is greatly underutilized in many regions worldwide. The prevalence of use of CAPD in Egypt is 0.29/million population in 2017. The aim of this study is to describe the current state and practice of CAPD in Egypt and included 22 adult patients who were treated by CAPD. All the study patients were switched to CAPD after treatment with HD failed due to vascular access problems. Patients were mainly female (68.2 % with the mean age of 49.77 ± 11.41 years. The average duration on CAPD was 1.76 ± 1.30 years. Hypertension was the main cause of end-stage renal disease (ESRD constituting 36.4%, followed by diabetes (27.3 %, and toxic nephropathy (4.5%. Of importance is that about 31.8% of patients had ESRD of unknown etiology. The mean weekly Kt/V urea of patients on PD was 1.92 ± 0.18. The mean hemoglobin, serum calcium, phosphorus, parathormone, and albumin levels were 10.27 ± 1.98 g/dL, 8.36 ± 1.19 mg/dL, 5.70 ± 1.35 mg/dL, 541.18 ± 230.12 pg/mL, and 2.98 ± 0.73 g/dL, respectively. There was no significant difference between diabetic and nondiabetic CAPD patients regarding demographic and laboratory data. Our data indicate that there is continuing underutilization of CAPD in Egypt which may be related to nonavailability of CAPD fluid, patient factors (education and motivation, gradual decline of the efficiency of health-care professionals, and lack of a national program to start PD as the first modality for renal replacement therapy. It is advised to start an organized program to make CAPD widespread and encourage local production of PD fluids to reduce the cost of CAPD.

  19. Beat wave current drive experiment on the Davis Diverted Tokamak (DDT)

    International Nuclear Information System (INIS)

    Hwang, D.Q.; Horton, R.D.; Rogers, J.H.

    1993-01-01

    The beatwave current drive experiment is summarized. The first phase of the experiment was the construction of the microwave sources and the diagnostics needed to demonstrate the beat wave effects, i.e. the measurement of the electrostatic plasma wave produced by the beating of two high intensity electromagnetic waves. In order to keep the cost of the experiments to a minimum, a low density filament plasma source (10 8 ) to (10 10 particles cm -3 ) was employed and the magnetic field in the toroidal plasma was produced by a dc power supply

  20. Current drive in a tokamak reactor during the heating of fast α particles

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Soboleva, T.K.

    1987-01-01

    Expressions are derived for the efficiency of the current drive in the approximation of a straight magnetic field through a solution of the kinetic equation for the distribution function of α particles as they are heated by rf waves. Three mechanisms for the absorption of the rf power in plasma are examined: cyclotron absorption at the fundamental frequency, Landau damping, and magnetic Landau damping. The efficiency of this method is shown to be at worst no lower than the efficiencies of methods involving electron heating

  1. Development of hybrid frequency couplers for non-inductive current drive in a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, St.

    1996-11-04

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead of the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a non-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (author). 53 refs.

  2. Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)

    1996-12-31

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.

  3. Tail anisotropy instability during plasma current rise by lower-hybrid waves in a tokamak

    International Nuclear Information System (INIS)

    Yamagiwa, Mitsuru.

    1986-01-01

    Tail anisotropy instability during lower-hybrid current rise is investigated. Tail formation by lower-hybrid waves is studied by using a Fokker-Planck equation combined with the return field and the rf associated terms. Quasi-linear relaxation of the electron tail distribution under the influence of the plasma waves excited due to the instability is examined. It is found that the instability condition is related to the strength of the parallel diffusion by lower-hybrid waves and the ratio of the electron cyclotron frequency to the electron plasma frequency. The time scale between the instability spikes and the suppression of the instability by electron cyclotron heating are also discussed. (author)

  4. Full-wave calculation of fast-wave current drive in tokamaks including kparallel upshifts

    International Nuclear Information System (INIS)

    Jaeger, E.F.; Batchelor, D.B.

    1991-01-01

    Numerical calculations of fast-wave current drive (FWCD) efficiency have generally been of two types: ray tracing or global wave calculations. Ray tracing shows that the projection of the wave number (k parallel) along the magnetic field can vary greatly over a ray trajectory, particularly when the launch point is above or below the equatorial plane. As the wave penetrates toward the center of the plasma, k parallel increases, causing a decrease in the parallel phase speed and a corresponding decrease in the current drive efficiency, γ. But the assumptions of geometrical optics, namely short wavelength and strong single-pass absorption, are not greatly applicable in FWCD scenarios. Eigenmode structure, which is ignored in ray tracing, can play an important role in determining electric field strength and Landau damping rates. In such cases, a full-wave or global solution for the wave fields is desirable. In full-wave calculations such as ORION k parallel appear as a differential operator (rvec B·∇) in the argument of the plasma dispersion function. Since this leads to a differential system of infinite order, such codes of necessity assume k parallel ∼ k var-phi = const, where k var-phi is the toroidal wave number. Thus, it is not possible to correctly include effects of the poloidal magnetic field on k parallel. The problem can be alleviated by expressing the electric field as a superposition of poloidal modes, in which case k parallel is purely algebraic. This paper describes a new full-wave calculation, Poloidal Ion Cyclotron Expansion Solution, which uses poloidal and toroidal mode expansions to solve the wave equation in general flux coordinates. The calculation includes a full solution for E parallel and uses a reduced-order form of the plasma conductivity tensor to eliminate numerical problems associated with resolution of the very short wavelength ion Bernstein wave

  5. A simultaneous description of fast wave e-TTMP and ion current drive effects on shear in a tokamak: theory and experiments in JET

    International Nuclear Information System (INIS)

    Bhatnagar, V.P.; Bosia, G.; Jacquinot, J.; Porcelli, F.

    1993-01-01

    A controlled local modification of the plasma-current profile, the safety factor q or shear (dq/dr) in a tokamak can lead to an improvement in its performance. For example, enhanced confinement in JET discharges with deep pellet injection is found to be associated with a reversal of the shear. Also, a significant control over the sawteeth behaviour in the JET tokamak has been found to occur when the shear at the q = 1 surface is modified by a dipolar-current driven by ICRF in the minority-ion heating regime. This could give a handle on the ejection of fast particles and hence on burn control in a reactor. The above sawtooth control may also be used to ease the ash removal in a reactor. When an ICRH antenna array is phased (Δφ ≠ 0 or π), the excited asymmetric k // -spectrum can drive non inductive currents by interaction of waves both with electrons (TTMP and e-Landau damping) and ions at minority (fundamental) or harmonic cyclotron resonances depending upon the scenario. Therefore, in any modeling of ICRF current drive, both (electron and ion) current drive mechanisms must be included simultaneously to correctly represent the non inductive current drive profile. To devise scenarios of shear control by minority current drive, that take advantage of the inherent electron current drive as well, we have developed a model based on earlier theories to calculate, for the first time, the two effects simultaneously. (author) 11 refs., 5 figs

  6. Optimization of OH coil recharging scenario of quasi-steady operation in tokamak fusion reactor by lower hybrid wave current drive

    International Nuclear Information System (INIS)

    Sugihara, M.; Fujisawa, N.; Nishio, S.; Iida, H.

    1984-01-01

    Using simple physical model equations optimum plasma and rf parameters for an OH coil recharging scenario of quasi-steady operation in tokamak fusion reactors by lower hybrid wave current drive are studied. In this operation scenario, the minimization of the recharge time of OH coils or stored energy for it will be essential and can be realized by driving sufficient current without increasing the plasma temperature too much. Low density and broad spectrum are shown to be favorable for the minimization. In the case of FER (Fusion Experimental Reactor under design study in JAERI) baseline parameters, the minimum recharge time is 3-5 s/V s. (orig.)

  7. Social Work Continuing Education: Current Issues and Future Direction

    Science.gov (United States)

    Kurzman, Paul A.

    2016-01-01

    Continuing education is arising as an area of rapid growth and increased attention in the social work profession. Conceptually, the impetus and focus are on the promotion of the principles of lifelong learning and professional replenishment; but pragmatically, the driving force has been the virtually universal requirement of continuing education…

  8. Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I; Calculo sobre una modificacion al campo magnetico toroidal del Tokamak Novillo. Parte I

    Energy Technology Data Exchange (ETDEWEB)

    Chavez A, E.; Melendez L, L.; Colunga S, S.; Valencia A, R.; Lopez C, R.; Gaytan G, E

    1991-07-15

    The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

  9. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  10. Resistive evolution of current profile in tokamaks, application to the optimization of Tore-supra plasma discharges; Evolution resistive du profil de courant dans les Tokamaks, application a l'optimisation des decharges de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Bregeon, R

    1999-03-01

    In Tokamak plasma physics, current profile shaping has now become a key issue to improve the confinement properties of the plasma discharge. The objective of this work is to study the processes governing the current diffusion when non-inductive current are playing a major role in the discharge. Ultimately, this study aims to identify the key parameters to control the plasma current density profile with external current drive heating systems such as Lower Hybrid Current drive (LHCD) or self generated current drive such as the bootstrap current. Principles of non inductive current drive and heating systems are introduced as well as bootstrap current mechanisms. Then we present the experimental study of plasma parallel electric conductivity to validate existing models. Using these results, the poloidal magnetic field flux diffusion is modelled, using toroidal co-ordinates in order to give an accurate description of the current density profiles evolution. The initial and boundary conditions required for numerical resolution of the diffusion equation are also presented. Finally, we conclude this work with the simulations of two discharges: one with Fast Wave Electron Heating and the second using Lower Hybrid Current Drive. These simulations have multiples aims: validity test of our numerical tool and to show some limits of cylindrical models. Test of electric conductivity and bootstrap current models. To identify the key parameters involved in the current diffusion processes of a high performance plasma discharge on Tore Supra. Such simulations are crucial to determine the amount of non-inductive current required to control and sustain long plasma discharges in steady state. (author)

  11. Observation of bulk-ion heating in a tokamak plasma by application of positive and negative current pulses in TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-09-01

    A positive of negative current pulse induced by a pulsed toroidal electric field much higher than the Dreicer field increases the bulk-ion temperature of the plasma centre two to three times, without macroscopic plasma destruction. The decay time of the raised ion temperature agrees well with the prediction from neoclassical transport theory. The magnitude of the positive current pulse is limited by violent current disruption, and that of the negative current by a lack of MHD equilibrium which is due to a marked reduction of the total plasma current. The relevant current-driven instabilities in the turbulent heating of a tokamak plasma, skin heating and inward transfer of the energy deposition in the skin layer are briefly discussed.

  12. 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

  13. On self-consistent ray-tracing and Fokker-Planck modeling of the hard X-ray emission during lower-hybrid current driven in Tokamaks

    International Nuclear Information System (INIS)

    Bizarro, J.P.; Peysson, Y.; Bonoli, P.T.; Carrasco, J.; Dudok de Wit, T.; Fuchs, V.; Hoang, G.T.; Litaudon, X.; Moreau, D.; Pocheau, C.; Shkarofsky, I.P.

    1993-04-01

    A detailed investigation is presented on the ability of combined ray-tracing and Fokker-Planck calculations to predict the hard x-ray (HXR) emission during lower-hybrid (LH) current drive in tokamaks when toroidally induced-ray-stochasticity is important. A large number of rays is used and the electron distribution function is obtained by self-consistently iterating the appropriate LH power deposition and Fokker-Planck calculations. Most of the experimentally observed features of the HXR emission are correctly predicted. It is found that corrections due to radial diffusion of suprathermal electrons and to radiation scattering by the inner wall can be significant

  14. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

    1992-01-01

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

  15. Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

    International Nuclear Information System (INIS)

    Kaita, R.; Bernabei, S.; Budny, R.

    1996-01-01

    The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

  16. A programmatic framework for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

    1993-01-01

    Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

  17. Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I

    International Nuclear Information System (INIS)

    Chavez A, E.; Melendez L, L.; Colunga S, S.; Valencia A, R.; Lopez C, R.; Gaytan G, E.

    1991-07-01

    The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

  18. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  19. Spectral dependence, efficiency and localization of non-inductive current drive via helicity injection by global Alfven waves in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Komoshvili, K.; Cuperman, S.; Bruma, C. [Tel Aviv Univ. (Israel). Sackler Faculty of Exact Sciences

    1997-04-01

    A systematic study of non-inductive current drive via helicity injection by global Alfven eigenmode (GAE) waves is carried out. For illustration, the first radial mode of the discrete resonant GAE spectrum is considered. The following aspects are given special attention: spectral analysis, radial dependence and efficiency - all of these functions of the characteristics of the waves launched by an external, concentric antenna (i.e. wave frequency and poloidal and toroidal wavenumbers). The tokamak plasma is simulated by a current-carrying cylindrical plasma column surrounded by a helical sheet current and situated inside a perfectly conducting shell, with incorporation of equilibrium (simulated) toroidal field, magnetic shear and a relatively large poloidal magnetic field component. Within the framework of low-{beta} MHD model equations and for typical tokamak physical parameters, the following basic results are obtained: (1) in the range of poloidal wavenumbers -3{<=} m {<=} 3 and toroidal wavenumbers -20{<=} n {<=}20, resonant GAE peaks below the Alfven continuum are found; (2) the power absorption (P), current drive (I) and corresponding frequency of the GAE modes depend strongly on the sets of (m,n) values considered; (3) the `net` current drive is positive (i.e. flows in the direction of the equilibrium current j{sub 0z} for m = -1, -2, -3 and -20 {<=} n {<=} -1 as well as for m +1, +2, +3 and n > 10); (4) in the cases m = -1, -2, -3, the efficiency of current drive, I/P, increases with /m/ and I/n/; (5) the radial localization of the current drive in each of the cases considered is determined and tabulated. (Author).

  20. Spectral dependence, efficiency and localization of non-inductive current drive via helicity injection by global Alfven waves in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1997-01-01

    A systematic study of non-inductive current drive via helicity injection by global Alfven eigenmode (GAE) waves is carried out. For illustration, the first radial mode of the discrete resonant GAE spectrum is considered. The following aspects are given special attention: spectral analysis, radial dependence and efficiency - all of these functions of the characteristics of the waves launched by an external, concentric antenna (i.e. wave frequency and poloidal and toroidal wavenumbers). The tokamak plasma is simulated by a current-carrying cylindrical plasma column surrounded by a helical sheet current and situated inside a perfectly conducting shell, with incorporation of equilibrium (simulated) toroidal field, magnetic shear and a relatively large poloidal magnetic field component. Within the framework of low-β MHD model equations and for typical tokamak physical parameters, the following basic results are obtained: (1) in the range of poloidal wavenumbers -3≤ m ≤ 3 and toroidal wavenumbers -20≤ n ≤20, resonant GAE peaks below the Alfven continuum are found; (2) the power absorption (P), current drive (I) and corresponding frequency of the GAE modes depend strongly on the sets of (m,n) values considered; (3) the 'net' current drive is positive (i.e. flows in the direction of the equilibrium current j 0z for m = -1, -2, -3 and -20 ≤ n ≤ -1 as well as for m +1, +2, +3 and n > 10; (4) in the cases m = -1, -2, -3, the efficiency of current drive, I/P, increases with /m/ and I/n/; (5) the radial localization of the current drive in each of the cases considered is determined and tabulated. (Author)

  1. Quantization of edge currents for continuous magnetic operators

    CERN Document Server

    Kellendonk, J

    2003-01-01

    For a magnetic Hamiltonian on a half-plane given as the sum of the Landau operator with Dirichlet boundary conditions and a random potential, a quantization theorem for the edge currents is proven. This shows that the concept of edge channels also makes sense in presence of disorder. Moreover, gaussian bounds on the heat kernel and its covariant derivatives are obtained.

  2. Electromagnetic torques and forces due to misalignment effects and eddy currents in the homopolar generator, power supply for the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Driga, M.D.; Bird, W.L.; Tolk, K.M.; Weldon, W.F.; Rylander, H.G.; Woodson, H.H.

    1977-01-01

    Asymmetries in the applied magnetic field due to manufacturing tolerances and rotor-stator misalignments can cause significant forces and moments in a homopolar generator. Parasitic eddy-currents in the rotor, brushes and bearings are also important effects of such asymmetries. The finite element method is used to calculate the magnetic flux distributions in the TEXT homopolar generators. The axial magnetic thrust force and the magnetic tilt moment acting on the rotor are calculated. Eddy-current torques opposing rotor motion are determined using the theory for eddy-current brakes. The results have been used in the design of the TEXT homopolar generator which have been proposed to provide the energy store and conversion for the toroidal field and ohmic heating coils of the new Texas Experimental Tokamak

  3. Nasal continuous positive airway pressure treatment: current realities and future.

    Science.gov (United States)

    Berthon-Jones, M; Lawrence, S; Sullivan, C E; Grunstein, R

    1996-11-01

    Nasal continuous positive airway pressure (CPAP) is a highly effective treatment for obstructive sleep apnea syndrome. The apnea/hypopnea index (AHI) is reduced 10-fold, but the patient dropout rate is up to 30%, and usage is typically 20/hour were recruited, with written informed consent. Subjects slept for a diagnostic night, followed by a treatment night, in the laboratory, using the AutoSet system with full polysomnographic monitoring of respiratory and sleep variables. Arousals were scored using ASDA criteria. Hypopneas were scored when there was a 50% reduction in ventilation for > 10 seconds, associated with a 4% drop in oxygen saturation. For comparison, the ASDA arousal index in 16 normal subjects (without nasal CPAP) is provided. Results are given as mean +/- standard error of the mean. AHI was reduced from 55 +/- 3 to 1.5 +/- 0.35 events/hour (p < 0.0001). The arousal index was reduced from 65 +/- 3 to 18 +/- 2 events/hour (p < 0.0001), identical to the value in the 16 healthy normal subjects. There was a 158% +/- 21% increase in slow-wave sleep (p = 0.01) and a 186% +/- 27% increase in rapid eye movement sleep (p = 0.013). The AutoSet self-adjusting nasal CPAP system adequately treats obstructive sleep apnea syndrome on the first night under laboratory conditions.

  4. 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

  5. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  6. Dependence of synergy current driven by lower hybrid wave and electron cyclotron wave on the frequency and parallel refractive index of electron cyclotron wave for Tokamaks

    International Nuclear Information System (INIS)

    Huang, J.; Chen, S. Y.; Tang, C. J.

    2014-01-01

    The physical mechanism of the synergy current driven by lower hybrid wave (LHW) and electron cyclotron wave (ECW) in tokamaks is investigated using theoretical analysis and simulation methods in the present paper. Research shows that the synergy relationship between the two waves in velocity space strongly depends on the frequency ω and parallel refractive index N // of ECW. For a given spectrum of LHW, the parameter range of ECW, in which the synergy current exists, can be predicted by theoretical analysis, and these results are consistent with the simulation results. It is shown that the synergy effect is mainly caused by the electrons accelerated by both ECW and LHW, and the acceleration of these electrons requires that there is overlap of the resonance regions of the two waves in velocity space

  7. 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

  8. Current-drive and plasma formation experiments on the Versator-II tokamak using lower-hybrid and electron-cyclotron waves

    International Nuclear Information System (INIS)

    Colborn, J.A.

    1992-01-01

    During lower-hybrid current-driven (LHCD) tokamak discharges with thermal electron temperature T e ∼ 150 eV, a two-parallel-temperature tail is observed in the electron distribution function. The cold tail extends to parallel energy E parallel ∼ 4.5 keV with temperature T cold tail ∼ 1.5 keV, and the hot tail extends to E parallel > 150 keV with T hot tail > 40 keV. Fokker-Planck computer simulations suggest the cold tail is created by low power, high-N parallel sidelobes in the lower-hybrid antenna spectrum, and that these sidelobes bridge the spectral gap, enabling current drive on small tokamaks such as Versator. During plasma-formation experiments using 28 GHz electroncyclotron (EC) waves, the plasma is born near the EC layer, then moves toward the upper-hybrid (UH) layer within 100-200μs. Wave power is detected in the plasma with frequency f = 300 MHz. Measured turbulent plasma fluctuations are correlated with decay-wave amplitude. Electron-cyclotron current-drive (ECCD) is observed with loop voltage V loop ≤ 0 and fully sustained plasma current I p approx-lt 15 kA at densities up to [n e ] = 2 x 10 12 cm -3 . The efficiency falls rapidly to zero as the density is raised, suggesting the ECCD depends on low collisonality. The EC waves enhance magnetic turbulence in the frequency range 50 kHz approx-lt f approx-lt 400 kHz by up to an order of magnitude. The time-of-arrival of the turbulence to probes at the plasma boundary is longer when the EC layer is farther from the probes

  9. Physics of the interaction between runaway electrons and the background plasma of the current quench in tokamak disruptions

    Science.gov (United States)

    Reux, Cedric

    2017-10-01

    Runaway electrons are created during disruptions of tokamak plasmas. They can be accelerated in the form of a multi-MA beam at energies up to several 10's of MeV. Prevention or suppression of runaway electrons during disruptions will be essential to ensure a reliable operation of future tokamaks such as ITER. Recent experiments showed that the suppression of an already accelerated beam with massive gas injection was unsuccessful at JET, conversely to smaller tokamaks. This was attributed to a dense, cold background plasma (up to several 1020 m-3 accompanying the runaway beam. The present contribution reports on the latest experimental results obtained at JET showing that some mitigation efficiency can be restored by changing the features of the background plasma. The density, temperature, position of the plasma and the energy of runaways were characterized using a combined analysis of interferometry, soft X-rays, bolometry, magnetics and hard X-rays. It showed that lower density background plasmas were obtained using smaller amounts of gas to trigger the disruption, leading to an improved penetration of the mitigation gas. Based on the observations, a physical model of the creation of the background plasma and its subsequent evolution is proposed. The plasma characteristics during later stages of the disruption are indeed dependent on the way it was initially created. The sustainment of the plasma during the runaway beam phase is then addressed by making a power balance between ohmic heating, power transfer from runaway electrons, radiation and atomic processes. Finally, a model of the interaction of the plasma with the mitigation gas is proposed to explain why massive gas injection of runaway beams works only in specific situations. This aims at pointing out which parameters bear the most importance if this mitigation scheme is to be used on larger devices like ITER. Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium

  10. 30 CFR 77.801-1 - Grounding resistors; continuous current rating.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Grounding resistors; continuous current rating... OF UNDERGROUND COAL MINES Surface High-Voltage Distribution § 77.801-1 Grounding resistors; continuous current rating. The ground fault current rating of grounding resistors shall meet the “extended...

  11. 30 CFR 77.901-1 - Grounding resistor; continuous current rating.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Grounding resistor; continuous current rating. 77.901-1 Section 77.901-1 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF... resistor; continuous current rating. The ground fault current rating of grounding resistors shall meet the...

  12. Application and Continued Development of Thin Faraday Collectors as a Lost Ion Diagnostic for Tokamak Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    F. Ed Cecil

    2011-06-30

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  13. Generation of runaway electrons during deterioration of lower hybrid power coupling in lower hybrid current drive plasmas in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Chen, Z Y; Ju, H J; Zhu, J X; Li, M; Cai, W D; Liang, H F; Wan, B N; Shi, Y J; Xu, H D

    2009-01-01

    Efficient coupling of lower hybrid (LH) power from the wave launcher to the plasma is a very important issue in lower hybrid current drive (LHCD) experiments. The large unbalanced reflections in the grill trigger the LH protection system, which will trip the power, resulting in the reduction of the coupled LH power. The generation of runaway electrons has been investigated in LHCD plasmas with deterioration of LH coupling in the HT-7 tokamak. The deterioration of LH coupling results in an increase of the loop voltage and a more energetic fast electron population. These two effects favor the generation of a runaway population. It is found that most of the fast electrons generated by LH waves through parallel electron Landau damping were converted into a runaway population through the acceleration from the toroidal electric field when significant deterioration of LH coupling occurs.

  14. Fast ion confinement during high power tangential neutral beam injection into low plasma current discharges on the ISX-B tokamak

    International Nuclear Information System (INIS)

    Carnevali, A.; Scott, S.D.; Neilson, H.; Galloway, M.; Stevens, P.; Thomas, C.E.

    1988-01-01

    The beam ion thermalization process during tangential neutral beam injection in the ISX-B tokamak is investigated. The classical model is tested in co- and counter-injected discharges at low plasma current, a regime where large orbit width excursions enhance the importance of the loss regions. To test the model, experimental charge exchange spectra are compared with the predictions of an orbit following Monte Carlo code. Measurements of beam-plasma neutron emission and measured decay rates of the emission following beam turnoff provide additional information. Good agreement is found between theory and experiment. Furthermore, beam additivity experiments show that, globally, the confinement of beam ions remains classical, independently of the injected beam power. However, some experimental evidence suggests that the fast ion density in the plasma core did not increase with beam power in a way consistent with classical processes. (author). 35 refs, 17 figs, 3 tabs

  15. A study on the continuing education of radiologic technologists: Focused on current status and satisfaction of continuing education

    International Nuclear Information System (INIS)

    Min, Hye Lim; Choi, In Seok; Nam, So Ra; Kim, Hyun Ji; Yoon, Yong Su; Her, Jae; Han, Seong Gyu; Kim, Jung Min; Ahn, Duck Sun

    2014-01-01

    In this study, we surveyed the current status, satisfaction and demand of radiologic technologist continuing education for 93 radiologic technologists who participated in the continuing education. To understand the current status and general evaluation and to find out the improvement direction, survey was conducted on 3 categories: participation, satisfaction and demand of continuing education. In addition, we analyzed the continuing education implementation status and the management system by collecting related regulations. As a result, the education completion rates of radiologic technologists from 2010 to 2012 were respectively 42.6%, 43.4% and 34.2%; the rates were similar to other medical technician’s average education completion rates. According to the survey, in case of participation, the most frequent answer was ‘more than five times less than 10 times per year’ with 48.4% and in satisfaction section, the most common answer was ‘Average(3)’ with 34.4%. In demand of continuing education section, 32.8% of the respondents chose ‘Clinical skill training in major field’. In the results of this research, continuing education needs to be managed in the direction of helping radiologists improve their personal ability and self development. Furthermore, to meet the demand of radiologists, the quality of continuing education should be improved to satisfy the educatee

  16. X-ray measurements during plasma current start-up experiments using the lower hybrid wave on the TST-2 spherical tokamak

    International Nuclear Information System (INIS)

    Wakatsuki, Takuma; Ejiri, Akira; Kakuda, Hidetoshi

    2012-01-01

    Non-inductive plasma current start-up experiments using RF power in the lower hybrid frequency range is being conducted on the TST-2 spherical tokamak. Plasma currents of up to 15 kA have been achieved. The effect of direct current drive can be seen by comparing the cases with co-drive and counter-drive. X-rays in various energy ranges were measured to investigate the interaction between the wave and the electrons. Soft X-ray (SX) measurements revealed that the perpendicular SX emission increased significantly as the plasma current increased, and that the tangential SX emission in the direction of RF drive was enhanced more strongly in the co-drive case compared to the counter-drive case. These observations imply that the fast electrons accelerated by the lower hybrid wave contribute to the plasma current. However, RF amplitude modulation experiments showed that the confinement time of these fast electrons are very short (less than 0.05 ms), much shorter than the collisional slowing down time. Hard X-ray spectral measurements showed that the radiation temperature of fast electrons in the co-direction for current drive was higher than that in the counter-direction. These observations are consistent with the existence of RF-driven fast electrons. (author)

  17. 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)

  18. 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

  19. Observation of a new turbulence-driven limit-cycle state in H-modes with lower hybrid current drive and lithium-wall conditioning in the EAST superconducting tokamak

    DEFF Research Database (Denmark)

    Wang, H.Q.; Xu, G.S.; Guo, H.Y.

    2012-01-01

    The first high confinement H-mode plasma has been obtained in the Experimental Advanced Superconducting Tokamak (EAST) with about 1 MW lower hybrid current drive after wall conditioning by lithium evaporation and real-time injection of Li powder. Following the L–H transition, a small-amplitude, low...

  20. Theory of current-drive in plasmas

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1986-12-01

    The continuous operation of a tokamak fusion reactor requires, among other things, a means of providing continuous toroidal current. Such operation is preferred to the conventional pulsed operation, where the plasma current is induced by a time-varying magnetic field. A variety of methods has been proposed to provide continuous current, including methods which utilize particle beams or radio frequency waves in any of several frequency regimes. Currents as large as half a mega-amp have now been produced in the laboratory by such means, and experimentation in these techniques has now involved major tokamak facilities worldwide

  1. Continuous improvement in the Netherlands: A survey-based study into the current practices of continuous improvement

    NARCIS (Netherlands)

    Middel, H.G.A.; op de Weegh, S.; Gieskes, J.F.B.; Schuring, R.W.

    2004-01-01

    Continuous Improvement is a well-known and consolidated concept in literature and practice and is considered vital in today¿s business environment. In 2003 a survey, as part of the international CINet survey, has been performed in the Netherlands in order to gain insight into the current practices

  2. Computer simulation of lower-hybrid heating and current drive in tokamaks. Progress report, September 1, 1982-July 1, 1983

    International Nuclear Information System (INIS)

    Ogden, J.M.

    1983-07-01

    A lower hybrid ray tracing package has been adapted for use in the PPPL 1-D tokamak transport code TRANSP. The code LHRAY has been written in OLYMPUS format and is suitable for use as a separate simulation program or in conjunction with TRANSP. The generality of the OLYMPUS conventions was chosen in order to make LHRAY easily transferable to other OLYMPUS style transport codes such as BALDUR. The details of LHRAY are described in this report. The physical model documented in our first progress report has been used with one major modification. Instead of solving the 1-D Fokker-Planck equations numerically to give the electron distribution function F/sub e/ in the presence of a background electric field, we have approximated F/sub e/ analytically using the theory of Liu et al for runaway electron distributions. The organization of LHRAY is given and the naming conventions are noted. Finally, preliminary results are presented. Program documentation and a listing of the code are included as appendices

  3. 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.)

  4. Recent QUEST experiments on non-inductive current drive and plasma-wall interaction towards steady state operation of spherical tokamak

    International Nuclear Information System (INIS)

    Hanada, K.; Zushi, H.; Idei, H.; Nakamura, K.; Nagashima, Y.; Hasegawa, M.; Fujisawa, A.; Higashijima, A.; Kawasaki, S.; Nakashima, H.; Ishiguro, M.; Tashima, S.; Kalinnikova, E.I.; Mitarai, O.; Maekawa, T.; Fukuyama, A.; Takase, Y.; Gao, X.; Liu, H.; Qian, J.; Ono, M.; Raman, R.; Peng, M.

    2015-01-01

    Full text of publication follows. Steady state operation (SSO) of magnetic fusion devices is one of the goals for fusion research. Development of non-inductive current drive and investigation of plasma-wall interaction (PWI) are issues to be resolved for SSO. Because of the very limited central solenoid (CS) flux in a spherical tokamak (ST), methods for non-inductive plasma current start-up and sustainment are necessary. Fully non-inductive plasma up to approximately 5 min was successfully demonstrated on the spherical tokamak QUEST. Furthermore, recharging of the center solenoid coil was also achieved in OH+RF plasmas with plasma current feedback using the CS. During the plasma start-up phase, precession motion of trapped electrons can drive some current, which plays an essential role in forming a closed flux surface. On QUEST, the main parts of the plasma facing components (PFCs) are covered by tungsten plates (W) or coated by W plasma spray and are actively cooled by water circulation. The increase in water temperature quantitatively provides the deposited power to each PFC. The power balance during long duration discharges has been studied for various types of magnetic configurations such as limiter, upper and lower single-null divertor discharges. As, the temperature of any PFCs reaches a steady-state condition during long pulse, the power balance can be obtained. It is found that the discharge duration of QUEST is significantly limited by particle imbalance shown by gradual increment of plasma and neutral density. The additional influx of neutrals was provided by recycling of hydrogen, which is still uncontrollable. A point model of particle balance was applied to a long-duration divertor discharge, and it was found that a small increment of particle-influx occurred around the end of the long duration discharge. A post-mortem analysis of surface-attaching specimen during an experimental campaign indicates that the increased amount of neutral influx could be

  5. Structure and relative importance of ponderomotive forces and current drive generated by converted fast waves in pre-heated low aspect ratio tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-05-12

    The generation in low aspect ratio tokamaks (LARTs) of ponderomotive forces and non-inductive current drive by the resonant fast wave-plasma interaction with mode conversion to kinetic Alfven waves (KAWs) and subsequent deposition, mainly by resonant electron Landau damping, is considered. The calculations follow the rigorous solution of the full wave equations upon using a dielectric tensor operator consisting of (i) a parallel conductivity including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped electrons and passing electrons+ions and (ii) perpendicular components provided by the resistive two-fluid model equations. The fast waves are launched by an antenna located on the low field side and extending {+-}45 deg. about the equatorial plane. A parametric investigation of the structure and importance of the various components of the ponderomotive forces and current drive generated in START-like plasmas is carried out and their suitability for supplementing the required non-rf toroidal equilibrium current is demonstrated.

  6. Study of the fast electron distribution function in lower hybrid and electron cyclotron current driven plasmas in the WT-3 tokamak

    International Nuclear Information System (INIS)

    Ogura, K.; Tanaka, H.; Ide, S.

    1991-01-01

    The distribution function f(p-vector) of fast electrons produced by lower hybrid current drive (LHCD) is investigated in the WT-3 tokamak, using a combination of measurements of the hard X-ray (HXR) angular distribution with respect to the toroidal magnetic field and observations of the HXR radial profile. The data obtained indicate the formation of a plateau-like region in f(p-vector) which corresponds to a region of resonant interaction between the lower hybrid (LH) wave and the electrons. The energy of the fast electrons in the peripheral plasma region is observed to be higher than that in the central plasma region under operational conditions with a high plasma current (I p ≥ 80 kA). At low current (I p < or approx. 50 kA), however, the energy of fast electrons is constant along the plasma radius. In the current ramp-up phase, fast electrons are generated in the directions normal to and opposite to the LH wave propagation. The latter case is ascribed to a negatively biased toroidal electric field induced by the current ramp-up. To study the characteristic change of f(p-vector) for various current drive mechanisms, HXR measurements are performed in electron cyclotron current driven (ECCD) plasma and in Ohmic heating (OH) plasma. In ECCD plasma, the perpendicular energy of fast electrons increases, which indicates that fast electrons are accelerated perpendicularly by electron cyclotron heating. In both LHCD and ECCD plasmas, fast electrons flow in the direction opposite to the wave propagation, while no such fast electrons are formed in OH plasma. (author). 33 refs, 16 figs, 1 tab

  7. The steady-state tokamak program

    International Nuclear Information System (INIS)

    Politzer, D.A.; Nevins, W.M.

    1992-01-01

    This paper reports on a steady-state tokamak experiment (STE) needed to develop the technology and physics data base required for construction of a steady-state fusion power demonstration reactor in the early 21st century. The STE will provide an integrated facility for the development and demonstration of steady-state and particle handling, low-activation high-heat-flux components and materials, efficient current drive, and continuous plasma performance in steady-state, with reactor-like plasma conditions under severe conditions of heat and particle bombardment of the wall. The STE facility will also be used to develop operation and control scenarios for ITER

  8. Inter-machine comparison of the termination phase and energy conversion in tokamak disruptions with runaway current plateau formation and implications for ITER

    International Nuclear Information System (INIS)

    Martín-Solís, J.R.; Loarte, A.; Hollmann, E.M.; Esposito, B.; Riccardo, V.

    2014-01-01

    The termination of the current and the loss of runaway electrons following runaway current plateau formation during disruptions have been investigated in the JET, DIII-D and FTU tokamaks. Substantial conversion of magnetic energy into runaway kinetic energy, up to ∼10 times the initial plateau runaway kinetic energy, has been inferred for the slowest current terminations. Both modelling and experiment suggest that, in present devices, the efficiency of conversion into runaway kinetic energy is determined to a great extent by the characteristic runaway loss time, τ diff , and the resistive time of the residual ohmic plasma after the disruption, τ res , increasing with the ratio τ diff /τ res . It is predicted that, in large future devices such as ITER, the generation of runaways by the avalanche mechanism will play an important role, particularly for slow runaway discharge terminations, increasing substantially the amount of energy deposited by the runaways onto the plasma-facing components by the conversion of magnetic energy of the runaway plasma into runaway kinetic energy. Estimates of the power fluxes on the beryllium plasma-facing components during runaway termination in ITER indicate that for runaway currents of up to 2 MA no melting of the components is expected. For larger runaway currents, minimization of the effects of runaway impact on the first wall requires a reduction in the kinetic energy of the runaway beam before termination and, in addition, high plasma density n e and low ohmic plasma resistance (long τ res ) to prevent large conversion of magnetic into runaway kinetic energy during slow current terminations. (paper)

  9. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  10. Observation of inward and outward particle convection in the core of electron cyclotron heated and current driven plasmas in the Tokamak a Configuration Variable

    International Nuclear Information System (INIS)

    Furno, I.; Weisen, H.

    2003-01-01

    In the Tokamak a Configuration Variable [F. Hofmann, J.B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)], inward or outward convection in the core of electron cyclotron heated and current driven plasmas is observed, depending on discharge conditions. In sawtoothing discharges with central electron cyclotron heating, outward convection is observed when a quasicontinuous m=1 kink mode is present, resulting in inverted sawteeth on the central electron density, while in the absence thereof, inward convection between successive sawtooth crashes leads to 'normal' sawteeth. The occurrence of a kink mode depends sensitively on plasma triangularity. When sawteeth are stabilized with central co- or counterelectron cyclotron current drive, stationary hollow electron density profiles are observed in the presence of m=1 modes, while peaked or flat profiles are observed in magnetohydrodynamic quiescent discharges. The observation of peaked density profiles in fully electron cyclotron driven plasmas demonstrates that pinch processes other than the Ware pinch must be responsible for these phenomena

  11. Study of the dynamics of the lower hybrid wave during current drive in tokamaks and of the Weyl-Wigner in quantum mechanics

    International Nuclear Information System (INIS)

    Bizarro, J.P.

    1993-10-01

    A comprehensive and detailed investigation is presented on the dynamics of the lower hybrid wave during current drive in tokamaks in situations where toroidally induced ray stochasticity is important and on the Weyl-Wigner formalism for rotation angle and angular momentum variables in quantum mechanics. It is shown that ray-tracing and Fokker-Planck codes are reliable tools for modelling the physics of lower-hybrid current drive provided a large number of rays is used when stochastic effects are important, and, in particular, that such codes are capable of reproducing the experimentally observed features of the hard X-ray emission. The balance between the wave damping and the stochastic divergence of nearby ray trajectories appears to be of great importance in governing the dynamics of the launched power spectrum and in establishing the characteristics of the deposition patterns. The implications of rotational periodicity and of angular momentum quantization for the Weyl-Wigner formalism are analyzed. Particular attention is paid to discreteness and its consequences: importance of evenness and oddness, use of two difference operators instead of one differential operator. 24 refs

  12. Conversion of continuous-direct-current TIG welder to pulse-arc operation

    Science.gov (United States)

    Lien, D. R.

    1969-01-01

    Electronics package converts a continuous-dc tungsten-inert gas welder for pulse-arc operation. Package allows presetting of the pulse rate, duty cycle, and current value, and enables welding of various alloys and thicknesses of materials.

  13. 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.

  14. Simulation of a major tokamak disruption

    International Nuclear Information System (INIS)

    White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

    1977-08-01

    It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

  15. MAST: a Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

    1995-01-01

    The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

  16. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  17. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  18. Effect of Continuous Current during Pauses between Successive Strokes on the Decay of the Lightning Channel

    International Nuclear Information System (INIS)

    Aleksandrov, N.L.; Bazelyan, E.M.; Shneider, M. N.

    2000-01-01

    A one-dimensional model is used to study the dynamics of the hydrodynamic parameters of the lightning channel in the return stroke and after the pulse current is damped. The effect of the continuous residual electric current during pauses between the successive strokes on the plasma cooling in the channel is analyzed. It is shown that a continuous electric current, which is several orders of magnitude lower than the peak current in the return stroke, is capable of maintaining the channel conductivity. This effect cannot be explained merely by Joule heating but is largely governed by the fact that the turbulent heat transport is substantially suppressed. In this case, even a continuous current as low as 50-100 A is capable of maintaining the conductivity of the lightning channel at a level at which only M-components can develop in the channel rather than the dart leader of the subsequent stroke

  19. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

    2000-12-01

    Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

  20. Energy losses on tokamak startup

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells

  1. Hamiltonian study of the response of a tokamak plasma to the ion cyclotron heating wave: minor heating and current generation by the fast wave

    International Nuclear Information System (INIS)

    Becoulet, A.

    1990-06-01

    The role of additional Heatings, such as the Ion Cyclotron Heating, is to raise magnetic fusion plasmas to higher temperatures, to satisfy the ignition condition. The understanding of the wave absorption mechanisms by the plasma first requires a precise description of the particle individual trajectories. The Hamiltonian mechanics, through action-angle variables, allows this description, and makes the computation of the wave-particle interaction easier. We then derive a quantitative evaluation of the intrinsic stochasticity for ionic trajectories perturbated by the fast wave. This stochasticity, combinated to the collisional effects, gives the validity domain for a quasilinear approximation of the evolution equation. This equation is then written under a variational formulation, and solved semi-analytically. Results conclude to the importance of the Hamiltonian chaos in the formation of the deeply anisotropic distribution tails, encountered in minority heating scenarios. Direct interaction of the electrons and the fast wave is similarly analysed. The influence of the various parameters (wave spectrum, magnetic configuration, frequency,...) is then examined in order to optimize this scenario of fast wave current drive in tokamaks [fr

  2. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  3. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  4. 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)

  5. Overview of the Tokamak de Varennes program

    International Nuclear Information System (INIS)

    Pacher, H.D.

    1986-01-01

    The Tokamak de Varennes will be the major Canadian experiment in the magnetic fusion domain. It has a toroidal field of 1.5 tesla, major radius of 0.85 m, a minor radius of 0.25 m, and will study long pulses, up to 30 seconds duration. Initially, a series of successive plasma pulses, each of the order of seconds, will yield a duty factor of over 50 percent. During this phase, the major emphasis will be on the study of impurity generation, transport, and control, plasma-wall interactions and material properties. The program will include studies of fast current rampdown and the resultant current profile modifications. The development of advanced diagnostics will also be undertaken. To attain a higher duty factor with continuous plasma operation, noninductive current drive by radio=frequency will be added as an early upgrade. This will introduce current drive investigations such as transformer recharge and profile relaxation, and enhance the wall and materials study program. In this context, the Tokamak de Varennes will concentrate on the study of impurity exhaust and retention as well as net erosion of the limiter and neutralization plate materials

  6. Ion energy spectrum just after the application of current pulse for turbulent heating in the TRIAM-1 tokamak

    International Nuclear Information System (INIS)

    Nakamura, Kazuo; Nakamura, Yukio; Hiraki, Naoji; Itoh, Satoshi

    1981-01-01

    Temporal evolution and spatial profile of ion energy spectrum just after the application of current pulse for turbulent heating are investigated experimentally in TRIAM-1 and numerically with a Fokker-Planck equation. Two-component ion energy spectrum formed by turbulent heating relaxes to single one within tau sub(i) (ion collision time). (author)

  7. Direct observation of current in type-I edge-localized-mode filaments on the ASDEX upgrade tokamak

    DEFF Research Database (Denmark)

    Vianello, N.; Zuin, M.; Cavazzana, R.

    2011-01-01

    Magnetically confined plasmas in the high confinement regime are regularly subjected to relaxation oscillations, termed edge localized modes (ELMs), leading to large transport events. Present ELM theories rely on a combined effect of edge current and the edge pressure gradients which result...

  8. Ion energy spectrum just after the application of current pulse for turbulent heating in the TRIAM-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-07-01

    Temporal evolution and spatial profile of ion energy spectrum just after the application of current pulse for turbulent heating are investigated experimentally in TRIAM-1 and numerically with a Fokker-Planck equation. Two-component ion energy spectrum formed by turbulent heating relaxes to single one within tau sub(i) (ion collision time).

  9. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

  10. Current distributions in superconducting wires subject to a random orientation magnetic field, and corresponding to the Tokamak usual conditions

    International Nuclear Information System (INIS)

    Artaud, J.F.

    1994-01-01

    The main themes of this thesis are: review of superconductivity principles; critical current in a random orientation magnetic field; the MHD model applied to superconductors (with comprehensive calculation of the field in a plate type conductor); the magnetization created by a variation of a random orientation magnetic field; the electric field in a superconductor in steady or quasi-steady state (MHD displacement, pinning and thermal effects). 145 figs., 166 refs

  11. Transcranial direct current stimulation in refractory continuous spikes and waves during slow sleep: a controlled study

    DEFF Research Database (Denmark)

    Varga, Edina T; Terney, Daniella; Atkins, Mary D

    2011-01-01

    Cathodal transcranial direct current stimulation (tDCS) decreases cortical excitability. The purpose of the study was to investigate whether cathodal tDCS could interrupt the continuous epileptiform activity. Five patients with focal, refractory continuous spikes and waves during slow sleep were...... recruited. Cathodal tDCS and sham stimulation were applied to the epileptic focus, before sleep (1 mA; 20 min). Cathodal tDCS did not reduce the spike-index in any of the patients....

  12. EDITORIAL: Special section on recent progress on radio frequency heating and current drive studies in the JET tokamak Special section on recent progress on radio frequency heating and current drive studies in the JET tokamak

    Science.gov (United States)

    Ongena, Jef; Mailloux, Joelle; Mayoral, Marie-Line

    2009-04-01

    This special cluster of papers summarizes the work accomplished during the last three years in the framework of the Task Force Heating at JET, whose mission it is to study the optimisation of heating systems for plasma heating and current drive, launching and deposition questions and the physics of plasma rotation. Good progress and new physics insights have been obtained with the three heating systems available at JET: lower hybrid (LH), ion cyclotron resonance heating (ICRH) and neutral beam injection (NBI). Topics covered in the present issue are the use of edge gas puffing to improve the coupling of LH waves at large distances between the plasma separatrix and the LH launcher. Closely linked with this topic are detailed studies of the changes in LH coupling due to modifications in the scrape-off layer during gas puffing and simultaneous application of ICRH. We revisit the fundamental ICRH heating of D plasmas, include new physics results made possible by recently installed new diagnostic capabilities on JET and point out caveats for ITER when NBI is simultaneously applied. Other topics are the study of the anomalous behaviour of fast ions from NBI, and a study of toroidal rotation induced by ICRH, both again with possible implications for ITER. In finalizing this cluster of articles, thanks are due to all colleagues involved in preparing and executing the JET programme under EFDA in recent years. We want to thank the EFDA leadership for the special privilege of appointing us as Leaders or Deputies of Task Force Heating, a wonderful and hardworking group of colleagues. Thanks also to all other European and non-European scientists who contributed to the JET scientific programme, the Operations team of JET and the colleagues of the Close Support Unit (CSU). Thanks are also due to the Editors, Editorial Board and referees of Plasma Physics and Controlled Fusion together with the publishing staff of IOP Publishing who have supported and contributed substantially to

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

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  14. Non-linear Simulations of MHD Instabilities in Tokamaks Including Eddy Current Effects and Perspectives for the Extension to Halo Currents

    International Nuclear Information System (INIS)

    Hoelzl, M; Merkel, P; Lackner, K; Strumberger, E; Huijsmans, G T A; Aleynikova, K; Liu, F; Atanasiu, C; Nardon, E; Fil, A; McAdams, R; Chapman, I

    2014-01-01

    The dynamics of large scale plasma instabilities can be strongly influenced by the mutual interaction with currents flowing in conducting vessel structures. Especially eddy currents caused by time-varying magnetic perturbations and halo currents flowing directly from the plasma into the walls are important. The relevance of a resistive wall model is directly evident for Resistive Wall Modes (RWMs) or Vertical Displacement Events (VDEs). However, also the linear and non-linear properties of most other large-scale instabilities may be influenced significantly by the interaction with currents in conducting structures near the plasma. The understanding of halo currents arising during disruptions and VDEs, which are a serious concern for ITER as they may lead to strong asymmetric forces on vessel structures, could also benefit strongly from these non-linear modeling capabilities. Modeling the plasma dynamics and its interaction with wall currents requires solving the magneto-hydrodynamic (MHD) equations in realistic toroidal X-point geometry consistently coupled with a model for the vacuum region and the resistive conducting structures. With this in mind, the non-linear finite element MHD code JOREK [1, 2] has been coupled [3] with the resistive wall code STARWALL [4], which allows us to include the effects of eddy currents in 3D conducting structures in non-linear MHD simulations. This article summarizes the capabilities of the coupled JOREK-STARWALL system and presents benchmark results as well as first applications to non-linear simulations of RWMs, VDEs, disruptions triggered by massive gas injection, and Quiescent H-Mode. As an outlook, the perspectives for extending the model to halo currents are described

  15. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

    1990-10-01

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  16. The effects of electron cyclotron heating and current drive on toroidal Alfvén eigenmodes in tokamak plasmas

    Science.gov (United States)

    Sharapov, S. E.; Garcia-Munoz, M.; Van Zeeland, M. A.; Bobkov, B.; Classen, I. G. J.; Ferreira, J.; Figueiredo, A.; Fitzgerald, M.; Galdon-Quiroga, J.; Gallart, D.; Geiger, B.; Gonzalez-Martin, J.; Johnson, T.; Lauber, P.; Mantsinen, M.; Nabais, F.; Nikolaeva, V.; Rodriguez-Ramos, M.; Sanchis-Sanchez, L.; Schneider, P. A.; Snicker, A.; Vallejos, P.; the AUG Team; the EUROfusion MST1 Team

    2018-01-01

    Dedicated studies performed for toroidal Alfvén eigenmodes (TAEs) in ASDEX-Upgrade (AUG) discharges with monotonic q-profiles have shown that electron cyclotron resonance heating (ECRH) can make TAEs more unstable. In these AUG discharges, energetic ions driving TAEs were obtained by ion cyclotron resonance heating (ICRH). It was found that off-axis ECRH facilitated TAE instability, with TAEs appearing and disappearing on timescales of a few milliseconds when the ECRH power was switched on and off. On-axis ECRH had a much weaker effect on TAEs, and in AUG discharges performed with co- and counter-current electron cyclotron current drive (ECCD), the effects of ECCD were found to be similar to those of ECRH. Fast ion distributions produced by ICRH were computed with the PION and SELFO codes. A significant increase in T e caused by ECRH applied off-axis is found to increase the fast ion slowing-down time and fast ion pressure causing a significant increase in the TAE drive by ICRH-accelerated ions. TAE stability calculations show that the rise in T e causes also an increase in TAE radiative damping and thermal ion Landau damping, but to a lesser extent than the fast ion drive. As a result of the competition between larger drive and damping effects caused by ECRH, TAEs become more unstable. It is concluded, that although ECRH effects on AE stability in present-day experiments may be quite significant, they are determined by the changes in the plasma profiles and are not particularly ECRH specific.

  17. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  18. Correlation analysis of motor current and chatter vibration in grinding using complex continuous wavelet coherence

    International Nuclear Information System (INIS)

    Liu, Yao; Wang, Xiufeng; Lin, Jing; Zhao, Wei

    2016-01-01

    Motor current is an emerging and popular signal which can be used to detect machining chatter with its multiple advantages. To achieve accurate and reliable chatter detection using motor current, it is important to make clear the quantitative relationship between motor current and chatter vibration, which has not yet been studied clearly. In this study, complex continuous wavelet coherence, including cross wavelet transform and wavelet coherence, is applied to the correlation analysis of motor current and chatter vibration in grinding. Experimental results show that complex continuous wavelet coherence performs very well in demonstrating and quantifying the intense correlation between these two signals in frequency, amplitude and phase. When chatter occurs, clear correlations in frequency and amplitude in the chatter frequency band appear and the phase difference of current signal to vibration signal turns from random to stable. The phase lead of the most correlated chatter frequency is the largest. With the further development of chatter, the correlation grows up in intensity and expands to higher order chatter frequency band. The analyzing results confirm that there is a consistent correlation between motor current and vibration signals in the grinding chatter process. However, to achieve accurate and reliable chatter detection using motor current, the frequency response bandwidth of current loop of the feed drive system must be wide enough to response chatter effectively. (paper)

  19. Continuous counter-current chromatography for capture and polishing steps in biopharmaceutical production.

    Science.gov (United States)

    Steinebach, Fabian; Müller-Späth, Thomas; Morbidelli, Massimo

    2016-09-01

    The economic advantages of continuous processing of biopharmaceuticals, which include smaller equipment and faster, efficient processes, have increased interest in this technology over the past decade. Continuous processes can also improve quality assurance and enable greater controllability, consistent with the quality initiatives of the FDA. Here, we discuss different continuous multi-column chromatography processes. Differences in the capture and polishing steps result in two different types of continuous processes that employ counter-current column movement. Continuous-capture processes are associated with increased productivity per cycle and decreased buffer consumption, whereas the typical purity-yield trade-off of classical batch chromatography can be surmounted by continuous processes for polishing applications. In the context of continuous manufacturing, different but complementary chromatographic columns or devices are typically combined to improve overall process performance and avoid unnecessary product storage. In the following, these various processes, their performances compared with batch processing and resulting product quality are discussed based on a review of the literature. Based on various examples of applications, primarily monoclonal antibody production processes, conclusions are drawn about the future of these continuous-manufacturing technologies. Copyright © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. Development of Operation Scenario for Spherical Tokamak at SNU

    International Nuclear Information System (INIS)

    Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

    2009-01-01

    Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

  1. Stratospheric Joule heating by lightning continuing current inferred from radio remote sensing

    DEFF Research Database (Denmark)

    Fullekrug, M.; Ignaccolo, M.; Kuvshinov, A.

    2006-01-01

    to ground deposits electrical energy into the stratosphere resulting from quasi-static ( Joule) heating. The energy deposition is dominated by the lightning continuing current, and it is similar to 10(-5) J/m(3) at 30 km height. It is speculated that the initiation of blue jets and gigantic jets...

  2. Continuous measurements of discharge from a horizontal acoustic Doppler current profiler in a tidal river

    NARCIS (Netherlands)

    Hoitink, A.J.F.; Buschman, F.A.; Vermeulen, B.

    2009-01-01

    Acoustic Doppler current profilers (ADCPs) can be mounted horizontally at a river bank, yielding single-depth horizontal array observations of velocity across the river. This paper presents a semideterministic, semistochastic method to obtain continuous measurements of discharge from horizontal ADCP

  3. Current reversal in a continuously periodic system driven by an additive noise and a multiplicative noise

    International Nuclear Information System (INIS)

    Wang Canjun; Chen Shibo; Mei Dongcheng

    2006-01-01

    We study the noise-induce transport and current reversal of Brownian particles in a continuously periodic potential driven by cross correlation between a multiplicative white noise and an additive white noise. We find that directed motion of the Brownian particles can be induced by the correlation between the additive noise and the multiplicative noise. The current reversal and the direction of the current is controlled by the values of the intensity (λ) of the correlated noises and a dimensionless parameter R (R=α/D, D is the intensity of multiplicative noise and α is the intensity of additive noise)

  4. Magnetic island formation in tokamaks

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs

  5. 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

  6. Study on H-mode access at low density with lower hybrid current drive and lithium-wall coatings on the EAST superconducting tokamak

    DEFF Research Database (Denmark)

    Xu, G.S.; Wan, B.N.; Li, J.G.

    2011-01-01

    The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of HIPB98(y,2) ~ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before ...

  7. 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

  8. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  9. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  10. Education technology with continuous real time monitoring of the current functional and emotional students' states

    Science.gov (United States)

    Alyushin, M. V.; Kolobashkina, L. V.

    2017-01-01

    The education technology with continuous monitoring of the current functional and emotional students' states is suggested. The application of this technology allows one to increase the effectiveness of practice through informed planning of the training load. For monitoring the current functional and emotional students' states non-contact remote technologies of person bioparameters registration are encouraged to use. These technologies are based on recording and processing in real time the main person bioparameters in a purely passive mode. Experimental testing of this technology has confirmed its effectiveness.

  11. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

  12. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

    Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

    1996-01-01

    We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

  13. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  14. Continuous Glucose Monitoring in the Cardiac ICU: Current Use and Future Directions.

    Science.gov (United States)

    Scrimgeour, Laura A; Potz, Brittany A; Sellke, Frank W; Abid, M Ruhul

    2017-11-01

    Perioperative glucose control is highly important, particularly for patients undergoing cardiac surgery. Variable glucose levels before, during and after cardiac surgery lead to increased post-operative complications and patient mortality. [1] Current methods for intensive monitoring and treating hyperglycemia in the Intensive Care Unit (ICU) usually involve hourly glucose monitoring and continuous intravenous insulin infusions. With the advent of more accurate subcutaneous glucose monitoring systems, the role of improved glucose control with newer systems deserves consideration for widespread adoption.

  15. 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

  16. Low Power Continuous-Time Delta-Sigma ADC with Current Output DAC

    DEFF Research Database (Denmark)

    Marker-Villumsen, Niels; Jørgensen, Ivan Harald Holger; Bruun, Erik

    2015-01-01

    The paper presents a continuous-time (CT) DeltaSigma (∆Σ) analog-to-digital converter (ADC) using a current output digital-to-analog converter (DAC) for the feedback. From circuit analysis it is shown that using a current output DAC makes it possible to relax the noise requirements of the 1st...... integrator of the loopfilter, and thereby reduce the current consumption. Furthermore, the noise of the current output DAC being dependent on the ADC input signal level, enabling a dynamic range that is larger than the peak signal-to-noise ratio (SNR). The current output DAC is used in a 3rd order multibit...... CT ∆Σ ADC for audio applications, designed in a 0.18 µm CMOS process, with active-RC integrators, a 7-level Flash ADC quantizer and current output DAC for the feedback. From simulations the ADC achieves a dynamic range of 95.0 dB in the audio band, with a current consumption of 284 µA for a 1.7 V...

  17. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  18. Runaway electrons during tokamak startup

    International Nuclear Information System (INIS)

    Sharma, A.S.; Jayakumar, R.

    1988-01-01

    Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs

  19. 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

  20. 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

  1. Empirical scaling for present Ohmically heated tokamaks

    International Nuclear Information System (INIS)

    Daughney, C.

    1975-01-01

    Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)

  2. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  3. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  4. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-01-01

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics

  5. Disruptions in Tokamaks

    International Nuclear Information System (INIS)

    Bondeson, A.

    1987-01-01

    This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

  6. Spherical tokamak without external toroidal fields

    International Nuclear Information System (INIS)

    Kaw, P.K.; Avinash, K.; Srinivasan, R.

    2001-01-01

    A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

  7. Preliminary results of the TBR small tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.

    1982-01-01

    The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)

  8. Full power in the European tokamak

    International Nuclear Information System (INIS)

    Lallia, P.P.; Hugon, M.

    1987-01-01

    A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

  9. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  10. 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.

  11. Economic considerations of commercial tokamak options

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-05-01

    Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)

  12. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  13. Continuing-education needs of the currently employed public health education workforce.

    Science.gov (United States)

    Allegrante, J P; Moon, R W; Auld, M E; Gebbie, K M

    2001-08-01

    This study examined the continuing-education needs of the currently employed public health education workforce. A national consensus panel of leading health educators from public health agencies, academic institutions, and professional organizations was convened to examine the forces creating the context for the work of public health educators and the competencies they need to practice effectively. Advocacy; business management and finance; communication; community health planning and development, coalition building, and leadership; computing and technology; cultural competency; evaluation; and strategic planning were identified as areas of critical competence. Continuing education must strengthen a broad range of critical competencies and skills if we are to ensure the further development and effectiveness of the public health education workforce.

  14. Disruption generated secondary runaway electrons in present day tokamaks

    International Nuclear Information System (INIS)

    Pankratov, I.M.; Jaspers, R.

    2000-01-01

    An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

  15. 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

  16. 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.

  17. 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.

  18. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  19. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  20. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  1. High Beta Tokamak research

    International Nuclear Information System (INIS)

    Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.

    1993-01-01

    During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year

  2. 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)

  3. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  4. 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.)

  5. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

    2004-01-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

  6. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2004-09-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

  7. Smoke and mirrors: the perceived benefits of continued tobacco use among current smokers

    Directory of Open Access Journals (Sweden)

    Hugh Klein

    2014-09-01

    Full Text Available Despite 50+ years of public health efforts to reduce smoking rates in the United States, approximately one-fifth of the adults living in this country continue to smoke cigarettes. Previous studies have examined smokers’ risk perceptions of cigarette smoking, as well as the perceived benefits of quitting smoking. Less research has focused on the perceived benefits of smoking among current cigarette smokers. The latter is the main focus of the present paper. Questionnaire-based interviews were conducted with a community-based sample of 485 adult current cigarette smokers recruited from the Atlanta, Georgia, metropolitan area between 2004 and 2007. Active and passive recruiting approaches were used, along with a targeted sampling strategy. Results revealed that most current cigarette smokers perceive themselves to experience benefits as a result of their cigarette use, including (among others increased relaxation, diminished nervousness in social situations, enjoyment of the taste of cigarettes when smoking, and greater enjoyment of parties when smoking. Perceiving benefits from cigarette smoking was associated with a variety of tobacco use measures, such as smoking more cigarettes, an increased likelihood of chain smoking, and overall negative attitude toward quitting smoking, among others. Several factors were associated with the extent to which smokers perceived themselves to benefit from their tobacco use, including education attainment, the age of first purchasing cigarettes, the proportion of friends who smoked, hiding smoking from others, being internally-oriented regarding locus of control, and self-esteem.

  8. Self-consistent kinetic simulations of lower hybrid drift instability resulting in electron current driven by fusion products in tokamak plasmas

    International Nuclear Information System (INIS)

    Cook, J W S; Chapman, S C; Dendy, R O; Brady, C S

    2011-01-01

    We present particle-in-cell (PIC) simulations of minority energetic protons in deuterium plasmas, which demonstrate a collective instability responsible for emission near the lower hybrid frequency and its harmonics. The simulations capture the lower hybrid drift instability in a parameter regime motivated by tokamak fusion plasma conditions, and show further that the excited electromagnetic fields collectively and collisionlessly couple free energy from the protons to directed electron motion. This results in an asymmetric tail antiparallel to the magnetic field. We focus on obliquely propagating modes excited by energetic ions, whose ring-beam distribution is motivated by population inversions related to ion cyclotron emission, in a background plasma with a temperature similar to that of the core of a large tokamak plasma. A fully self-consistent electromagnetic relativistic PIC code representing all vector field quantities and particle velocities in three dimensions as functions of a single spatial dimension is used to model this situation, by evolving the initial antiparallel travelling ring-beam distribution of 3 MeV protons in a background 10 keV Maxwellian deuterium plasma with realistic ion-electron mass ratio. These simulations provide a proof-of-principle for a key plasma physics process that may be exploited in future alpha channelling scenarios for magnetically confined burning plasmas.

  9. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

    1992-03-01

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  10. Transport Barriers in Bootstrap Driven Tokamaks

    Science.gov (United States)

    Staebler, Gary

    2017-10-01

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

  11. Continuous Glucose Monitoring: Current Use in Diabetes Management and Possible Future Applications.

    Science.gov (United States)

    Vettoretti, Martina; Cappon, Giacomo; Acciaroli, Giada; Facchinetti, Andrea; Sparacino, Giovanni

    2018-05-01

    The recent announcement of the production of new low-cost continuous glucose monitoring (CGM) sensors, the approval of marketed CGM sensors for making treatment decisions, and new reimbursement criteria have the potential to revolutionize CGM use. After briefly summarizing current CGM applications, we discuss how, in our opinion, these changes are expected to extend CGM utilization beyond diabetes patients, for example, to subjects with prediabetes or even healthy individuals. We also elaborate on how the integration of CGM data with other relevant information, for example, health records and other medical device/wearable sensor data, will contribute to creating a digital data ecosystem that will improve our understanding of the etiology and complications of diabetes and will facilitate the development of data analytics for personalized diabetes management and prevention.

  12. Current Research on Containment Technologies for Verification Activities: Advanced Tools for Maintaining Continuity of Knowledge

    International Nuclear Information System (INIS)

    Smartt, H.; Kuhn, M.; Krementz, D.

    2015-01-01

    The U.S. National Nuclear Security Administration (NNSA) Office of Non-proliferation and Verification Research and Development currently funds research on advanced containment technologies to support Continuity of Knowledge (CoK) objectives for verification regimes. One effort in this area is the Advanced Tools for Maintaining Continuity of Knowledge (ATCK) project. Recognizing that CoK assurances must withstand potential threats from sophisticated adversaries, and that containment options must therefore keep pace with technology advances, the NNSA research and development on advanced containment tools is an important investment. The two ATCK efforts underway at present address the technical containment requirements for securing access points (loop seals) and protecting defined volumes. Multiple U.S. national laboratories are supporting this project: Sandia National Laboratories (SNL), Savannah River National Laboratory (SRNL), and Oak Ridge National Laboratory (ORNL). SNL and SRNL are developing the ''Ceramic Seal,'' an active loop seal that integrates multiple advanced security capabilities and improved efficiency housed within a small-volume ceramic body. The development includes an associated handheld reader and interface software. Currently at the prototype stage, the Ceramic Seal will undergo a series of tests to determine operational readiness. It will be field tested in a representative verification trial in 2016. ORNL is developing the Whole Volume Containment Seal (WCS), a flexible conductive fabric capable of enclosing various sizes and shapes of monitored items. The WCS includes a distributed impedance measurement system for imaging the fabric surface area and passive tamper-indicating features such as permanent-staining conductive ink. With the expected technology advances from the Ceramic Seal and WCS, the ATCK project takes significant steps in advancing containment technologies to help maintain CoK for various verification

  13. [Current status of continuous subcutaneous insulin infusion and continuous glucose monitoring systems in the Community of Madrid].

    Science.gov (United States)

    Arranz Martín, Alfonso; Calle Pascual, Alfonso; Del Cañizo Gómez, Francisco Javier; González Albarrán, Olga; Lisbona Gil, Arturo; Botella Serrano, Marta; Pallardo Sánchez, Luis Felipe

    2015-04-01

    To analyze the available information about continuous subcutaneous insulin infusion (CSII) and continuous glucose monitoring (CGM) systems in the public health care system of the Community of Madrid. A survey consisting of 31 items was sent to the 28 endocrinology department of the Madrid public hospitals. Items focused on CSII and CGM and included patients' registrations, as well as data regarding healthcare, administrative, and logistic aspects. Responses from a total of 20 hospitals where these procedures are used were received from March 2013 to May 2014. Data about pediatric patients were obtained from adult endocrinology departments, except for two hospitals which directly reported the information. A total of 1256 CSII pumps were recorded in the Madrid region, of which 1089 were used by adults, and the remaining 167 by pediatric patients. During 2013, 151 new CSII systems were implanted (12% of the total), while 14 pumps were withdrawn. Availability of human resources (medical assistance) and the number of staff practitioners experienced in management of these systems widely varied between hospitals. Eighty-five percent of hospitals used retrospective CGM systems, and 40% routinely placed them before starting an insulin pump. Thirteen hospitals (65%) used long-term, real-time CGM systems in selected cases (a total of 67 patients). Use of these technologies in diabetes is unequal between public health care hospitals in Madrid, and is still significantly lower as compared to other countries with similar incomes. However, there appears to be a trend to an increase in their use. Copyright © 2014 SEEN. Published by Elsevier España, S.L.U. All rights reserved.

  14. Tokamak control simulator

    International Nuclear Information System (INIS)

    Edelbaum, T.N.; Serben, S.; Var, R.E.

    1976-01-01

    A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

  15. Kirchhoff and Ohm in action: solving electric currents in continuous extended media

    Science.gov (United States)

    Dolinko, A. E.

    2018-03-01

    In this paper we show a simple and versatile computational simulation method for determining electric currents and electric potential in 2D and 3D media with arbitrary distribution of resistivity. One of the highlights of the proposed method is that the simulation space containing the distribution of resistivity and the points of external applied voltage are introduced by means of digital images or bitmaps, which easily allows simulating any phenomena involving distributions of resistivity. The simulation is based on the Kirchhoff’s laws of electric currents and it is solved by means of an iterative procedure. The method is also generalised to account for media with distributions of reactive impedance. At the end of this work, we show an example of application of the simulation, consisting in reproducing the response obtained with the geophysical method of electric resistivity tomography in presence of soil cracks. This paper is aimed at undergraduate or graduated students interested in computational physics and electricity and also researchers involved in the area of continuous electric media, which could find a simple and powerful tool for investigation.

  16. Design and construction of electronic components for a ''Novillo'' Tokamak

    International Nuclear Information System (INIS)

    Lopez C, R.

    1986-07-01

    The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

  17. Physics design requirements for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.

    1993-01-01

    The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust

  18. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

  19. Predicting core losses and efficiency of SRM in continuous current mode of operation using improved analytical technique

    International Nuclear Information System (INIS)

    Parsapour, Amir; Dehkordi, Behzad Mirzaeian; Moallem, Mehdi

    2015-01-01

    In applications in which the high torque per ampere at low speed and rated power at high speed are required, the continuous current method is the best solution. However, there is no report on calculating the core loss of SRM in continuous current mode of operation. Efficiency and iron loss calculation which are complex tasks in case of conventional mode of operation is even more involved in continuous current mode of operation. In this paper, the Switched Reluctance Motor (SRM) is modeled using finite element method and core loss and copper loss of SRM in discontinuous and continuous current modes of operation are calculated using improved analytical techniques to include the minor loop losses in continuous current mode of operation. Motor efficiency versus speed in both operation modes is obtained and compared. - Highlights: • Continuous current method for Switched Reluctance Motor (SRM) is explained. • An improved analytical technique is presented for SRM core loss calculation. • SRM losses in discontinuous and continuous current operation modes are presented. • Effect of mutual inductances on SRM performance is investigated

  20. Predicting core losses and efficiency of SRM in continuous current mode of operation using improved analytical technique

    Energy Technology Data Exchange (ETDEWEB)

    Parsapour, Amir, E-mail: amirparsapour@gmail.com [Department of Electrical Engineering, University of Isfahan, Isfahan (Iran, Islamic Republic of); Dehkordi, Behzad Mirzaeian, E-mail: mirzaeian@eng.ui.ac.ir [Department of Electrical Engineering, University of Isfahan, Isfahan (Iran, Islamic Republic of); Moallem, Mehdi, E-mail: moallem@cc.iut.ac.ir [Department of Electrical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of)

    2015-03-15

    In applications in which the high torque per ampere at low speed and rated power at high speed are required, the continuous current method is the best solution. However, there is no report on calculating the core loss of SRM in continuous current mode of operation. Efficiency and iron loss calculation which are complex tasks in case of conventional mode of operation is even more involved in continuous current mode of operation. In this paper, the Switched Reluctance Motor (SRM) is modeled using finite element method and core loss and copper loss of SRM in discontinuous and continuous current modes of operation are calculated using improved analytical techniques to include the minor loop losses in continuous current mode of operation. Motor efficiency versus speed in both operation modes is obtained and compared. - Highlights: • Continuous current method for Switched Reluctance Motor (SRM) is explained. • An improved analytical technique is presented for SRM core loss calculation. • SRM losses in discontinuous and continuous current operation modes are presented. • Effect of mutual inductances on SRM performance is investigated.

  1. Comparative studies of stellarator and tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  2. Resistive instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

  3. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

  4. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

  5. Simulation models: a current indispensable tool in studies of the continuous water-soil-plant - atmosphere

    International Nuclear Information System (INIS)

    Lopez Seijas, Teresa; Gonzalez, Felicita; Cid, G.; Osorio, Maria de los A.; Ruiz, Maria Elena

    2008-01-01

    Full text: This work assesses the current use of simulation models as a tool useful and indispensable for the advancement in the research and study of the processes related to the continuous water-soil - plant-atmosphere. In recent years they have reported in the literature many jobs where these modeling tools are used as a support to the decision-making process of companies or organizations in the agricultural sphere and in Special for the design of optimal management of irrigation and fertilization strategies of the crops. Summarizes some of the latest applications reported with respect to the use of water transfers and solutes, such simulation models mainly to nitrate leaching and groundwater contamination problems. On the other hand also summarizes important applications of simulation models of growth of cultivation for the prediction of effects on the performance of different conditions of water stress, and finally some other applications on the management of the different irrigation technologies as kingpins, superfiail irrigation and drip irrigation. Refer also the main work carried out in Cuba. (author)

  6. Design of a New Water Load for S-band 750 kW Continuous Wave High Power Klystron Used in EAST Tokamak

    Science.gov (United States)

    Liu, Liang; Liu, Fukun; Shan, Jiafang; Kuang, Guangli

    2007-04-01

    In order to test the klystrons operated at a frequency of 3.7 GHz in a continuous wave (CW) mode, a type of water load to absorb its power up to 750 kW is presented. The distilled water sealed with an RF ceramic window is used as the absorbent. At a frequency range of 70 MHz, the VSWR (Voltage Standing Wave Ratio) is below 1.2, and the rise in temperature of water is about 30 oC at the highest power level.

  7. Helicity content and tokamak applications of helicity

    International Nuclear Information System (INIS)

    Boozer, A.H.

    1986-05-01

    Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities

  8. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  9. 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)

  10. Energy storage for tokamak reactor cycles

    International Nuclear Information System (INIS)

    Buchanan, C.H.

    1979-01-01

    The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion

  11. Continuous improvement in the Netherlands: current practices and experiences in Dutch manufacturing industry (awarded with ANBAR Citation of excellence)

    NARCIS (Netherlands)

    Gieskes, J.F.B.; Baudet, F.C.M.; Baudet, Frank; Schuring, R.W.; Boer, Harm

    1997-01-01

    In order to get insight into the current continuous-improvement practices in European industry, EuroCINet carried out a survey in its member countries. In this article, continuous-improvement activities in a sample of 135 Dutch industrial companies are described. The results show that CI is a

  12. 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).

  13. 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

  14. 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

  15. Continuous development of current sheets near and away from magnetic nulls

    International Nuclear Information System (INIS)

    Kumar, Sanjay; Bhattacharyya, R.

    2016-01-01

    The presented computations compare the strength of current sheets which develop near and away from the magnetic nulls. To ensure the spontaneous generation of current sheets, the computations are performed congruently with Parker's magnetostatic theorem. The simulations evince current sheets near two dimensional and three dimensional magnetic nulls as well as away from them. An important finding of this work is in the demonstration of comparative scaling of peak current density with numerical resolution, for these different types of current sheets. The results document current sheets near two dimensional magnetic nulls to have larger strength while exhibiting a stronger scaling than the current sheets close to three dimensional magnetic nulls or away from any magnetic null. The comparative scaling points to a scenario where the magnetic topology near a developing current sheet is important for energetics of the subsequent reconnection.

  16. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

    Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

    1999-01-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

  17. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  18. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Zoletnik, S.

    1988-11-01

    Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

  19. Gas blanket fueling of a tokamak reactor

    International Nuclear Information System (INIS)

    Gralnick, S.L.

    1978-01-01

    The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs

  20. Alfven wave heating in a tokamak reactor

    International Nuclear Information System (INIS)

    Borg, G.G.; Appert, K.; Knight, A.J.; Lister, J.B.; Vaclavik, J.

    1990-01-01

    A number of features of Alfven wave heating make it potentially attractive for use in large tokamak reactors. Among them are the availability and relativity low cost of the power supplies, the potential ability to act selectively on the current profile, and the probable absence of operational limits in size, fields or density. The physics of Alfven wave heating in a large tokamak is assessed. Present theoretical understanding of mode coupling and antenna loading is extrapolated to a large machine. The problem of a recessed antenna is analysed. Calculations of loading and discussion of various heating scenarios for the particular case of NET are also presented. (author). 23 refs, 18 figs, 4 tabs

  1. 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)

  2. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  3. 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)

  4. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Purcell, J.R.; Chen, W.; Thomas, R.

    1976-01-01

    Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

  5. Surface tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Kurita, Gen-ichi; Azumi, Masafumi; Takeda, Tatsuoki

    1985-10-01

    Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)

  6. 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

  7. Scattering measurements in Tokamak type devices

    International Nuclear Information System (INIS)

    Matoba, Tohru

    1975-03-01

    Theories, experiments and proposals for light scattering in Tokamak type devices are reviewed. Thomson scattering, measuring method of the current density distribution by scattering and resonance fluorescence are summarily described. These methods may be useful for diagnosis of the fusion plasmas. The report may help planning of the measuring apparatus for the fusion plasmas in future. (auth.)

  8. Tokamak power plant burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1994-06-01

    Experiments show that tokamaks can operate in various fashions. Economic analyses show that steady state is most attractive provided the physics and technology of current drive (CD) can be modestly improved. Even with very conservative CD assumptions a hybrid operating mode seems superior to conventional, simple inductive operation

  9. 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

  10. Runaway electrons in the TRIAM-1 tokamak

    International Nuclear Information System (INIS)

    Satoh, Takemichi; Nakamura, Kazuo; Toi, Kazuo; Nakamura, Yukio; Hiraki, Naoji

    1981-01-01

    Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges. (author)

  11. Runaway electrons in the TRIAM-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

    Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges.

  12. Radiation scattering back to the plasma by the tokamak inner wall in the energy range 50-500 keV during lower hybrid current drive

    International Nuclear Information System (INIS)

    Peysson, Y.

    1990-10-01

    We describe the wall reflectivity by the ratio between the number of photons emerging from the wall and the number entering - and determine the proportion of the reflected contribution to the detected radiations. Various emission profiles and plasma positions in the tokamak chamber have been considered. The contribution of multiple reflections has also be investigated. The wall reflectivity can lead to spurious conclusions for a peaked radial profile in the vicinity of the plasma edge. The next step is devoted to the resolution of the radiation transport equation in solid matter. As an heterogeneous medium is considered - carbon tiles brazed on an iron bulk -, the solution is determined by a numerical Monte-Carlo method. The reflectivity is greatly enhanced by a carbon layer between 50 keV and 150 keV, even for a thickness of one centimeter. The reflectivity is then nearly independent of the energy of the entering photons up to 500 KeV, and lies between 0.15 and 0.4 from a perpendicular to a nearly tangential incidence. Angular corrections have also been considered. Finally, a fully description of the X-ray reflectivity in the high energy range has been performed, taking account of the toroidal geometry and the exact solution of the radiation transport equation. Comparison between theoretical and experimental results obtained with the Tore-Supra high energy X-ray spectrometer has been done. A strong reflectivity effect is observed for the more peripheral line of sight when the plasma emission profile is peaked. There is a good agreement for the total number of detected photons with an energy greater than 100 keV The measured energy spectrum lies up to 200 keV when the photon energy spectrum of the plasma determined from the central chords extends up to 500 keV. A procedure to determine the energy threshold above which the photon energy spectrum is free of the reflected contribution is proposed

  13. Magnetic diagnostics for the proto-eta Tokamak

    International Nuclear Information System (INIS)

    Ferreira, J.L.; Aso, Y.; Ueda, M.; Ferreira, J.G.

    1991-04-01

    This work gives a general view of the magnetic diagnostics rat will be used in the Proto-Eta Tokamak. These diagnostics will be useful tools to measure currents, electric and magnetic fields involved in the plasma magnetic confinement. (author)

  14. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

    Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

    1989-01-01

    The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

  15. Magnetic field structure of experimental high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Deniz, A.V.

    1986-01-01

    The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  17. Effects of isotropic alpha populations on tokamak ballooning stability

    International Nuclear Information System (INIS)

    Spong, D.A.; Sigmar, D.J.; Tsang, K.T.; Ramos, J.J.; Hastings, D.E.; Cooper, W.A.

    1986-12-01

    Fusion product alpha populations can significantly influence tokamak stability due to coupling between the trapped alpha precessional drift and the kinetic ballooning mode frequency. Careful, quantitative evaluations of these effects are necessary in burning plasma devices such as the Tokamak Fusion Test Reactor and the Joint European Torus, and we have continued systematic development of such a kinetic stability model. In this model we have considered a range of different forms for the alpha distribution function and the tokamak equilibrium. Both Maxwellian and slowing-down models have been used for the alpha energy dependence while deeply trapped and, more recently, isotropic pitch angle dependences have been examined

  18. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  19. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  20. Research tokamak system with multi-mode discharges using inverter power supply

    International Nuclear Information System (INIS)

    Kojima, Hiroki; Kobayashi, Masahiro; Takagi, Makoto; Takamura, Shuichi; Tashiro, Kenji

    1999-01-01

    In Current Sustaining Tokamak in Nagoya university (CSTN)-IV research tokamak system using a compact 40kHz pulse width modulation (PWM) inverter power supply, which is controlled through LabVIEW program, we construct a new tokamak discharge system with multi-mode including a stable alternating current discharge and a high-repetition high-duty one. These discharge modes can be operated continuously for as long as 60sec. The continuous discharge with long duration is able to simulate the important physical and chemical processes of long time discharges in fusion devices, in which the heat load to the wall and the particle balance in the plasma-wall system are crucial topics in order to realize a long pulse fusion reactor, like ITER. Employing ergodic divertor (ED) is one of tools to control the particle balance and the heat load to the wall. In addition, we installed another inverter power supply to generate a rotating magnetic perturbation for dynamic ergodic divertor (DED) with the appropriate measurement system so that we may carry out experiments on heat and particle control with DED at long time operation. (author)