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Sample records for tokamak current drive

  1. 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)

  2. 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

  3. Current drive for spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

    Very low aspect ratio spherical tokamaks have proved to have some very useful and remarkable properties including very high values of the plasma pressure to magnetic field pressure. Following the construction of the Start tokamak, a number of such configurations have been constructed. One of the difficulties encountered is in providing sufficient inductive current drive due to the competing requirements of the need to keep the aspect ratio low and providing the space for the central current-carrying rod with an internal inductive coil. An alternative current drive technique would be very useful. In a parallel development it has been shown that a rotating magnetic field can drive a significant non-linear Hall current in a spherical plasma. Successful experiments of this concept have been made with a device called the Rotamak. In its original configuration this device was a field reversed configuration without a toroidal magnetic field but with a vertical field to establish the magnetic hydrodynamical equilibrium. However, recent modifications have shown that increased current can be driven if a central current-carrying rod is used to provide an applied toroidal field. The new Rotamak has then a spherical tokamak magnetic field structure. This work will present new calculations which model the above structure and include the effect of the applied toroidal field in addition to the steady vertical field and the rotating (current-drive) magnetic field. The problem is fully three dimensional and non-linear and involves the application of interesting computational techniques. The potential of using the rotating field current drive technique for spherical tokamaks will be evaluated

  4. Current drive in high density tokamak plasma

    International Nuclear Information System (INIS)

    Sakaguchi, Seiichiro; Jotaki, Eriko; Kawasaki, Shoji; Moriyama, Shin-ichi; Nagao, Akihiro; Nakamura, Kazuo; Nakamura, Yukio; Hiraki, Naoki; Itoh, Satoshi

    1989-01-01

    Current drive in high density tokamak plasma is investigated, with special attention given to mode conversion and proximity conditions that characterize the propagation of electromagnetic waves in the case of current drive by lower hybrid waves. A simple model is used to evaluate the current drive efficiency, and its dependence on various parameters associated with equipment is investigated to provide information required in designing experimental equipment. A strong troidal magnetic field is necessary to produce high density plasma, and incident electromagnetic waves should have a high frequency to prevent the mode conversion, suggesting that a high frequency and a strong troidal field are essential to permit desirable propagation of incident electromagnetic waves. The evaluation of the current drive efficiency shows that the proximity conditions and the power spectrum of the lower hybrid waves entering the plasma are of importance. The average refraction factor in the direction of the troidal field should be larger than but close to that determined from the proximity conditions in order to increase the drive efficiency. As the intensity of the troidal field increases, the refraction factor determined from the proximity conditions decreases, leading to an increase in the drive efficiency. (N.K.)

  5. Current drive in spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

    The early experiments on a spherical rotamak showed that a rotating magnetic field could be used to drive substantial currents and create a compact torus magnetic field configuration. The theoretical analysis of the spherical rotamak has been essentially confined to this class. Recent experiments on the Flinders Rotamak-ST have included a toroidal field, produced by a current-carrying central rod, with encouraging results; for it has been shown that an enhanced current can be driven with this configuration which is the equivalent of a spherical tokamak. This paper will be devoted to a theoretical and computational analysis of this situation. We use a model where the rotating magnetic field is applied to a spherical plasma, with the rotating field oriented parallel to the equatorial plane, taken to be the x-y plane. In our model the ions form a uniform background and the frequency of the rotating Held is very much less than the electron cyclotron frequency (with respect to the rotating field strength) and very much greater than the ion cyclotron frequency. This condition is satisfied by the rotamak experiments

  6. 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.)

  7. Fast wave current drive in reactor scale tokamak plasmas

    International Nuclear Information System (INIS)

    Becoulet, A.; Moreau, D.; Saoutic, B.

    1991-01-01

    The possibility for driving current in large tokamak plasmas using the fast magnetosonic wave is analysed in terms of linear propagation-absorption, and also in terms of quasilinear absorption through an hamiltonian analysis of the wave-particle interaction. The tokamak geometry is shown to strongly influence the capability for the fast wave to sustain a significant part of the toroidal current. Synergetic effects with other scenarios are also discussed

  8. 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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Batchelor, D.B.; Baity, F.W.; Carter, M.D. [and others

    1994-12-31

    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.

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. Novel current drive experiments on the CDX-U, HIT, and DIII-D Tokamaks

    International Nuclear Information System (INIS)

    Ono, M.; Forest, C.B.; Hwang, Y.S.; Armstrong, R.J.; Choe, W.; Darrow, D.S.; Greene, G.; Jones, T.; Schaffer, M.J.; Hyatt, A.W.; Pinsker, R.I.; Staebler, G.M.; Stambaugh, R.D.; Strait, E.J.; Greene, K.L.; Leuer, J.A.; Lohr, J.M.

    1992-01-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges have been developed on the CDX-U, HIT, and DIII-D Tokamaks. On CDX-U, a new, non-inductive current drive technique utilizing fully internally generated pressure driven currents has been demonstrated. The measured current density profile shows a non-hollow profile which agrees with a modeling calculation including helicity conserving non-classical current transport providing the ''seed current''. Another current drive concept, dc-helicity injection, has been investigated on, CDX-U, HIT and DIII-D. This method utilizes injection of magnetic helicity via low energy electron currents, maintaining the plasma current through helicity conserving relaxiation. In these experiments, non-ohmic tokamak plasmas were formed and maintained in the tens of kA range

  18. On current drive by Ohkawa mechanism of electron cyclotron wave in large inverse aspect ratio tokamaks

    Science.gov (United States)

    Zheng, Pingwei; Gong, Xueyu; Lu, Xingqiang; He, Lihua; Cao, Jingjia; Huang, Qianhong; Deng, Sheng

    2018-03-01

    A localized and efficient current drive method in the outer-half region of the tokamak with a large inverse aspect ratio is proposed via the Ohkawa mechanism of electron cyclotron (EC) waves. Further off-axis Ohkawa current drive (OKCD) via EC waves was investigated in high electron beta β e HL-2M-like tokamaks with a large inverse aspect ratio, and in EAST-like tokamaks with a low inverse aspect ratio. OKCD can be driven efficiently, and the driven current profile is spatially localized in the radial region, ranging from 0.62 to 0.85, where the large fraction of trapped electrons provides an excellent advantage for OKCD. Furthermore, the current drive efficiency increases with an increase in minor radius, and then drops when the minor radius beyond a certain value. The effect of trapped electrons greatly enhances the current driving capability of the OKCD mechanism. The highest current drive efficiency can reach 0.183 by adjusting the steering mirror to change the toroidal and poloidal incident angle, and the total driven current by OKCD can reach 20–32 kA MW‑1 in HL-2M-like tokamaks. The current drive is less efficient for the EAST-like scenario due to the lower inverse aspect ratio. The results show that OKCD may be a valuable alternative current drive method in large inverse aspect ratio tokamaks, and the potential capabilities of OKCD can be used to suppress some important magnetohydrodynamics instabilities in the far off-axis region.

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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)

  4. 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.

    2001-01-01

    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)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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)

  14. Electrical conductivity and electron cyclotron current drive efficiencies for non-circular flux surfaces in tokamaks

    International Nuclear Information System (INIS)

    O'Brien, M.R.

    1989-01-01

    As is well known, the presence of electron trapping can strongly reduce the electrical conductivity and rf current drive efficiencies of tokamak plasmas. For example, the conductivity (in the low collisionality limit) of a flux surface with inverse aspect ratio ε=0.1 is approximately one half of the Spitzer conductivity (σ sp )for uniform magnetic fields. Previous estimates of these effects have assumed that the variation of magnetic field strength around a flux surface is given by the standard form for circular flux surfaces. (author) 11 refs., 4 figs

  15. 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

  16. The ergodic limit of multipass absorption for fast wave current drive in tokamaks

    Science.gov (United States)

    Kupfer, K.; Forest, C. B.; Petty, C. C.; Pinsker, R. I.

    1994-12-01

    In many parameter regimes of interest for fast wave current drive (FWCD) in tokamaks, direct absorption of the fast wave by resonant electrons is a weak process and multipass absorption is an important issue. Although both full wave codes and ray tracing codes have been developed to model FWCD, in the multipass regime these tools are computationally intensive, and yield little insight into the nature of the solutions. In this work, an alternative approach is considered. Based on the wave kinetic equation, a natural limit emerges for the multipass regime, where wave energy density, convected along stochastic ray trajectories, uniformly fills the entire accessible phase space. In this ergodic, weak damping limit, the absorbed power density and corresponding wave-driven current density are readily obtained by calculating the appropriate set of one-dimensional k-space integrals at every point in configuration space. The method is used here to model FWCD on the DIII-D tokamak [R. I. Pinsker and the DIII-D Team, Plasma Physics and Controlled Nuclear Fusion Research 1992 (International Atomic Energy Agency, Vienna, 1993), Vol. 1, p. 683]. An example for reactor-grade plasma parameters is also considered.

  17. 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)

  18. 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

  19. 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

  20. Simulations of current density profile control using lower hybrid current drive in the TdeV tokamak

    International Nuclear Information System (INIS)

    Fuchs, V.; Bonoli, P.T.; Shkarofsky, I.P.; Cote, A.; Demers, Y.; Janicki, C.

    1995-01-01

    The physics basis of a simulation model for lower hybrid current drive (LHCD) is discussed. Issues associated with LH power deposition - wave propagation, mode conversion and cut-offs in toroidal geometry, as well as linear and quasi-linear Landau damping - are analysed. A simulation model (ACCOME) is applied to the LHCD experiment now operating on the Tokamak de Varennes (TdeV). The profiles and values of density and temperature needed as inputs to ACCOME are taken from the experiment. The predictions of current density and loop voltage from ACCOME are then compared with experimental LHCD results. Possible LH current profile control experiments are also analysed for TdeV using composite LH spectra to control the location of RF power deposition. Finally, the relevance of these current profile control results to future devices is discussed and an example is shown for ITER-like parameters. (author). 44 refs, 15 figs, 4 tabs

  1. 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.)

  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. High power lower hybrid current drive experiment in TORE SUPRA tokamak

    International Nuclear Information System (INIS)

    Peysson, Y.

    2001-01-01

    A review of the Lower Hybrid (LH) current drive experiments carried out on the TORE SUPRA tokamak is presented. This work highlights the issues for an effective application of the LH wave at high power in reactor relevant conditions. Very promising performances have been obtained with the new launcher that is designed to couple up to 4 MW during 1000 s at a power density of 25 MWm -2 . The heat load on the guard limiter of the antenna and the fast electron acceleration in the near electric field of the grill mouth remain at a low level, while the mean reflection coefficient never exceeds 10%. The powerful diagnosis capabilities of the hard x-ray (HXR) fast electron bremsstrahlung tomography has led to significant progresses in the understanding of the LH wave dynamics. The role of the fastest electrons driven by the LH wave is clearly identified. From HXR measurements, an increase of the LH current drive efficiency with the plasma current is predicted and confirmed by a direct determination at zero loop voltage. LH power absorption is observed to be off-axis in almost all plasma conditions, and its radial width clearly depends of antenna phasing conditions. A correlation between the HXR profiles and the onset of an improved core confinement is identified in fully non-inductive discharges. This regime ascribed to some vanishing of the magnetic shear is found to be transient and usually ends when the minimum of the safety factor becomes very close to 2, leading to a large MHD activity. Experimental observations and numerical simulations suggest that LH power is absorbed in a few number of passes. However, besides toroidal mode coupling, additional mechanisms may likely contribute to a spectral broadening to the LH wave. (author)

  4. 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)

  5. 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)

  6. 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)

  7. 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)

  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. 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)

  10. 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)

  11. Non inductive formation of an extremely overdense spherical Tokamak by electron Bernstein wave heating and current drive on LATE

    Directory of Open Access Journals (Sweden)

    Uchida Masaki

    2015-01-01

    Full Text Available An extremely overdense special Tokamak plasma has been non-inductively formed and maintained by electron Bernstein (EB wave heating and current drive in the Low Aspect ratio Torus Experiment (LATE device. The plasma current reaches 12 kA and the line-averaged electron density exceeds 7 times the plasma cut off density by injecting a 2.45 GHz microwave power of 60 kW. Such a highly overdense plasma is obtained when the upper hybrid resonance layer lies to the higher field side of the 2nd harmonic ECR layer, which may realize a good coupling to EB waves at their first propagation band. The effect of the injection polarization on the mode conversion rate to EB waves at the extremely overdense regime has been investigated and an improvement in the plasma current is observed.

  12. 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

  13. 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.))

  14. Theoretical studies of lower hybrid current drive and ion-cyclotron heating in tokamaks

    International Nuclear Information System (INIS)

    Perkins, F.W.; Valeo, E.J.; Eder, D.C.

    1985-02-01

    A computational model for PLT lower hybrid current drive and ramp-up experiments combines a parallel velocity Fokker-Planck treatment of lower hybrid current drive with minor radius flux diffusion and toroidal ray-tracing wave propagation. Computational and experimental results are in good accord. Analytic solutions of the two-dimensional velocity space (v/sub perpendicular/, v/sub parallel/) diffusion problem give values of the current drive parameter J/P/sub d/ which agree with numerical results, both relativistically and nonrelativistically. Turning to ICRF heating, two new all-metal antenna designs will permit power flux up to 10 kW/cm 2 . A full wave solution to the magnetosonic wave equation, based on the parabolic method, yields cylindrical convergence and treats the diffraction limitation on intensity correctly. Mode conversion with energy absorption has been added to the BALDUR ICRF modeling code. A Fokker-Planck treatment of high energy ion tail formation by ICRF finds that enhanced thermonuclear reactivity can occur

  15. 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.)

  16. 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.

  17. Quasilinear description of heating and current drive in tokamaks by means of test particle Fokker-Planck equation

    International Nuclear Information System (INIS)

    Faulconer, D.W.; Evrard, M.P.

    1991-01-01

    The Fokker-Planck equation is employed to describe wave heating and current drive in fields which are strong enough to distort the zero-order velocity distribution in the face of Maxwellianization by collisions. The effect of the imposed HF wave enters this equation as an additional diffusive term alongside the usual collision term, with a diffusion coefficient, D, calculated from quasilinear theory. In order to account for the irregular motion of trapped and passing particles in a tokamak, this equation is 'bounce averaged' over this motion. Such approaches employ either of two general procedures: 1) averaging of the D which follows from locally homogeneous theory or 2) deriving D from a quasilinear theory which employs the motion of individual particles in the inhomogeneous tokamak magnetic field. Considering low collisionality, we adopt the latter approach as more fundamental in allowing retention of wave-particle phase effects which can limit heating and current drive due to the coherent HF perturbation. An earlier model of the second type which incorporates this limitation has adopted a Hamiltonian formalism, deriving D by proceeding through the Vlasov equation written in action-angle variables. In contrast to this work which uses the kinetic equation, we derive D directly, employing a test particle approach along the lines of work on electron cyclotron resonance. The full Hamiltonian formalism is not used, nor is modal decomposition of poloidal spatial dependence imposed, these points favoring adaptation to diverse field geometries and wave codes (e.g. the SPRUCE 2-D finite difference full wave code which uses Fourier analysis only in toroidal angle). (author) 3 refs

  18. 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

  19. 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)

  20. Lower hybrid current drive system operated with two klystrons in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Rao Jun; Li Xiaodong; Liu Yong; Xuan Weimin; Yuan Yong; Yang Maoyong; Xie Lifeng

    1999-03-01

    The lower hybrid current drive system on HL-1M was improved significantly in the past two years. The output power for one klystron was increased to 500 kW from previous 200 kW. In order to carry out LHCD experiments with higher RF power, the operation mode of the two klystrons in parallel was used. Some problems were solved, which involved frequency, balance of powers and phase control. With this operation mode, maximum output power of 850 kW was realized, and a series of experiment results were obtained

  1. 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

  2. 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.

  3. 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

  4. Lower hybrid heating and current drive, and ion cyclotron heating experiments on the Alcator C and the Versator II tokamaks

    International Nuclear Information System (INIS)

    Porkolab, M.; Blackwell, B.; Bonoli, P.

    1984-08-01

    Lower hybrid electron Landau heting experiments have been carried out on the Alcator C tokamak at densities 0.8 x 10 14 less than or equal to anti n(cm -3 ) less than or equal to 1.7 x 10 14 , and at B = 7 T - 11 T. Using SiC coated graphite limiters, upon injection of 850 kW of power at a density of anti n/sub e/ approx. = 1.3 x 10 14 cm -3 in deuterium typical temperature increases of ΔTi approx. = 0.7 keV, ΔT/sub e/ approx. = 1.0 keV were observed. Current drive and ramping experiments have also been carried out at densities 10 13 less than or equal to anti n/sub e/(cm -3 ) less than or equal to 10 14 . The maximum current drive efficiency, defined as eta = anti n(10 14 cm -3 )I(MA)R(m)/P(MW), was eta approx. = 0.12 at B = 10 T and eta approx. = 0.08 at B = 8 T. Current ramping experiments have resulted in ramping rates of ΔI/Δt approx. = 400 kA/sec with P/sub rf/ approx. = 860 kW at a density of anti n/sub e/ approx. = 3.0 x 10 13 cm -3 . Initial results from the Alcator C 180 MHz ICRF heating program are reported. Significant heating results were obtained in the hydrogen minority regime: upon injection of 400 kW of rf power the ion temperature rose by as much as 600 eV at a density of anti n/sub e/ less than or equal to 1.9 x 10 14 cm -3 . In Versator II, during lower hybrid current drive experiments the global particle confinement increased by approximately a factor of two above the ohmic discharge value when the Parail-Pogutse electron tail instability was stabilized

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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

  10. Synergy in RF Current Drive

    International Nuclear Information System (INIS)

    Dumont, R.J.; Giruzzi, G.

    2005-01-01

    Auxiliary methods for efficient non-inductive current drive in tokamaks generally involve the interaction of externally driven waves with superthermal electrons. Among the possible schemes, Lower Hybrid (LH) and Electron Cyclotron (EC) current drive have been so far the most successful. An interesting aspect of their combined use is the fact that since they involve possibly overlapping domains in velocity and configuration spaces, a synergy between them is expected for appropriate parameters. The signature of this effect, significant improvement of the EC current drive efficiency, results from a favorable interplay of the quasilinear diffusions induced by both waves. Recently, improvements of the EC current drive efficiency in the range of 2-4 have been measured in fully non-inductive discharges in the Tore Supra tokamak, providing the first clear evidence of this effect in steady-state conditions. We present here the experimental aspects of these discharges. The associated kinetic modeling and current state of understanding of the LH-EC synergy phenomenon are also discussed. (authors)

  11. Synergy in RF Current Drive

    International Nuclear Information System (INIS)

    Dumont, R.J.; Giruzzi, G.

    2005-01-01

    Auxiliary methods for efficient non-inductive current drive in tokamaks generally involve the interaction of externally driven waves with superthermal electrons. Among the possible schemes, Lower Hybrid (LH) and Electron Cyclotron (EC) current drive have been so far the most successful. An interesting aspect of their combined use is the fact that since they involve possibly overlapping domains in velocity and configuration spaces, a synergy between them is expected for appropriate parameters. The signature of this effect, significant improvement of the EC current drive efficiency, results from a favorable interplay of the quasilinear diffusions induced by both waves. Recently, improvements of the EC current drive efficiency in the range of 2-4 have been measured in fully non-inductive discharges in the Tore Supra tokamak, providing the first clear evidence of this effect in steady-state conditions. We present here the experimental aspects of these discharges. The associated kinetic modeling and current state of understanding of the LH-EC synergy phenomenon are also discussed

  12. 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

  13. Stability, current drive and heating, energetic particles

    International Nuclear Information System (INIS)

    Razumova, K.

    2001-01-01

    The paper summarizes the results presented at the conference Fusion Energy 2000 (FEC 2000) in relation to the following subjects: 1. The possibility of realizing plasma parameters for ITER needs, advanced regimes in tokamaks and stellarators. 2. Stability of plasmas with an appreciable component of fast particles. 3. Low aspect ratio tokamaks. 4. New results with auxiliary heating and current drive methods. 5. β limit and neoclassical tearing mode (NTM) stabilization. 6. Internal transport barriers. (author)

  14. Eddy currents in the Alcator Tokamak

    International Nuclear Information System (INIS)

    Schram, D.C.; Rem, J.

    1975-03-01

    A one-dimensional model of an aircore transformer has been developed through which it is possible to analyze the effect of eddy currents in the primary windings and of similar currents in the field coils for the toroidal magnetic field, on the time dependence of the current in a Tokamak experiment. The model is applied to the 'Alcator' Tokamak at MIT and its accuracy is tested by comparing analytical results for the harmonic behaviour of the transformer, with experimental data. The time-dependent behaviour of the plasma current for a constant plasma resistance shows that eddy currents in the primary windings will lead to a reduction of 8% of the current maximum. The eddy currents in the 'Bitter' coils are found to affect predominantly the initial current rise; they lead to a steepening of the current rise. Finally, the influence of the time dependence of the plasma resistance is investigated

  15. Modeling tokamak discharges with current holes

    International Nuclear Information System (INIS)

    Jensen, T.H.

    2002-01-01

    Tokamaks with current holes [T.S. Taylor, et al., Bull. Am. Phys. Soc. 43 (1998) 1783; N.C. Hawkes, et al., Phys. Rev. Lett. 87 (2001) 115001; T. Fujita, et al., Phys. Rev. Lett. 87 (2001) 245001] are interesting, in part, because discharges with true current holes do not consume poloidal flux. The modeling of this Letter suggests that under steady-state conditions their currents may be driven by radial flow of plasma resulting from neutral beam injection

  16. Turbulent current drive mechanisms

    Science.gov (United States)

    McDevitt, Christopher J.; Tang, Xian-Zhu; Guo, Zehua

    2017-08-01

    Mechanisms through which plasma microturbulence can drive a mean electron plasma current are derived. The efficiency through which these turbulent contributions can drive deviations from neoclassical predictions of the electron current profile is computed by employing a linearized Coulomb collision operator. It is found that a non-diffusive contribution to the electron momentum flux as well as an anomalous electron-ion momentum exchange term provide the most efficient means through which turbulence can modify the mean electron current for the cases considered. Such turbulent contributions appear as an effective EMF within Ohm's law and hence provide an ideal means for driving deviations from neoclassical predictions.

  17. Theory of beat-wave current drive

    Energy Technology Data Exchange (ETDEWEB)

    Mendonca, J.T.; Galvao, R.M.O.

    1986-06-01

    The beat-wave scheme for current drive is studied in the frame of plasma weak-turbulence theory. The value of the driven current is limited by a quasi-linear diffusion mechanism. The one-dimensional problem, corresponding to the resonant excitation of electron plasma oscillations by two beating electro-magnetic beams is discussed in detail. The results may be relevant to the continuous operation of tokamak discharges.

  18. 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 ...

  19. Plasma current profile during current reversal in a tokamak

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  20. Analysis of fast wave current drive from the Alcyon code

    International Nuclear Information System (INIS)

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

    1991-01-01

    Fast Wave Current Drive simulations have been performed with the 2-D full wave code ALCYON. These simulations include the computation of the RF field in a Tokamak geometry, for a given launched power spectrum, a linear estimation of the driven current profile using this RF field, and the numerical derivation of the quasilinear diffusion operator for a complete Fokker-Planck calculation. Results concerning the ITER, and the DIII-D Tokamaks are presented and discussed

  1. Fast wave current drive above the slow wave density limit

    International Nuclear Information System (INIS)

    McWilliams, R.; Sheehan, D.P.; Wolf, N.S.; Edrich, D.

    1989-01-01

    Fast wave and slow wave current drive near the mean gyrofrequency were compared in the Irvine Torus using distinct phased array antennae of similar principal wavelengths, frequencies, and input powers. The slow wave current drive density limit was measured for 50ω ci ≤ω≤500ω ci and found to agree with trends in tokamaks. Fast wave current drive was observed at densities up to the operating limit of the torus, demonstrably above the slow wave density limit

  2. 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

  3. Transport effects on current drive efficiency and localisation

    International Nuclear Information System (INIS)

    Cox, M.; McKenzie, J.S.; O'Brien, M.R.

    1990-01-01

    In this paper we discuss the effects of radial transport of electrons on the efficiency and profiles of a radiofrequency and Ohmic current drive in tokamaks. It has been recognised theoretically and experimentally that such processes can reduce the potential current drive efficiency of both Lower Hybrid (LH) and Electron Cyclotron Resonance Heating (ECRH) in tokamaks in which the energy confinement time (τ E ) is comparable with or less than the collision time of the heated electrons. Also, even in tokamaks in which this condition is not satisfied, radial transport can broaden the driven current profile and perhaps limit the effectiveness of the use of current drive for tailoring the current profile to control localised MHD modes. Here we solve numerically for the perturbed current-carrying component of the electron distribution function produced by balancing collisional, heating and transport processes. Three cases are considered: current drive by LH waves, the ECRH current drive experiments on CLEO in which the discrepancy between observed and predicted driven current was attributed to these effects; and the effect on Ohmic current drive. (author) 7 refs., 4 figs

  4. 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

  5. High Field Side Lower Hybrid Current Drive Simulations for Off- axis Current Drive in DIII-D

    Science.gov (United States)

    Wukitch, S. J.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Holcomb, C.; Pinsker, R. I.

    2017-10-01

    Efficient off-axis current drive scalable to reactors is a key enabling technology for developing economical, steady state tokamak. Previous studies have focussed on high field side (HFS) launch of lower hybrid current drive (LHCD) in double null configurations in reactor grade plasmas and found improved wave penetration and high current drive efficiency with driven current profile peaked near a normalized radius, ρ, of 0.6-0.8, consistent with advanced tokamak scenarios. Further, HFS launch potentially mitigates plasma material interaction and coupling issues. For this work, we sought credible HFS LHCD scenario for DIII-D advanced tokamak discharges through utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D) constrained by experimental considerations. For a model and existing discharge, HFS LHCD scenarios with excellent wave penetration and current drive were identified. The LHCD is peaked off axis, ρ˜0.6-0.8, with FWHM Δρ=0.2 and driven current up to 0.37 MA/MW coupled. For HFS near mid plane launch, wave penetration is excellent and have access to single pass absorption scenarios for variety of plasmas for n||=2.6-3.4. These DIII-D discharge simulations indicate that HFS LHCD has potential to demonstrate efficient off axis current drive and current profile control in DIII-D existing and model discharge.

  6. Tokamak formation and sustainment by tokamak injection

    International Nuclear Information System (INIS)

    Farengo, R.; Jarboe, T.R.

    1991-01-01

    The authors propose here a new helicity injection method for tokamak formation and sustainment that has high efficiency, conserves toroidal symmetry and is inductively driven. The basic idea is to inject a small tokamak (source tokamak) into a larger tokamak (steady tokamak). This current drive scheme eliminates the need for the ohmic heating transformer in the steady tokamak allowing the formation of very small aspect ratio tokamaks (Spherical Tori). Thus, steady state operation and high beta can be realized simultaneously. The method can also be applied to a larger aspect ratio tokamak and used in conjunction with the standard inductive formation technique. In order to allow for translation the ohmic heating coil used to produce the source tokamaks must be fed from one end (as in the CSS device) and the toroidal field coil must link both tokamaks. After formation the source tokamaks are accelerated towards the steady tokamak by a mirror field and the tension of the field lines that wrap around both tokamaks (producing a doublet type configuration). In a tokamak the helicity is proportional to the current. This indicates that (assuming helicity is conserved during the merging process) a steady state situation will result if the helicity supplied by the source tokamaks is equal to the helicity dissipated by the steady tokamak. Assuming that source tokamaks of helicity K s are injected with frequency f, the steady state condition can be written as: fK s = 2V t Ψ t = K t /τ K where V t , Ψ t , K t and τ K are the ohmic loop voltage, toroidal flux, helicity and helicity decay time of the steady tokamak. A simple calculation shows that the DIII-D tokamak could be sustained by injecting source tokamaks with R = 1.20 m, a = 0.23 m and I = 151 kA at a frequency of 120 Hz. 1 ref

  7. Current drive experiments at the electron cyclotron frequency

    International Nuclear Information System (INIS)

    Erckmann, V.; Gasparino, U.; Maassberg, H.; Renner, H.; Tutter, M.; Kasparek, W.; Mueller, G.A.; Schueller, P.G.; Thumm, M.

    1991-01-01

    The experimental investigation of non-inductive current drive by electromagnetic waves in the electron cyclotron range of frequencies and the comparison with theoretical predictions attracts increasing interest in both, tokamak as well as stellarator research. In spite of the low current drive efficiency (compared to Lower Hybrid Current Drive) Electron Cyclotron Current Drive (ECCD) is a candidate for MHD-mode control and current profil shaping in tokamaks and stellarators due to the high localization of the driven currents and the capability to drive currents in the plasma centre in large machines. ECCD is an appropriate tool for the control of the pressure effects on the profile of the rotational transform, particularly the bootstrap current in stellarators. This is a crucial condition to maintain good confinement properties in low shear configurations such as W VII-AS. Basic experimental investigations were performed at the W VII-AS stellarator, where the small EC-driven currents are not masked by large inductively driven currents as in tokamaks. The theoretical treatment of ECCD in stellarators would require a Fokker Planck solution in full phase space taking into account the complex magnetic field configuration (trapped particles, loss cone effects) which is untractable. In a first approach, we have compared our experimental data to a simple analytical current drive model (linearized in slab geometry) which is incorporated in a ray-tracing code. In a second step, we have analysed the sensitivity of the model with respect to simplified assumptions on trapped particles and quasi-linear effects. (orig.)

  8. Reconstruction of plasma current profile of Tokamaks using genetic algorithm

    International Nuclear Information System (INIS)

    Kishimoto, Maki

    1996-01-01

    A new method to reconstruct plasma shape and plasma current distribution from magnetic measurements using a combinatorial optimization technique are proposed. The reconstruction of plasma current profile from magnetic measurements is regarded as an optimum allocation problem of currents into cross section of the vacuum vessel of the Tokamak. In order to solve this optimization problem, we use a genetic algorithm. The effectiveness of this method is shown by the application of this technique to JT-60U plasmas. (author)

  9. Current drive for rotamak plasmas

    Indian Academy of Sciences (India)

    Abstract. Experiments which have been undertaken over a number of years have shown that a rotating magnetic field can drive a significant non-linear Hall current in a plasma. Successful experiments of this concept have been made with a device called rotamak. In its original configuration this device was a field reversed ...

  10. Theory of current-drive in plasmas

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. 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

  12. Current challenges in autonomous driving

    Science.gov (United States)

    Barabás, I.; Todoruţ, A.; Cordoş, N.; Molea, A.

    2017-10-01

    Nowadays the automotive industry makes a quantum shift to a future, where the driver will have smaller and smaller role in driving his or her vehicle ending up being totally excluded. In this paper, we have investigated the different levels of driving automatization, the prospective effects of these new technologies on the environment and traffic safety, the importance of regulations and their current state, the moral aspects of introducing these technologies and the possible scenarios of deploying the autonomous vehicles. We have found that the self-driving technologies are facing many challenges: a) They must make decisions faster in very diverse conditions which can include many moral dilemmas as well; b) They have an important potential in reducing the environmental pollution by optimizing their routes, driving styles by communicating with other vehicles, infrastructures and their environment; c) There is a considerable gap between the self-drive technology level and the current regulations; fortunately, this gap shows a continuously decreasing trend; d) In case of many types of imminent accidents management there are many concerns about the ability of making the right decision. Considering that this field has an extraordinary speed of development, our study is up to date at the submission deadline. Self-driving technologies become increasingly sophisticated and technically accessible, and in some cases, they can be deployed for commercial vehicles as well. According to the current stage of research and development, it is still unclear how the self-driving technologies will be able to handle extreme and unexpected events including their moral aspects. Since most of the traffic accidents are caused by human error or omission, it is expected that the emergence of the autonomous technologies will reduce these accidents in their number and gravity, but the very few currently available test results have not been able to scientifically underpin this issue yet. The

  13. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    conditions. In the ohmic discharges of the SINP tokamak, we observed 'negative edge plasma currents' at the edge region (r/a ∼ 0.7–0.8, where a = 7.5 cm is the ... The horizontal plasma column (figure 1c) is observed to move inwards/ ... (b) 3 by 3 NaI(Tl) output (HX) in MeV, (c) horizontal plasma column motion ( hor).

  14. 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

  15. Numerical modeling of lower hybrid heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Valeo, E.J.; Eder, D.C.

    1986-03-01

    The generation of currents in toroidal plasma by application of waves in the lower hybrid frequency range involves the interplay of several physical phenomena which include: wave propagation in toroidal geometry, absorption via wave-particle resonances, the quasilinear generation of strongly nonequilibrium electron and ion distribution functions, and the self-consistent evolution of the current density in such a nonequilibrium plasma. We describe a code, LHMOD, which we have developed to treat these aspects of current drive and heating in tokamaks. We present results obtained by applying the code to a computation of current ramp-up and to an investigation of the possible importance of minority hydrogen absorption in a deuterium plasma as the ''density limit'' to current drive is approached.

  16. Numerical modeling of lower hybrid heating and current drive

    International Nuclear Information System (INIS)

    Valeo, E.J.; Eder, D.C.

    1986-03-01

    The generation of currents in toroidal plasma by application of waves in the lower hybrid frequency range involves the interplay of several physical phenomena which include: wave propagation in toroidal geometry, absorption via wave-particle resonances, the quasilinear generation of strongly nonequilibrium electron and ion distribution functions, and the self-consistent evolution of the current density in such a nonequilibrium plasma. We describe a code, LHMOD, which we have developed to treat these aspects of current drive and heating in tokamaks. We present results obtained by applying the code to a computation of current ramp-up and to an investigation of the possible importance of minority hydrogen absorption in a deuterium plasma as the ''density limit'' to current drive is approached

  17. Fast wave current drive system design for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D. (General Atomics, San Diego, CA (United States)); Baity, F.W.; Hoffman, D.J.; Taylor, D.J. (Oak Ridge National Lab., TN (United States)); Arnold, W.; Martin, S. (ANT-Nachrichtentechnik GmbH, Backnang (Germany))

    1992-09-01

    DIII-D has a major effort underway to develop the physics and technology of fast wave electron heating and current drive in conjunction with electron cyclotron heating. The present system consists of a four strap antenna driven by one 2 MW transmitter in the 32--60 MHz band. Experiments have been successful in demonstrating the physics of heating and current drive. In order to validate fast wave current drive for future machines a greater power capability is necessary to drive all of the plasma current. Advanced tokamak modeling for DIII-D has indicated that this goal can be met for plasma configurations of interest (i.e. high [beta] VH-mode discharges) with 8 MW of transmitter fast wave capability. It is proposed that four transmitters drive fast wave antennas at three locations in DIII-D to provide the power for current drive and current profile modification. As the next step in acquiring this capability, two modular four strap antennas are in design and the procurement of a high power transmitter in the 30--120 MHz range is in progress. Additionally, innovations in the technology are being investigated, such as the use of a coupled combine antenna to reduce the number of required feedthroughs and to provide for parallel phase velocity variation with a relatively small change in frequency, and the use of fast ferrite tuners to provide millisecond timescale impedance matching. A successful test of a low power fast ferrite prototype was conducted on DIII-D.

  18. Fast wave current drive system design for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D. [General Atomics, San Diego, CA (United States); Baity, F.W.; Hoffman, D.J.; Taylor, D.J. [Oak Ridge National Lab., TN (United States); Arnold, W.; Martin, S. [ANT-Nachrichtentechnik GmbH, Backnang (Germany)

    1992-09-01

    DIII-D has a major effort underway to develop the physics and technology of fast wave electron heating and current drive in conjunction with electron cyclotron heating. The present system consists of a four strap antenna driven by one 2 MW transmitter in the 32--60 MHz band. Experiments have been successful in demonstrating the physics of heating and current drive. In order to validate fast wave current drive for future machines a greater power capability is necessary to drive all of the plasma current. Advanced tokamak modeling for DIII-D has indicated that this goal can be met for plasma configurations of interest (i.e. high {beta} VH-mode discharges) with 8 MW of transmitter fast wave capability. It is proposed that four transmitters drive fast wave antennas at three locations in DIII-D to provide the power for current drive and current profile modification. As the next step in acquiring this capability, two modular four strap antennas are in design and the procurement of a high power transmitter in the 30--120 MHz range is in progress. Additionally, innovations in the technology are being investigated, such as the use of a coupled combine antenna to reduce the number of required feedthroughs and to provide for parallel phase velocity variation with a relatively small change in frequency, and the use of fast ferrite tuners to provide millisecond timescale impedance matching. A successful test of a low power fast ferrite prototype was conducted on DIII-D.

  19. 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.

  20. Physics of electron cyclotron current drive on DIII-D

    CERN Document Server

    Petty, C C; Harvey, R W; Kinsey, J E; Lao, L L; Lohr, J; Luce, T C; Makowski, M A; Prater, R

    2002-01-01

    OAK A271 PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The measured ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection. The measured ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (Toray-GA) as well as a quasilinear Fokker-Planck model (CQL3D). The experimental ECCD is found to be in better agreement with the more complete Fokker-Planck calculation, especially for cases of high rf power density and/or loop voltage. The narrow width of the measured ECCD profile is consistent with only low levels of radial transport for the current carrying electrons.

  1. Fast wave current drive in DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Callis, R.W.; Chiu, S.C.; deGrassie, J.S.; Forest, C.B.; Freeman, R.L.; Gohil, P.; Harvey, R.W.; Ikezi, H.; Lin-Liu, Y.-R.

    1995-02-01

    The non-inductive current drive from fast Alfven waves launched by a directional four-element antenna was measured in the DIII-D tokamak. The fast wave frequency (60 MHz) was eight times the deuterium cyclotron frequency at the plasma center. An array of rf pickup loops at several locations around the torus was used to verify the directivity of the four-element antenna. 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 discharges with steady plasma current, up to 110 kA of FWCD was inferred from an analysis of the loop voltage, with a maximum non-inductive current (FWCD, ECCD, and bootstrap) of 195 out of 310 kA. The FWCD efficiency increased linearly with central electron temperature. For low current discharges, the FWCD efficiency was degraded due to incomplete fast wave damping. The experimental FWCD was found to agree with predictions from the CURRAY ray-tracing code only when a parasitic loss of 4% per pass was included in the modeling along with multiple pass damping

  2. 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

  3. Optimized calculation of the synergy conditions between electron cyclotron current drive and lower hybrid current drive on EAST

    International Nuclear Information System (INIS)

    Wei Wei; Ding Bo-Jiang; Li Miao-Hui; Zhang Xin-Jun; Wang Xiao-Jie; Peysson, Y; Decker, J; Zhang Lei

    2016-01-01

    The optimized synergy conditions between electron cyclotron current drive (ECCD) and lower hybrid current drive (LHCD) with normal parameters of the EAST tokamak are studied by using the C3PO/LUKE code based on the understanding of the synergy mechanisms so as to obtain a higher synergistic current and provide theoretical reference for the synergistic effect in the EAST experiment. The dependences of the synergistic effect on the parameters of two waves (lower hybrid wave (LHW) and electron cyclotron wave (ECW)), including the radial position of the power deposition, the power value of the LH and EC waves, and the parallel refractive indices of the LHW (N ∥ ) are presented and discussed. (paper)

  4. Scrape-off-layer current and EUV diagnostics and control on the HBT-EP tokamak

    Science.gov (United States)

    Levesque, J. P.; Mauel, M. E.; Bialek, J.; Navratil, G. A.; Delgado-Aparicio, L.; Hansen, C. J.

    2015-11-01

    Non-axisymmetric currents in the scrape-off-layer (SOL) and conducting structure of a tokamak can produce severe forces at high plasma performance, compromising the device's structural integrity. Diagnosing these currents during disruptions is important for extrapolating forces in future machines including ITER. Progress on designing components to measure and control SOL and vessel currents in the HBT-EP tokamak is presented. Movable tiles positioned around limiting surfaces will measure SOL and vessel currents during mode activity and disruptions. Biasable plates at divertor strike points will allow control of field-aligned SOL currents for kink mode control studies and will drive convection in the plasma edge. In-vessel Rogowski coils will measure currents in wall components with high spatial resolution. A planned EUV diagnostic upgrade is also presented. Four sets of 16 poloidal views will allow tomographic reconstruction of plasma emissivity and internal kink mode structure. A separate two-color, 16-chord tangential system will allow reconstruction of temperature profiles versus time. Measurements will be input to HBT-EP's GPU-based feedback system, providing active feedback for kink modes using only optical sensors and both magnetic and edge current actuators. Supported by U.S. DOE Grant DE-FG02-86ER53222.

  5. PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    International Nuclear Information System (INIS)

    PETTY, C.C.; PRATER, R.; LUCE, T.C.; ELLIS, R.A.; HARVEY, R.W.; KINSEY, J.E.; LAO, L.L.; LOHR, J.; MAKOWSKI, M.A.

    2002-01-01

    OAK A271 PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The measured ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection. The measured ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (Toray-GA) as well as a quasilinear Fokker-Planck model (CQL3D). The experimental ECCD is found to be in better agreement with the more complete Fokker-Planck calculation, especially for cases of high rf power density and/or loop voltage

  6. The essential theory of fast wave current drive with full wave method

    International Nuclear Information System (INIS)

    Liu Yan; Gong Xueyu; Yang Lei; Yin Chenyan; Yin Lan

    2007-01-01

    The full wave numerical method is developed for analyzing fast wave current drive in the range of ion cyclotron waves in tokamak plasmas, taking into account finite larmor radius effects and parallel dispersion. the physical model, the dispersion relation on the assumption of Finite Larmor Radius (FLR) effects and the form of full wave be used for computer simulation are developed. All of the work will contribute to further study of fast wave current drive. (authors)

  7. RF current drive by electron cyclotron waves in the presence of magnetic islands

    Energy Technology Data Exchange (ETDEWEB)

    Da Silva Rosa, P.; Giruzzi, G

    1999-11-01

    The influence of the presence of magnetic islands, and the consequent modification of the tokamak magnetic surface topology, on electron current drive is analyzed. To this end, a new 3D Fokker-Planck code has been developed, taking into account the modifications of the magnetic equilibrium topology owing to the presence of the islands. Significant differences between electron cyclotron current drive efficiency with and without island inside the plasma are found, particularly in the case of interaction with locked modes. (authors)

  8. 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

  9. Fast-ion transport and neutral beam current drive in ASDEX upgrade

    DEFF Research Database (Denmark)

    Geiger, B.; Weiland, M.; Jacobsen, Asger Schou

    2015-01-01

    The neutral beam current drive efficiency has been investigated in the ASDEX Upgrade tokamak by replacing on-axis neutral beams with tangential off-axis beams. A clear modification of the radial fast-ion profiles is observed with a fast-ion D-alpha diagnostic that measures centrally peaked profiles...

  10. ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY

    International Nuclear Information System (INIS)

    PRATER, R; PETTY, CC; LUCE, TC; HARVEY, RW; CHOI, M; LAHAYE, RJ; LIN-LIU, Y-R; LOHR, J; MURAKAMI, M; WADE, MR; WONG, K-L

    2003-01-01

    A271 ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY. Experiments on the DIII-D tokamak in which the measured off-axis electron cyclotron current drive has been compared systematically to theory over a broad range of parameters have shown that the Fokker-Planck code CQL3D provides an excellent model of the relevant current drive physics. This physics understanding has been critical in optimizing the application of ECCD to high performance discharges, supporting such applications as suppression of neoclassical tearing modes and control and sustainment of the current profile

  11. Tokamak

    International Nuclear Information System (INIS)

    Meglicki, Z.

    1995-01-01

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs

  12. D-3He tokamak reactor with inductive a.c. and unidirectional current operation

    International Nuclear Information System (INIS)

    Mitarai, O.

    1995-01-01

    We propose inductively operated D- 3 He tokamak reactors (R=9.5m, a=2.8 and 3.2m, B t =10 and 7.5T, and I p =53.6 and 52.5MA respectively) with an a.c. or unidirectional current (UDC) mode, operated in the first stability regime, with a small current drive power of about 500kW in the plasma side, and about 2MW in the transformer side. The high electron temperature T e (0)∼62.5keV together with a high bootstrap current fraction greater than 70% decreases the plasma resistance (to less than 1nΩ) and the loop voltage (less than 10mV range) and then prolongs the pulse length to more than 15h. The ignition characteristics of a D- 3 He tokamak reactor are analyzed by the operation path method on the P ht τ E 2 -T plane and POPCON for different confinement scalings. ((orig.))

  13. Synergy between electron cyclotron and lower hybrid current drive on Tore Supra

    International Nuclear Information System (INIS)

    Giruzzi, G.; Artaud, J.F.; Dumont, R.J.; Imbeaux, F.; Bibet, P.; Berger-By, G.; Bouquey, F.; Clary, J.; Darbos, C.; Ekedahl, A.; Hoang, G.T.; Lennholm, M.; Maget, P.; Magne, R.; Segui, J.L.; Bruschi, A.; Granucci, G.

    2005-01-01

    Improvement (up to a factor ∼ 4) of the electron cyclotron (EC) current drive efficiency in plasmas sustained by lower hybrid (LH) current drive has been demonstrated in stationary conditions on the Tore Supra tokamak. This was made possible by feedback controlled discharges at zero loop voltage, constant plasma current and density. This effect, predicted by kinetic theory, results from a favorable interplay of the velocity space diffusions induced by the two waves: the EC wave pulling low-energy electrons out of the Maxwellian bulk, and the LH wave driving them to high parallel velocities. (author)

  14. Determination of the centre of gravity of the current distribution in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Zoletnik, S.; Montvai, A.

    1986-03-01

    A simple software method is described for measuring the plasma current channel position from Mirnov coil signals on the MT-1 tokamak. Plasma equilibrium calculations are not involved. The method was also applied to unstable tokamak discharges, and examples based on the results are presented. (author)

  15. Technology of fast-wave current drive antennas

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Goulding, R.H.; Haste, G.R.; Ryan, P.M.; Taylor, D.J.; Swain, D.W.; Mayberry, M.J.; Yugo, J.J.

    1989-01-01

    The design of fast-wave current drive (FWCD) antennas combines the usual antenna considerations (e.g., the plasma/antenna interface, disruptions, high currents and voltages, and thermal loads) with new requirements for spectral shaping and phase control. The internal configuration of the antenna array has a profound effect on the spectrum and the ability to control phasing. This paper elaborates on these considerations, as epitomized by a proof-of-principle (POP) experiment designed for the DIII-D tokamak. The extension of FWCD for machines such as the International Thermonuclear Engineering Reactor (ITER) will require combining ideas implemented in the POP experiment with reactor-relevant antenna concepts, such as the folded waveguide. 6 refs., 8 figs

  16. 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

  17. Fast wave current drive on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I. [and others

    1995-07-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as {gamma} = 0.4 {times} 10{sup 18} T{sub eo} (keV) [A/m{sup 2}W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with dear evidence for a toroidally directed wave with antenna phasing set for current drive. There is some experimental evidence for fast wave absorption by energetic beam ions at high cyclotron harmonic resonances.

  18. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.; Forest, C.B.; Ikezi, H.; Prater, R.; Baity, F.W.; Callis, R.W.; Cary, W.P.; Chiu, S.C.; Doyle, E.J.; Ferguson, S.W.; Hoffman, D.J.; Jaeger, E.F.; Kim, K.W.; Lee, J.H.; Lin-Liu, Y.R.; Murakami, M.; ONeill, R.C.; Porkolab, M.; Rhodes, T.L.; Swain, D.W.

    1996-01-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as γ=0.4x10 18 T e0 (keV) [A/m 2 W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with clear evidence for a toroidally directed wave with antenna phasing set for current drive. copyright 1996 American Institute of Physics

  19. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.

    1995-01-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as γ = 0.4 x 10 18 T eo (keV) [A/m 2 W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with dear evidence for a toroidally directed wave with antenna phasing set for current drive. There is some experimental evidence for fast wave absorption by energetic beam ions at high cyclotron harmonic resonances

  20. Electron cyclotron current drive experiments on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    James, R.A. (Lawrence Livermore National Lab., CA (USA)); Giruzzi, G.; Gentile, B. de; Rodriguez, L. (Association Euratom-CEA, Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France)); Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V. (Kurchatov Inst. of Atomic Energy, Moscow (USSR)); Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R. (General Atomics, San Di

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and {tau}{sub E} much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T{sub e}, {eta}{sub e} and Z{sub eff} are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs.

  1. Experimental demonstration of synergy between electron cyclotron and lower hybrid current drive on Tore Supra

    International Nuclear Information System (INIS)

    Artaud, J.F.; Giruzzi, G.; Dumont, R.J.; Imbeaux, F.; Bibet, P.; Bouquey, F.; Clary, J.; Ekedahl, A.; Hoang, G.T.; Lennholm, M.; Magne, R.; Segui, J.L.

    2004-01-01

    Non-inductive current drive (CD) has two main applications in tokamaks: sustainment of a substantial fraction of the toroidal plasma current necessary for the plasma confinement and control of the plasma stability and transport properties by appropriate shaping of the current density profile. For the first kind of applications, lower hybrid (LH) waves are known to provide the highest efficiency (defined as the ratio of the driven current to the injected wave power), although with limited control capability. Conversely, electron cyclotron (EC) waves drive highly localized currents, and are therefore particularly suited for control purposes, but their CD efficiency is much lower than that of LH waves (typically, an order of magnitude in present day experiments). Various calculations have demonstrated an interesting property: the current driven by the simultaneous use of the two waves, I(LH+EC), can be significantly larger than the sum I(LH)+I(EC) of the currents separately driven by the two waves in the same plasma conditions. This property, called synergy effect. The peculiar experimental conditions attainable on the Tore Supra tokamak have allowed the first experimental demonstration of the synergy between EC and LH current drive. The significant improvement of the electron cyclotron current drive (ECCD) efficiency in the presence of low hybrid current drive (LHCD), predicted by kinetic theory and confirmed by stationary experiments on Tore Supra, opens up the possibility of using ECCD as an efficient current profile control tool in LHCD plasmas

  2. Current drive for rotamak plasmas

    Indian Academy of Sciences (India)

    toroidal fields. It is clear that the current driven may depend on the actual experi- mental situation and that increasing either of these fields may not necessarily lead to better performance. Acknowledgement. The author acknowledges the financial support provided through the Australian. Research Council and the Australian ...

  3. Electron cyclotron heating and current drive in toroidal geometry

    Energy Technology Data Exchange (ETDEWEB)

    Kritz, A.H.

    1993-03-01

    The Principal Investigator has continued to work on problems associated both with the deposition and with the emission of electron cyclotron heating power electron cyclotron heating in toroidal plasmas. Inparticular, the work has focused on the use of electron cyclotron heating to stabilize q = 1 and q = 2 instabilities in tokamaks and on the use of electron cyclotron emission as a plasma diagnostic. The research described in this report has been carried out in collaboration with scientists at Princeton, MIT and Livermore. The Principal Investigator is now employed at Lehigh University, and a small group effort on electron cyclotron heating in plasmas has begun to evolve at Lehigh involving undergraduate and graduate students. Work has also been done in support of the electron cyclotron heating and current drive program at the Center for Research in Plasma Physics in Lausanne, Switzerland.

  4. Current-drive by lower hybrid waves in the presence of energetic alpha-particles

    Energy Technology Data Exchange (ETDEWEB)

    Fisch, N.J.; Rax, J.M.

    1991-10-01

    Many experiments have now proved the effectiveness of lower hybrid waves for driving toroidal current in tokamaks. The use of these waves, however, to provide all the current in a reactor is thought to be uncertain because the waves may not penetrate the center of the more energetic reactor plasma, and, if they did, the wave power may be absorbed by alpha particles rather than by electrons. This paper explores the conditions under which lower-hybrid waves might actually drive all the current. 26 refs.

  5. Current-drive by lower hybrid waves in the presence of energetic alpha-particles

    International Nuclear Information System (INIS)

    Fisch, N.J.; Rax, J.M.

    1991-10-01

    Many experiments have now proved the effectiveness of lower hybrid waves for driving toroidal current in tokamaks. The use of these waves, however, to provide all the current in a reactor is thought to be uncertain because the waves may not penetrate the center of the more energetic reactor plasma, and, if they did, the wave power may be absorbed by alpha particles rather than by electrons. This paper explores the conditions under which lower-hybrid waves might actually drive all the current. 26 refs

  6. Conditions for Lower Hybrid Current Drive in ITER

    Science.gov (United States)

    Cesario, R.; Amicucci, L.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Napoli, F.; Tuccillo, A. A.; Galli, A.; Schettini, G.

    2012-12-01

    To control the plasma current profile represents one of the most important problems of the research of nuclear fusion energy based on the tokamak concept, as in the plasma column the necessary conditions of stability and confinement should be satisfied. This problem can be solved by using the lower hybrid current drive (LHCD) effect, which was demonstrated to occur also at reactor grade high plasma densities provided that a proper method should be utilised, as assessed on FTU (Frascati Tokamak Upgrade). This method, based on theoretical predictions confirmed by experiment, produces relatively high electron temperature at the plasma periphery and scrape-off layer (SOL), consequently reducing the broadening of the spectrum launched by the antenna produced by parasitic wave physics of the edge, namely parametric instability (PI). The new results presented here show that, for kinetic profiles now foreseen for the SOL of ITER, PI is expected to hugely broaden the antenna spectrum and prevent any penetration in the core of the coupled LH power. However, considering the FTU method and assuming higher electron temperature at the edge (which would be however reasonable for ITER) the PI-produced spectral broadening would be mitigated, and enable the penetration of the coupled LH power in the main plasma. By successful LHCD effect, the control of the plasma current profile at normalised minor radius of about 0.8 would be possible, with much higher efficiency than that obtainable by other tools. A very useful reinforce of bootstrap current effects would be thus possible by LHCD in ITER.

  7. 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

  8. 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

  9. COMPASS-D magnetic equilibria with LH and NBI current drive

    Czech Academy of Sciences Publication Activity Database

    Hronová-Bilyková, Olena; Fuchs, Vladimír; Pánek, Radomír; Urban, Jakub; Žáček, František; Stöckel, Jan; Voitsekhovitch, I.; Valovič, M.; Fitzgerald, M.

    2006-01-01

    Roč. 56, suppl.B (2006), B24-B30 ISSN 0011-4626. [Symposium on Plasma Physics and Technology/22nd./. Praha, 26.6.2006-29.6.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * COMPASS-D * magnetic equilibrium * ACCOME code * ASTRA code * Neutral Beam Injection * Low Hybrid Current Drive Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.568, year: 2006

  10. 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

  11. Electron cyclotron resonance heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Fidone, I.; Castejon, F.

    1992-07-01

    A brief summary of the theory and experiments on electron- cyclotron heating and current drive is presented. The general relativistic formulation of wave propagation and linear absorption is considered in some detail. The O-mode and the X-mode for normal and oblique propagation are investigated and illustrated by several examples. The experimental verification of the theory in T-10 and D- III-D is briefly discussed. Quasilinear evolution of the momentum distribution and related applications as, for instance, non linear wave, damping and current drive, are also considered for special cases of wave frequencies, polarization and propagation. In the concluding section we present the general formulation of the wave damping and current drive in the absence of electron trapping for arbitrary values of the wave frequency. (Author) 13 refs.

  12. Electron - cyclotron resonance heating and current drive

    International Nuclear Information System (INIS)

    Fidone, I.; Castejon, F.

    1992-01-01

    A brief summary of the theory and experiments on electron- cyclotron heating and current drive is presented. The general relativistic formulation of wave propagation and linear absorption is considered in some detail. The O-mode and the X-mode for normal and oblique propagation are investigated and illustrated by several examples. The experimental verification of the theory in T-10 and D- III-D is briefly discussed. Quasilinear evolution of the momentum distribution and related applications as, for instance, non linear wave, damping and current drive, are also considered for special cases of wave frequencies, polarization and propagation. In the concluding section we present the general formulation of the wave damping and current drive in the absence of electron trapping for arbitrary values of the wave frequency. (Author) 13 refs

  13. 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

  14. 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

  15. Comments on ICRH current drive in JET

    International Nuclear Information System (INIS)

    Fried, B.; Hellsten, T.; Moreau, D.

    1989-01-01

    To study current drive via the mode-converted slow wave during ICRH an assessment for which plasma compositions and wave number mode conversion from the magnetosonic wave to the slow wave can dominate is made. A simple slab model is used to investigate the competition between mode conversion and minority cyclotron absorption for a deuterium plasma with H + and 3 He 2+ minority species in JET. A 3 He 2+ minority should be more appropriate for mode conversion current drive than H + because the 3 He 2+ concentration can be chosen near its optimum for the ''Budden absorption'' without bringing the ion hybrid resonance and the cyclotron resonance so close that the minority absorption dominates. 3 He 2+ minority also allows operation at toroidal numbers which are characteristic of present JET antennae. (author)

  16. Lower Hybrid Current Drive Experiments in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Wilson, S. Bernabei, P. Bonoli, A. Hubbard, R. Parker, A. Schmidt, G. Wallace, J. Wright, and the Alcator C-Mod Team

    2007-10-09

    A Lower Hybrid Current Drive (LHCD) system has been installed on the Alcator C-MOD tokamak at MIT. Twelve klystrons at 4.6 GHz feed a 4x22 waveguide array. This system was designed for maximum flexibility in the launched parallel wave-number spectrum. This flexibility allows tailoring of the lower hybrid deposition under a variety of plasma conditions. Power levels up to 900 kW have been injected into the tokomak. The parallel wave number has been varied over a wide range, n|| ~ 1.6–4. Driven currents have been inferred from magnetic measurements by extrapolating to zero loop voltage and by direct comparison to Fisch-Karney theory, yielding an efficiency of n20IR/P ~ 0.3. Modeling using the CQL3D code supports these efficiencies. Sawtooth oscillations vanish, accompanied with peaking of the electron temperature (Te0 rises from 2.8 to 3.8 keV). Central q is inferred to rise above unity from the collapse of the sawtooth inversion radius, indicating off-axis cd as expected. Measurements of non-thermal x-ray and electron cyclotron emission confirm the presence of a significant fast electron population that varies with phase and plasma density. The x-ray emission is observed to be radialy broader than that predicted by simple ray tracing codes. Possible explanations for this broader emission include fast electron diffusion or broader deposition than simple ray tracing predictions (perhaps due to diffractive effects).

  17. 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

  18. Tokamak SST-1: an over-view

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2002-01-01

    Steady State Tokamak SST-1 is in advanced stage of fabrication at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak with superconducting magnets. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas for 1000 s with significant elongation (K) and triangularity (δ). The choice of the parameters is dictated by the physics and technology goals viz. (a) to control and study strongly shaped single and double null divertor plasma, (b) explore advanced tokamak plasma regimes, (c) steady state particle and heat removal from the device, (d) design and operation of large volume superconducting magnets, (e) non-inductive steady state current drive, (f) methods of plasma heating and (g) material technologies

  19. Model for ICRF fast wave current drive in self-consistent MHD equilibria

    International Nuclear Information System (INIS)

    Bonoli, P.T.; Englade, R.C.; Porkolab, M.; Fenstermacher, M.E.

    1993-01-01

    Recently, a model for fast wave current drive in the ion cyclotron radio frequency (ICRF) range was incorporated into the current drive and MHD equilibrium code ACCOME. The ACCOME model combines a free boundary solution of the Grad Shafranov equation with the calculation of driven currents due to neutral beam injection, lower hybrid (LH) waves, bootstrap effects, and ICRF fast waves. The equilibrium and current drive packages iterate between each other to obtain an MHD equilibrium which is consistent with the profiles of driven current density. The ICRF current drive package combines a toroidal full-wave code (FISIC) with a parameterization of the current drive efficiency obtained from an adjoint solution of the Fokker Planck equation. The electron absorption calculation in the full-wave code properly accounts for the combined effects of electron Landau damping (ELD) and transit time magnetic pumping (TTMP), assuming a Maxwellian (or bi-Maxwellian) electron distribution function. Furthermore, the current drive efficiency includes the effects of particle trapping, momentum conserving corrections to the background Fokker Planck collision operator, and toroidally induced variations in the parallel wavenumbers of the injected ICRF waves. This model has been used to carry out detailed studies of advanced physics scenarios in the proposed Tokamak Physics Experiment (TPX). Results are shown, for example, which demonstrate the possibility of achieving stable equilibria at high beta and high bootstrap current fraction in TPX. Model results are also shown for the proposed ITER device

  20. 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].

  1. Heating, current drive and MHD control using ECH in TCV

    International Nuclear Information System (INIS)

    Goodman, T.

    2001-01-01

    The 6 beam 2nd harmonic X-mode (X2), 3MW, ECH/ECCD system of the TCV tokamak allows a fine tailoring of the deposition profiles in the plasma. The sensitivity of the sawtooth period to the deposition location is used to increase the equilibria reconstruction and ray-tracing accuracy. Off-axis ECH, followed by on-axis counter-ECCD produces improved central confinement regimes in which τ Ee exceeds RLW scaling by a factor of 3.5. The PRETOR transport code (incorporating an RLW local transport model but constrained by the experimental density profiles) predicts an extreme sensitivity of τ Ee to the deposition location of the counter-ECCD. This is confirmed by experiments. Sawtooth simulations using PRETOR, including the effects of current drive with inputs from the TORAY ray-tracing code, are in good agreement with experimental results. These results are an initial benchmark for the package of analysis codes, LIUQE/TORAY/PRETOR used during ECH/ECCD experiments on TCV. (author)

  2. Measurement of current drive profile using electron cyclotron wave attenuation near the O-mode cutoff

    International Nuclear Information System (INIS)

    Fidone, I.; Meyer, R.L.; Caron, X.

    1992-01-01

    A method for determining the radial profile of the lower-hybrid current drive in tokamaks using electron cyclotron attenuation of the O mode for frequencies ω near the cutoff frequency is discussed. The basic idea is that, for a given wave frequency, the cutoff plays the role of a spatial filter selecting a variable portion of the noninductive current. It is shown that the incremental attenuation resulting from a small increase of ω displays specific features related to the current density near the cutoff point. Using the relation between the wave damping and the current density, it is possible to determine the radial profile of the current drive from the wave attenuation measurements. A numerical application is also presented for plasma parameters in the reactor regime

  3. 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.

  4. Current density profile control by programming of gas puffing and plasma current waveform in the JIPPT-II tokamak

    International Nuclear Information System (INIS)

    Toi, K.; Itoh, S.; Kadota, K.; Kawahata, K.; Noda, N.

    1979-03-01

    In the resistive shell tokamak, JIPP T-II, the current density profile control is carried out by pre-programming of both gas puffing and plasma current waveform. The major disruptions are completely suppressed by the method and a high density tokamak plasma with low q(a) is obtained with better MHD stability, where the line-average electron density n sub(e) 13 cm -3 and safety factor at plasma surface q(a) >= 2.2. The control criterion that the current density profile is successfully controlled is derived as a function of the ratio of plasma current to the electron density I sub(p)/n sub(e) in the current rising phase, i.e., 20 x 10 -13 -13 kA.cm 3 . (author)

  5. Angular distribution of the bremsstrahlung emission during lower-hybrid current drive on PLT

    International Nuclear Information System (INIS)

    von Goeler, S.; Stevens, J.; Bernabei, S.

    1985-06-01

    The bremsstrahlung emission from the PLT tokamak during lower-hybrid current drive has been measured as a function of angle between the magnetic field and the emission direction. The emission is peaked strongly in the forward direction, indicating a strong anisotropy of the electron-velocity distribution. The data demonstrate the existence of a nearly flat tail of the velocity distribution, which extends out to approximately 500 keV and which is interpreted as the plateau created by Landau damping of the lower-hybrid waves

  6. 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)

  7. Current Drive in a Ponderomotive Potential with Sign Reversal

    International Nuclear Information System (INIS)

    Fisch, N.J.; Rax, J.M.; Dodin, I.Y.

    2003-01-01

    Noninductive current drive can be accomplished through ponderomotive forces with high efficiency when the potential changes sign over the interaction region. The effect can practiced upon both ions and electrons. The current drive efficiencies, in principle, might be higher than those possible with conventional radio-frequency current-drive techniques, since different considerations come into play

  8. 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)

  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. 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. FED-A, an advanced performance FED based on low safety factor and current drive

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Rutherford, P.H.

    1983-08-01

    The FED-A study aims to quantify the potential improvement in cost-effectiveness of the Fusion Engineering Device (FED) by assuming low safety factor q (less than 2 as opposed to about 3) at the plasma edge and noninductive current drive (as opposed to only inductive current drive). The FED-A performance objectives are set to be : (1) ignition assuming International Tokamak Reactor (INTOR) plamsa confinement scaling, but still achieving a fusion power amplification Q greater than or equal to 5 when the confinement is degraded by a factor of 2; (2) neutron wall loading of about 1 MW/m 2 , with 0.5 MW/m 2 as a conservative lower bound; and (3) more clearly power-reactor-like operations, such as steady state

  12. Two-dimensional analysis of beat wave current drive with intense microwave pulses

    International Nuclear Information System (INIS)

    Amin, M.R.; Cairns, R.A.

    1990-01-01

    Current drive in tokamak plasmas by a beat wave is considered in two-dimensional (2-D) geometry. The beat wave is excited by the non-linear interaction of two intense microwave pulses (free electron lasers) in the plasma. The three-wave non-linear interaction equations in steady state are solved numerically. The 2-D toroidal effect and the effect of finite spatial width of the pump microwave pulses are taken into account for the excitation of the beat mode. To illustrate the principle, two types of tokamak are considered: one is small, such as, typically, the Microwave Tokamak Experiment (MTX). and the other one is larger, such as the Joint European Torus (JET). In both cases, it is found that good beat wave coupling exists for a Langmuir beat wave with a phase velocity of around 2.0 to 4.0 times the thermal velocity of the electrons. The fraction of total input power of the right circularly polarized pump waves deposited in the beat mode can be as high as 29% in JET and 32% in MTX. In these cases, there is almost complete pump depletion of the higher frequency pump microwave. It is also found that, for the same input parameters, left circularly polarized pump waves are less efficient than right circularly polarized pump waves for depositing power in the beat mode. (author). 12 refs, 15 figs, 5 tabs

  13. Effects of Current on Behaviors of Saturated Magnetic Island in Tokamak

    International Nuclear Information System (INIS)

    Kanjanaput, W.; Picha, R.; Promping, J.; Poolyarat, N.; Onjun, T.

    2014-01-01

    Plasma current density gradient is known to be one of crucial parameters triggering neoclassical tearing mode in a tokamak plasma. This kind of instability can lead to a formation of magnetic islands, which results in the reduction of plasma pressure and, consequently the degradation of fusion performance. The ISLAND module, developed for determining multiple saturated island width due to different unstable modes, is used in this work. This calculation is based on a quasi-linear theory approach and can include the effect of the bootstrap cur- rent. Both geometry of tokamak and the operation conditions such as magnetic field strength, current and pressure profile are used as initial inputs. The different unstable modes (called m/n, where m and n are the poloidal and toroidal mode number, respectively) are considered. It is found in this work that the mode m/n =2/1 is found to produce the largest saturated island width in the JET and DIIID tokamaks, which agrees with what observed in those to- kamaks. The saturated width of this mode trends to get larger when the gradient of current between the magnetic axis and the mode rational surface increase. The detailed results will be investigated and discussed.

  14. Lower hybrid heating and current drive in Iter operation scenarios and outline system design

    International Nuclear Information System (INIS)

    1994-11-01

    Lower Hybrid Waves (LHW) are considered a valid method of plasma heating and the best demonstrated current drive method. Current drive by LHW possesses the unique feature, as compared to the other methods, to retain a good current drive efficiency in plasma regions of low to medium temperature, or in low-β phases of the discharges. This makes them an essential element to realize the so called 'advanced steady-state Tokamak scenarios' in which a hollow current density profile (deep shear reversal) - established during the ramp-up of the plasma current - offers the prospects of improved confinement and an MHD-stable route to continuous burn. This report contains both modelling and design studies of an LHW system for ITER. It aims primarily at the definition of concepts and parameters for steady-state operation using LHW combined with Fast Waves (FW), or other methods of generating a central seed current for high bootstrap current operation. However simulations addressing the use of LHW for current profile control in the high current pulsed operation scenario are also presented. The outline design of a LHW system which covers the needs for both pulsed and steady-state operation is described in detail. (author). 28 refs., 49 figs

  15. Current drive by asymmetrical heating in a toroidal plasma

    International Nuclear Information System (INIS)

    Gahl, J.M.

    1986-01-01

    This report describes the first experimental observation of current generation by asymmetrical heating of ions. A unidirectional fast Alfven wave launched by a slow-wave antenna inside the Texas Tech Tokamak, asymmetrically heated the ions. Measurements of the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column confirmed the current generation indirectly. Current generation, obtained in a one-species, hydrogen plasma, is a phenomenon which had not been predicted previously. Calculations of the dispersion relation for the fast Alfven wave near the fundamental cyclotron resonance in a one-species, hydrogen plasma, using warm plasma theory, support the experimental results

  16. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    X-ray flux ejected out from the plasma column, with reduced magnetic fluctuations. We have made use of an IRC and has been able to show that the improvement in confinement properties is observed in the presence of sustained negative currents. These negative cur- rents have been observed to be initiated after the fall of ...

  17. 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.

  18. 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

  19. Reconstruction of plasma current profile of tokamaks using combinatorial optimization techniques

    International Nuclear Information System (INIS)

    Kishimoto, Maki; Sakasai, Kaoru; Ara, Katuyuki; Suzuki, Yasuo; Fujita, Takaaki

    1996-01-01

    New methods to reconstruct plasma shape and plasma current distribution from magnetic measurements are proposed. The reconstruction of plasma current profile from magnetic measurements is regarded as an optimum allocation problem of currents into cross section of the vacuum vessel of the tokamak. For solving this optimization problem, the authors use two types of solutions: a genetic algorithm and a combined method of a Hopfield neural network and a genetic algorithm. The effectiveness of these methods is shown by the application of these techniques to JT-60U plasmas

  20. A direct calculation of current drive in toroidal geometry

    International Nuclear Information System (INIS)

    Wright, J.C.; Phillips, C.K.; Bonoli, P.T.

    1998-01-01

    The magnitude and radial profiles of noninductive currents driven by fast magnetosonic waves in tokamaks have been calculated directly from the wave-induced quasilinear flux in a toroidal geometry and a Green's function for the current. An expression for the quasilinear flux has been derived which accounts for coupling between modes in the spectrum of waves launched from the antenna. A Fokker-Planck code for the Green's function and a full wave code for the electric field in the quasilinear flux are used to evaluate the current in a specified toroidal geometry

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-26

    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.

  3. 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

  4. A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints

    International Nuclear Information System (INIS)

    Stewart, W.A.

    1990-05-01

    This thesis describes the development of a design tool for the poloidal field magnet system of a tokamak. Specifically, an existing program for determining the poloidal field coil currents has been modified to: support the general case of asymmetric equilibria and coil sets, determine the coil currents subject to constraints on the maximum values of those currents, and determine the coil currents subject to limits on the forces those coils may carry. The equations representing the current limits and coil force limits are derived and an algorithm based on Newton's method is developed to determine a set of coil currents which satisfies those limits. The resulting program allows the designer to quickly determine whether or not a given coil set is capable of supporting a given equilibrium. 25 refs

  5. 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.

  6. 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.

  7. 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

  8. Theoretical studies of non inductive current drive in compact toroids

    NARCIS (Netherlands)

    Farengo, R; Lifschitz, AF; Caputi, KI; Arista, NR; Clemente, RA

    Three non inductive current drive methods that can be applied to compact toroids axe studied. The use of neutral beams to drive current in field reversed configurations and spheromaks is studied using a Monte Carlo code that includes a complete ionization package and follows the exact particle

  9. Beat wave current drive with intense pulsed free-electron lasers

    International Nuclear Information System (INIS)

    Cohen, B.I.; Cohen, R.H.; Logan, B.G.; McCay Nevins, W.; Smith, G.R.; Kluge, A.V.; Kritz, A.H.

    1988-01-01

    High power free-electron lasers make possible new methods for driving current in toroidal devices with electromagnetic waves. Earlier considerations of beat wave current drive are applied to a hot magnetized plasma and an arbitrary beat wave. Here the beating of two electromagnetic waves resonantly excites a low frequency beat wave that accelerates and heats electrons and leads to a current. The absolute current drive efficiency depends non-linearly on the two pump wave intensities and is constrained by the Manley-Rowe relations. Accessibility at high plasma densities is not a difficulty, but a degree of frequency tunability of the wave sources is required. Particle simulations indicate that there is a good coupling to an electron velocity tail for a Langmuir beat wave with a phase velocity 1.5 to 3 times (T e /m e ) 1/2 , so that all of the high frequency wave source is absorbed and the beat wave damps completely on the electrons. A novel diagnostic, based on an analytical solution for the linearized Fokker-Planck equation describing electron scattering and slowing down, is added to the particle code. This permits the computation of the current drive efficiency, including both the non-linear beat wave coupling and the collisional relaxation of the electron distribution. A realistic scenario for a beat wave current drive experiment in the Livermore Microwave Tokamak Experiment is calculated using the TORAY toroidal ray tracing code, and the scaling to an engineering test reactor plasma is described. (author). 23 refs, 8 figs

  10. 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

  11. 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.

  12. MSE measurements for sawtooth and non-inductive current drive studies in KSTAR

    Science.gov (United States)

    Ko, J.; Park, H.; Bea, Y. S.; Chung, J.; Jeon, Y. M.

    2016-10-01

    Two major topics where the measurement of the magnetic-field-line rotational transform profiles in toroidal plasma systems include the long-standing issue of complete versus incomplete reconnection model of the sawtooth instability and the issue with future reactor-relevant tokamak devices in which non-inductive steady state current sustainment is essential. The motional Stark effect (MSE) diagnostic based on the photoelastic-modulator (PEM) approach is one of the most reliable means to measure the internal magnetic pitch, and thus the rotational transform, or its reciprocal (q), profiles. The MSE system has been commissioned for the Korea Superconducting Tokamak Advanced Research (KSTAR) along with the development of various techniques to minimize systematic offset errors such as Faraday rotation and mis-alignment of the bandpass filters. The diagnostic has revealed the central q is well correlated with the sawtooth oscillation, maintaining its value above unity during the MHD quiescent period and that the response of the q profile to external current drive such as electron cyclotron wave injection not only involves the local change of the pitch angle gradient but also a significant shift of the magnetic topology due to the wave energy transport. Work supported by the Ministry of Science, ICT and Future Planning, Korea.

  13. 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.

  14. Lower hybrid current drive for edge current density modification in DIII-D: Final status report

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Porkolab, M. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center

    1993-08-04

    Application of Lower Hybrid (LH) Current Drive (CD) in the DIII-D tokamak has been studied at LLNL, off and on, for several years. The latest effort began in February 1992 in response to a letter from ASDEX indicating that the 2.45 GHz, 3 MW system there was available to be used on another device. An initial assessment of the possible uses for such a system on DIII-D was made and documented in September 1992. Multiple meetings with GA personnel and members of the LH community nationwide have occurred since that time. The work continued through the submission of the 1995 Field Work Proposals in March 1993 and was then put on hold due to budget limitations. The purpose of this document is to record the status of the work in such a way that it could fairly easily be restarted at a future date. This document will take the form of a collection of Appendices giving both background and the latest results from the FY 1993 work, connected by brief descriptive text. Section 2 will describe the final workshop on LHCD in DIII-D held at GA in February 1993. This was an open meeting with attendees from GA, LLNL, MIT and PPPL. Summary documents from the meeting and subsequent papers describing the results will be included in Appendices. Section 3 will describe the status of work on the use of low frequency (2.45 GHZ) LH power and Parametric Decay Instabilities (PDI) for the special case of high dielectric in the edge regions of the DIII-D plasma. This was one of the critical issues identified at the workshop. Other potential issues for LHCD in the DIII-D scenarios are: (1) damping of the waves on fast ions from neutral beam injection, (2) runaway electrons in the low density edge plasma, (3) the validity of the WKB approximation used in the ray-tracing models in the steep edge density gradients.

  15. 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

  16. 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)

  17. 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

  18. High Harmonic Fast Wave heating and current drive for NSTX

    Science.gov (United States)

    Robinson, J. A.; Majeski, R.; Hosea, J.; Menard, J.; Ono, M.; Phillips, C. K.; Wilson, J. R.; Wright, J.; Batchelor, D. B.; Carter, M. D.; Jaeger, E. F.; Ryan, P.; Swain, D.; Mau, T. K.; Chiu, S. C.; Smithe, D.

    1997-11-01

    Heating and noninductive current drive in NSTX will initially use 6 MW of rf power in the high harmonic fast wave (HHFW) regime. We present numerical modelling of HHFW heating and current drive in NSTX using the PICES, CURRAY, FISIC, and METS95 codes. High electron β during the discharge flattop in NSTX is predicted to result in off-axis power deposition and current drive. However, reductions in the trapped electron fraction (due also to high β effects) are predicted to result in adequate current drive efficiency, with ~ 400 - 500 kA of noninductive current driven. Sufficient per-pass absorption (>10%) to ensure effective electron heating is also expected for the startup plasma. Present plans call for a single twelve strap antenna driven by six FMIT transmitters operating at 30 MHz. The design for the antenna and matching system will also be discussed.

  19. 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)

  20. MHD-mode locking by controlled halo-current in the T-10 tokamak

    International Nuclear Information System (INIS)

    Ivanov, N.V.; Chudnovskiy, A.N.; Gvozdkov, Yu.V.; Kakurin, A.M.; Orlovskiy, I.I.; Pavlov, Yu.D.; Piterskiy, V.V.; Safonova, M.B.; Volkov, V.V.

    2003-01-01

    Experiments on a non-disruptive halo-current influence on the m = 2 mode behaviour at the flat-top stage of a tokamak discharge are presented. The halo-current in the Rail Limiter - Plasma - Vacuum Vessel - External Circuit - Rail Limiter loop was used. An EMF source controlled with a pre-programmed signal or with a feedback m = 2 signal was introduced into the external part of the halo-current circuit. The EMF source generated oscillating halo-currents with up to 500 A amplitude in the frequency range 0-20 kHz. In the case of the pre-programmed control signal the switching on of the EMF source resulted in the shift of the m = 2 mode frequency to the frequency of the halo-current oscillations. In particular, the rotation of the m = 2 mode stopped under a pulse of zero-frequency halo-current. In the tokamak discharges when the mode rotation stopped by itself before the switching on of the oscillating halo-current, the mode rotation was restored at the halo-current frequency. In the case of the halo-current feedback control by the m = 2 mode signal, the effect depended on the choice of the phase shift in the feedback loop. Some increase or decrease of the m = 2 mode amplitude as well as some variations of the mode frequency were observed at different values of the phase shift. The halo-current effect on the m = 2 mode behaviour can be attributed to a coupling between the m/n = 2/1 magnetic islands and the halo-current magnetic field. The experiment was simulated on the assumption that the tearing mode is affected by the halo-current magnetic field helical component with the same space structure. The equation for the disturbed poloidal flux in the presence of the external helical surface current was used for the analysis. In the calculations for the T-10 conditions, the mode behaviour under the effect of the halo-current was similar to the experimental observations. (author)

  1. 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)

  2. Magnetic signature of current carrying edge localized modes filaments on the Joint European Torus tokamak

    DEFF Research Database (Denmark)

    Migliucci, P.; Naulin, Volker

    2010-01-01

    Fast magnetic pickup coils are used in forward modeling to match parameters in a simple edge localized mode (ELM) filament model. This novel method allows us to determine key parameters for the evolution of the ELM filaments, as effective mode number, radial and toroidal velocities, and average...... current from standard magnetic diagnostics. The method is employed on a number of Joint European Torus (JET) [ F. Romanelli, R. Kamendje, and JET-EFDA Contributors, Nucl. Fusion 49, 104006 (2009) ] pulses. The parameter values obtained are compared to ELM filament characterization from JET and other...... tokamaks, obtained by a range of different diagnostics. It is found that the forward modeling produces key parameters such as the number of filaments and their toroidal velocity in agreement with other observations and in addition allows an estimate of the filament current....

  3. Effects of ion cyclotron harmonic damping on current drive in the lower hybrid frequency range

    International Nuclear Information System (INIS)

    Wong, K.L.; Ono, M.

    1983-11-01

    We investigate the ion cyclotron harmonic damping effects on slow and fast waves in the lower hybrid frequency range for tokamak reactor parameters. Inclusion of the higher order terms in the hot plasma dielectric tensor introduces ion cyclotron harmonic damping; these terms also contribute to the real part of the dispersion relation and affect the wave trajectories. However, wave absorption by 15 keV deuterium and tritium ions can be avoided by choosing the slow wave frequency above the lower hybrid frequency and the fast wave frequency below the lower hybrid frequency. But preliminary estimates show that energetic alpha particles tend to absorb both the slow and the fast waves. This absorption may become a serious obstacle for fusion-reactor current drive in the lower hybrid frequency range

  4. Electron cyclotron heating and current drive in toroidal geometry. Technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    Kritz, A.H.

    1993-03-01

    The Principal Investigator has continued to work on problems associated both with the deposition and with the emission of electron cyclotron heating power electron cyclotron heating in toroidal plasmas. Inparticular, the work has focused on the use of electron cyclotron heating to stabilize q = 1 and q = 2 instabilities in tokamaks and on the use of electron cyclotron emission as a plasma diagnostic. The research described in this report has been carried out in collaboration with scientists at Princeton, MIT and Livermore. The Principal Investigator is now employed at Lehigh University, and a small group effort on electron cyclotron heating in plasmas has begun to evolve at Lehigh involving undergraduate and graduate students. Work has also been done in support of the electron cyclotron heating and current drive program at the Center for Research in Plasma Physics in Lausanne, Switzerland.

  5. Current drive by EC waves in the presence of magnetic islands and transport

    International Nuclear Information System (INIS)

    Rosa, P R da S; Ziebell, L F

    2008-01-01

    In this paper we address the problem of current drive by electron cyclotron (EC) waves in the presence of magnetic islands and transport. Our approach makes use of quasilinear theory by numerically solving the Fokker-Planck equation in cylindrical geometry. We take into account the actual geometry of the islands along the calculations as well as the changes in the plasma density profile due to the action of the radial particle transport. The particle transport is supposed to have a magnetic origin. The waves are assumed to be launched and propagated in the equatorial plane of the tokamak, as in the slab geometry. Our results show that the use of equilibrium profiles as usually done in the studies on neoclassical tearing mode control may not be a better choice and point to the need for taking into account the actual island geometry

  6. MHD-mode locking by controlled halo-current in T-10 tokamak

    International Nuclear Information System (INIS)

    Ivanov, N.V.

    2002-01-01

    The experiment on a non-disruptive halo-current influence on the m=2 mode rotation at the steady-state stage of tokamak discharge is presented. The halo current in the (Rail Limiter - Plasma - Vacuum Chamber - External Circuit - Rail Limiter) loop was used. The switching on of an EMF source in the external circuit resulted in locking of the m=2 magnetic islands by the halo current of 400 A amplitude. This effect can be attributed to a coupling between the halo-current magnetic field and the m=2/n=1 mode. A set of magnetic probes was used to measure the halo-current space structure in plasma. The dimensions of the halo-current path in plasma along the magnetic field were much shorter in poloidal and toroidal directions than the corresponding wavelengths for m=2/n=1 mode. The experiment was simulated in the assumption that the tearing mode is affected by halo-current helical component with the same space structure. The equation for disturbed poloidal flux in presence of external helical surface current was used for the analysis. In calculations for T-10 conditions the halo-current affected the mode rotation frequency, like it was observed in the experiment. (author)

  7. 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)

  8. 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

  9. 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

  10. Approximate quantitative relationships for rotating magnetic field current drive

    International Nuclear Information System (INIS)

    Hugrass, W.N.; Ohnishi, M.

    1999-01-01

    A simplified model for the rotating magnetic field (RMF) current drive in an infinitely long cylindrical plasma is used to obtain approximate relationships between the fluid flow velocities, collisionality and degree of nonlinearity. These approximate relationships provide simple quantitative estimates for the basic conditions required for the RMF current drive technique to be applied successfully. In particular, the condition required for the motion of the ion fluid not to be flux-preserving, is evaluated quantitatively for the first time. (author)

  11. Oversampled deadbeat current control strategy for PMSM drives

    OpenAIRE

    Rovere, Luca; Formentini, Andrea; Zanchetta, Pericle

    2016-01-01

    This paper presents a novel deadbeat current control approach for Permanent Magnet Synchronous Motors (PMSMs) drives capable of operating at a controller sampling frequency multiple of the power converter switching frequency. The proposed technique permits to achieve a constant switching frequency and an optimal current ripple along with a high current loop bandwidth and robust behaviour to parameter variation.

  12. 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)

  13. 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)

  14. Designs of the measurement system of the plamsa current and loop voltage on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Yang Qingwei; Feng Beibin; Lu Jie; Zhou Hangyu

    2004-11-01

    The diagnostic systems for plasma current and loop voltage measurement on the HL-2A tokamak have been described in detail. The measurement principles and diagnostic arrangements have been mentioned as well. The more carefully compensation techniques have been used to deal with the interferential magnetic field. (authors)

  15. Current Drive operation in a In-blanket ICRF Ring Array

    Directory of Open Access Journals (Sweden)

    G.Bosia

    2017-01-01

    Full Text Available The currently predicted Current Drive (CDefficiency in a tokamak reactor plasma is χCD ≤ 50 kA/MW for all auxiliary systems (NBCD, ICCD, ECCD. This would imply a relatively low reactor Q = Pelectric /Pauxiliary and an extremely high power density (>10 MW/m2 at the plasma boundary, if injected through the vacuum vessel ports or by in-port launchers.. In a previous paper the concept of an Ion Cyclotron Ring Array (RA of many elements, integrated in the reactor blanket first wall was proposed. An analysis of this RA operating in steady state Fast Wave Current Drive (FW CD with all elements individually controlled in magnitude and phase is presented in this paper. For CD operation the RA is operated at low frequency (f ≈ 20MHz so as to avoid the otherwise overwhelming RF power absorption by ions and α-particles. It is shown that the RA would feature a very directive radiation spectrum, a low power density operation (avoiding the drawbacks associated with high AC and DC potentials and a potentially high flexibility of operation..

  16. Current Drive operation in a In-blanket ICRF Ring Array

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2017-10-01

    The currently predicted Current Drive (CD)efficiency in a tokamak reactor plasma is χCD ≤ 50 kA/MW for all auxiliary systems (NBCD, ICCD, ECCD). This would imply a relatively low reactor Q = Pelectric /Pauxiliary and an extremely high power density (>10 MW/m2) at the plasma boundary, if injected through the vacuum vessel ports or by in-port launchers.. In a previous paper the concept of an Ion Cyclotron Ring Array (RA) of many elements, integrated in the reactor blanket first wall was proposed. An analysis of this RA operating in steady state Fast Wave Current Drive (FW CD) with all elements individually controlled in magnitude and phase is presented in this paper. For CD operation the RA is operated at low frequency (f ≈ 20MHz) so as to avoid the otherwise overwhelming RF power absorption by ions and α-particles. It is shown that the RA would feature a very directive radiation spectrum, a low power density operation (avoiding the drawbacks associated with high AC and DC potentials) and a potentially high flexibility of operation..

  17. Fast wave current drive in neutral beam heated plasmas on DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Forest, C.B.; Pinsker, R.I.

    1997-04-01

    The physics of non-inductive current drive and current profile control using the fast magnetosonic wave has been demonstrated on the DIII-D tokamak. In non-sawtoothing discharges formed by neutral beam injection (NBI), the radial profile of the fast wave current drive (FWCD) was determined by the response of the loop voltage profile to co, counter, and symmetric antenna phasings, and was found to be in good agreement with theoretical models. The application of counter FWCD increased the magnetic shear reversal of the plasma and delayed the onset of sawteeth, compared to co FWCD. The partial absorption of fast waves by energetic beam ions at high harmonics of the ion cyclotron frequency was also evident from a build up of fast particle pressure near the magnetic axis and a correlated increase in the neutron rate. The anomalous fast particle pressure and neutron rate increased with increasing NBI power and peaked when a harmonic of the deuterium cyclotron frequency passed through the center of the plasma. The experimental FWCD efficiency was highest at 2 T where the interaction between the fast waves and the beam ions was weakest; as the magnetic field strength was lowered, the FWCD efficiency decreased to approximately half of the maximum theoretical value

  18. 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.)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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/sup 19/ m/sup -3/) near the upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a/sub 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.

  1. Whistlers, helicons, and lower hybrid waves: The physics of radio frequency wave propagation and absorption for current drive via Landau damping

    International Nuclear Information System (INIS)

    Pinsker, R. I.

    2015-01-01

    This introductory-level tutorial article describes the application of plasma waves in the lower hybrid range of frequencies (LHRF) for current drive in tokamaks. Wave damping mechanisms in a nearly collisionless hot magnetized plasma are briefly described, and the connections between the properties of the damping mechanisms and the optimal choices of wave properties (mode, frequency, wavelength) are explored. The two wave modes available for current drive in the LHRF are described and compared. The terms applied to these waves in different applications of plasma physics are elucidated. The character of the ray paths of these waves in the LHRF is illustrated in slab and toroidal geometries. Applications of these ideas to experiments in the DIII-D tokamak are discussed

  2. RF current drive: common fallacies in simple theory

    International Nuclear Information System (INIS)

    Canobbio, E.; Croci, R.

    1982-01-01

    Two RF current drive models are briefly re-examined: 1) current drive parallel to B-vector 0 as given by the 1 D quasi-linear Fokker-Planck equation. It is shown that the predicted j and P values do not tend to the correct linearized 2 D values when the RF field amplitude is small, although the ratio j/P is essentially correct; 2) current drive perpendicular to B-vector 0 by means of rotating magnetic fields (the Blevin-Thonemann mechanism). The RF field configuration considered so far is found to be incomplete. As a result stable current generation not only occurs under different conditions but also requires much less power than recently estimated. (author)

  3. Stability, energetic particles, waves, and current drive summary

    International Nuclear Information System (INIS)

    Stambaugh, R.D.

    2005-01-01

    This is the summary paper for the subjects of plasma stability, energetic particles, waves, and current drive for the 20th IAEA Fusion Energy Conference, 1-6 November 2004, Vilamoura, Portugal. Material summarized herein was drawn from 65 contributed papers and 21 overview papers. The distribution of contributed papers by subjects is shown. Significant advances were reported on the principal instabilities in magnetically confined plasmas, even looking forward to the burning plasma state. Wave-plasma physics is maturing and novel methods of current drive and noninductive current generation are being developed. (author)

  4. Advanced induction motor drive control with single current sensor

    Directory of Open Access Journals (Sweden)

    Adžić Evgenije M.

    2016-01-01

    Full Text Available This paper proposes induction motor drive control method which uses minimal number of sensors, providing only DC-link current as a feedback signal. Improved DC-link current sampling scheme and modified asymmetrical switching pattern cancels characteristic waveform errors which exist in all three reconstructed motor line-currents. Motor linecurrent harmonic content is reduced to an acceptable level, eliminating torque and speed oscillations which were inherent for conventional single sensor drives. Consequently, use of single current sensor and line-current reconstruction technique is no longer acceptable only for low and medium performance drives, but also for drives where priority is obtaining a highly accurate, stable and fast response. Proposed control algorithm is validated using induction motor drive hardware prototype based on TMS320F2812 digital signal processor. [Projekat Ministarstva nauke Republike Srbije, br. III 042004 and by the Provincial Secretariat for Science and Technological Development of AP Vojvodina under contract No. 114-451-3508/2013-04

  5. 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

  6. Electron-cyclotron resonance heating and current drive

    International Nuclear Information System (INIS)

    Filone, I.

    1992-01-01

    A brief summary of the theory and experiments on electron-cyclotron heating and current drive is presented. the general relativistic formulation of wave propagation and linear absorption is considered in some detail. The O-mode and the X-mode for normal and oblique propagation are investigated and illustrated by several examples. The experimental verification of the theory in T-10 and D-III-D is briefly discussed. Quasilinear evolution of the momentum distribution and related applications as, for instance, non linear wave damping and current drive, are also considered for special cases of wave frequencies, polarization and propagation. In the concluding section we present the general formulation of the wave damping and current drive in the absence of electron trapping for arbitrary values of the wave frequency. (author) 8 fig. 13 ref

  7. High-current cyclotron to drive an electronuclear assembly

    CERN Document Server

    Alenitsky, Yu G

    2002-01-01

    The proposal on creation of a high-current cyclotron complex for driving an electronuclear assembly reported at the 17th Meeting on Accelerators of Charged Particles is discussed. Some changes in the basic design parameters of the accelerator are considered in view of new results obtained in the recent works. It is shown that the cyclotron complex is now the most real and cheapest accelerator for production of proton beams with a power of up to 10 MW. Projects on design of a high-current cyclotron complex for driving an electronuclear subcritical assembly are presented.

  8. Steady-state current drive by lower hybrid waves

    International Nuclear Information System (INIS)

    Belyanskaya, N.V.; Dnestrovskii, Y.N.; Kostomarov, D.P.; Smirnov, A.P.

    1986-01-01

    Steady-state current drive in a plasma by lower-hybrid waves with a square-wave spectrum is analyzed in the linear approximation. A linearized two-dimensional kinetic equation in velocity space for the electron distribution function is reduced to a one-dimensional expansion in Legendre polynomials. A numerical solution is found for the complete equation, and an analytic solution is found for the asymptotic equation. The results can be used to determine the effect of the spectral width of the excited waves on the efficiency of the current drive. The analytic solution agrees well with the numerical solution

  9. Creating poloidal flux in a tokamak plasma with low frequency waves

    International Nuclear Information System (INIS)

    Kirkwood, R.K.; Capewell, D.L.; Bellan, P.M.

    1993-01-01

    Using a fully toroidal, collisionless, low frequency model, we show that low amplitude, circularly polarized waves can, depending on antenna geometry (i) drive the toroidal EMF necessary to sustain a tokamak reactor, or (ii) shift the internal current profile. Measurements on a small tokamak to test (ii) agree with the model predictions. (orig.)

  10. HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

    International Nuclear Information System (INIS)

    Schultz, J.H.; Montgomery, D.B.

    1983-04-01

    HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

  11. Numerical Analysis on the Magneto-Elastic Stability of Current-Carrying Conductors: Aiming at Applications for the Tokamak System

    International Nuclear Information System (INIS)

    Li Weixin; Yuan Zhensheng; Wu Wenjing; Chen Zhenmao

    2013-01-01

    A novel method for calculating the magnetic stiffness matrix was proposed for the numerical analysis of the magneto-elastic stability of complicated current-carrying structures aiming for application in the magneto-elastic behavior of the tokamak system. A code based on the proposed method was developed and applied to the numerical analysis of two typical current-carrying structures. The good consistency of the numerical and analytical results validated the proposed method and the related numerical code. (fusion engineering)

  12. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-01-01

    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)

  13. Disturbance observer based current controller for vector controlled IM drives

    DEFF Research Database (Denmark)

    Teodorescu, Remus; Dal, Mehmet

    2008-01-01

    In order to increase the accuracy of the current control loop, usually, well known parameter compensation and/or cross decoupling techniques are employed for advanced ac drives. In this paper, instead of using these techniques an observer-based current controller is proposed for vector controlled...... induction motor (IM) drives. The control design, based on synchronously rotating d-q frame model of the machine, has a simple structure that combines the proportional portion of a conventional PI control and output of the observer. The observer is predicted to estimate the disturbances caused by parameters...... change in current control loop and, also to remove undesired cross coupling existing between components of the stator current. The observer uses the measured stator currents and estimated PWM voltages, and produces a disturbance signal with a low pass filter. The proposed control scheme reduces cross...

  14. On Ion Cyclotron Current Drive for sawtooth control

    International Nuclear Information System (INIS)

    Eriksson, L.-G.; Johnson, T.; Hellsten, T.; Mayoral, M.-L.; McDonald, D.; Santala, M.; Vries, P. de; Coda, S.; Sauter, O.; Mueck, A.; Buttery, R.J.; Mantsinen, M.J.; Noterdaeme, J.-M.; Westerhof, E.

    2006-01-01

    Experiments using Ion Cyclotron Current Drive (ICCD) to control sawteeth are presented. In particular, discharges demonstrating shortening of fast ion induced long sawteeth reported in [L.-G. Eriksson et al., Physical Review Letters 92, 235004 (2004)] by ICCD have been analysed in detail. Numerical simulations of the ICCD driven currents are shown to be consistent with the experimental observations. They support the hypothesis that an increase of the magnetic shear, due to the driven current, at the surface where the safety factor is unity was the critical factor for the shortening of the sawteeth. In view of the potential utility of ICCD, the mechanisms for the current drive have been further investigated experimentally. This includes the influence of the averaged energy of the resonating ions carrying the current and the spectrum of the launched waves. The results of these experiments are discussed in the light of theoretical considerations. (author)

  15. 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

  16. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Research on Predicting Drive Current of Shipborne Satcom Antenna

    Directory of Open Access Journals (Sweden)

    Kong Jinping

    2015-01-01

    Full Text Available Predicting the effect of antenna wind load on servo system precisely is meaningful to ensure the safety of satcom antenna on operation, which can avoid overload operation. In this paper, the computational fluid dynamics is used to proceed numerical computation on the pressure distribution of the reflector and torque of drive shaft under different wind speed, windward angle and angle of pitch of the antenna. The simulation model is built under MATLAB/Simulink simulation environment, and the drive current of the antenna servo system is analyzed under wind load effect and ship swing. Then, a method of predicting drive current of antenna servo system according to the wind speed, wind direction and attitude of the antenna is concluded. And this method is verified by simulation at last.

  18. Study of the Hamiltonian of the response of a tokamak plasma to the ion cyclotron heating wave: minor heating and generation of current by a 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 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. A quantitative evaluation of the intrinsic stochasticity is derived for ionic trajectories perturbated by the fast wave. The results show the importance of the Hamiltonian chaos in the formation of the deeply anisotropic distribution tails, encountered in minor heating scenarios. Direct interaction of the electrons and the fast wave is analysed. The influence of the various parameters is examined in order to optimize this scenario of fast wave current drive in tokamaks [fr

  19. Lower-hybrid heating and current drive on PLT

    International Nuclear Information System (INIS)

    Hooke, W.; Bernabei, S.; Boyd, D.

    1983-02-01

    Steady currents up to 165 kA for 3.5 seconds and 420 kA for 0.3 seconds have been maintained by 800 MHz lower hybrid waves. For line-averaged densities up to 7 x 10 12 cm - 3 the current is maintained with no input power from the ohmic heating transformer. The waves are launched with an array of six waveguides. Measurements of X rays and electron cyclotron radiation show that the rf power produces and maintains a suprathermal tail of electrons apparently independent of the number of fast electrons in the plasma prior to turning on the rf power. Measurements of current-drive efficiency and the electron tail provide direct evidence for a resonant wave-particle interaction. The radial profile of the rf-sustained current inferred from x-ray measurements is peaked in the center of the plasma and appears to obey the same q-value restraints as the inductively driven ohmic heating current. Current drive is observed to be accompanied always by radiation at frequencies greater than or equal to #betta#/sub ce/ and less than or equal to #betta#/sub pe/. The connection between this radiation and the current-drive mechanism is under study

  20. Langmuir probe study in the nonresonant current drive regime of ...

    Indian Academy of Sciences (India)

    Abstract. Characterization of the current drive regime is done for helicon wave- generated plasma in a torus, at a very high operating frequency. A radiofrequency- compensated Langmuir probe is designed and used for the measurement of plasma parameters along with the electron energy distributions in radial scans of the ...

  1. 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

  2. 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

  3. Evolution of the Tore Supra Lower Hybrid Current Drive System for WEST

    Energy Technology Data Exchange (ETDEWEB)

    Delpech, Léna, E-mail: lena.delpech@cea.fr [CEA, IRFM, F-13108 St Paul-Lez-Durance (France); Achard, Joelle; Armitano, Arthur; Berger-By, Gilles; Ekedahl, Annika; Gargiulo, Laurent; Goniche, Marc; Guilhem, Dominique; Hertout, Patrick; Hillairet, Julien; Magne, Roland; Mollard, Patrick [CEA, IRFM, F-13108 St Paul-Lez-Durance (France); Piluso, P. [CNIM Industrial Systems, 83507 La Seyne-sur-Mer (France); Poli, Serge; Prou, Marc; Saille, Alain; Samaille, Franck [CEA, IRFM, F-13108 St Paul-Lez-Durance (France)

    2015-10-15

    Highlights: • Describe the state of the Lower Hybrid heating system for the WEST project. • Detailed the experiments to assess the coupling in WEST configuration. • Give the modifications required on the launchers to be adapted to WEST configuration. • Detailed the technical modifications with the CNIM company on the launchers. - Abstract: The WEST-project (W-tungsten Environment in Steady-state Tokamak) involves equipping Tore Supra with a full tungsten divertor, capable of withstanding heat load of 10 MW/m{sup 2} in steady-state conditions, in discharges sustained by Lower Hybrid Current Drive (LHCD). The LHCD generator, recently upgraded to deliver 9.2 MW/1000 s, is equipped with sixteen TH2103C klystrons powering two launchers. The WEST transformation involves reducing the plasma volume, thus moving the launchers ∼10 cm closer to the tokamak centre. The toroidal curvature of the launchers no longer fits the plasma curvature due to the strong magnetic field ripple effect, leading to a degradation of the LH wave coupling, especially with the Full Active Multijunction Launcher (FAM). The toroidal curvature radius of the FAM launcher mouth will therefore be reshaped from 1700 mm to 2300 mm. The machining process is described in this article. In order to improve the coupling of the LH wave, the local gas injection has been modified to help to meet the requirement of 7 MW/1000 s of LH power coupled to the plasma in the WEST scenarios. Finally, the curvature radius of the waveguide septa are rounded to minimize the excitation of suprathermal electrons near the plasma edge, which can induce high power loads on the plasma facing components.

  4. Electric machine and current source inverter drive system

    Science.gov (United States)

    Hsu, John S

    2014-06-24

    A drive system includes an electric machine and a current source inverter (CSI). This integration of an electric machine and an inverter uses the machine's field excitation coil for not only flux generation in the machine but also for the CSI inductor. This integration of the two technologies, namely the U machine motor and the CSI, opens a new chapter for the component function integration instead of the traditional integration by simply placing separate machine and inverter components in the same housing. Elimination of the CSI inductor adds to the CSI volumetric reduction of the capacitors and the elimination of PMs for the motor further improve the drive system cost, weight, and volume.

  5. 4 MW upgrade to the DIII-D fast wave current drive system

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Pinsker, R.I.; Cary, W.P.

    1993-10-01

    The DIII-D fast wave current drive (FWCD) system is being upgraded by an additional 4 MW in the 30 to 120 MHz frequency range. This capability adds to the existing 2 MW 30 to 60 MHz system. Two new ABB transmitters of the type that are in use on the ASDEX-Upgrade tokamak in Garching will be used to drive two new water-cooled four-strap antennas to be installed in DIII-D in early 1994. The transmission and tuning system for each antenna will be similar to that now in use for the first 2 MW system on DIII-D, but with some significant improvements. One improvement consists of adding a decoupler element to counter the mutual coupling between the antenna straps which results in large imbalances in the power to a strap for the usual current drive intrastrap phasing of 90{degrees}. Another improvement is to utilize pressurized, ceramic-insulated transmission lines. The intrastrap phasing will again be controlled in pairs, with a pair of straps coupled in a resonant loop configuration, locking their phase difference at either 0 or 180{degrees}, depending upon the length of line installed. These resonant loops will incorporate a phase shifter so that they will be able to be tuned to resonance at several frequencies in the operating band of the transmitter. With the frequency change capability of the ABB generators, the FWCD frequency will thus be selectable on a shot-to-shot basis, from this preselected set of frequencies. The schedule is for experiments to begin with this added 4 MW capability in mid-1994. The details of the system are described.

  6. 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)

  7. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Scharer, J.E.

    1992-01-01

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  8. Current-Drive Efficiency in a Degenerate Plasma

    Energy Technology Data Exchange (ETDEWEB)

    S. Son and N.J. Fisch

    2005-11-01

    a degenerate plasma, the rates of electron processes are much smaller than the classical model would predict, affecting the efficiencies of current generation by external non-inductive means, such as by electromagnetic radiation or intense ion beams. For electron-based mechanisms, the current-drive efficiency is higher than the classical prediction by more than a factor of 6 in a degenerate hydrogen plasma, mainly because the electron-electron collisions do not quickly slow down fast electrons. Moreover, electrons much faster than thermal speeds are more readily excited without exciting thermal electrons. In ion-based mechanisms of current drive, the efficiency is likewise enhanced due to the degeneracy effects, since the electron stopping power on slow ion beams is significantly reduced.

  9. Helicity injection experiment in the SINP tokamak

    International Nuclear Information System (INIS)

    Bhattacharyya, Krishnendu; Ray, Nihar Ranjan

    2000-01-01

    The current drive or sustainment in magnetized toroidal resistive plasmas can be though of as a 'balance' between helicity injection and dissipation. In the present work, the mechanisms of the 'balance' in the fluctuating magnetized resistive plasmas of the SINP tokamak, have been studied experimentally. The result shows that the oscillatory vertical magnetic field and oscillatory plasmas' velocity in a definite phase relationship causes the balancing effect between helicity injection and dissipation and thus sustainment of plasma current for a longer period of time has been observed in the resistive plasmas of the SINP tokamak. (author)

  10. Dynamic modelling of tearing mode stabilization by RF current drive

    International Nuclear Information System (INIS)

    Giruzzi, G.; Zabiego, M.; Gianakon, T.A.; Garbet, X.; Bernabei, S.

    1998-01-01

    The theory of tearing mode stabilization in toroidal plasmas by RF-driven currents that are modulated in phase with the island rotation is investigated. A time scale analysis of the phenomena involved indicates that transient effects, such as finite time response of the driven currents, island rotation during the power pulses, and the inductive response of the plasma, are intrinsically important. A dynamic model of such effects is developed, based on a 3-D Fokker-Planck code coupled to both the electric field diffusion and the island evolution equations. Extensive applications to both Electron Cyclotron and Lower Hybrid current drive in ITER are presented. (author)

  11. Direct calculation of current drive efficiency in FISIC code

    International Nuclear Information System (INIS)

    Wright, J.C.; Phillips, C.K.; Bonoli, P.T.

    1996-01-01

    Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. copyright 1996 American Institute of Physics

  12. Direct calculation of current drive efficiency in FISIC code

    Science.gov (United States)

    Wright, J. C.; Phillips, C. K.; Bonoli, P. T.

    1996-02-01

    Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented.

  13. Direct calculation of current drive efficiency in FISIC code

    Energy Technology Data Exchange (ETDEWEB)

    Wright, J.C.; Phillips, C.K. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543-0451 (United States); Bonoli, P.T. [Plasma Fusion Center, MIT Cambridge, Massachusetts 02139 (United States)

    1996-02-01

    Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires {ital a} {ital priori} knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. {copyright} {ital 1996 American Institute of Physics.}

  14. RF heating and current drive in Tore Supra

    International Nuclear Information System (INIS)

    Litaudon, X.

    1995-01-01

    Recent RF heating and current drive experiments in the Lower Hybrid (LH) and Ion Cyclotron (IC) frequency ranges are reported. In the 4T improved confinement LHEP regime, steady-state LHCD operation has been realized with a new ''constant-flux'' scenario. A new, reversed shear, 2T improved confinement plasma regime has also been investigated when the core plasma is inaccessible to the LH waves. Stable, LH driven 0.4 MA discharges were obtained with H rlw = 2 at βp = 0.8, q o above 2 and with a reduced electron thermal diffusivity in the central reversed shear region. Efficient direct coupling of the fast magnetosonic wave to the electrons for heating and current drive is observed during 48 MHz/2T operation. Fast wave electron heating has produced improved confinement with H rlw = 2 at βp = 1.6, and a bootstrap current fraction up to 45%. Fast wave current drive has been observed at the level of 80 kA in a 0.4 MA discharge. (authors). 28 refs., 7 figs

  15. Hard X-ray studies on the Castor tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.

    1990-04-01

    The electron runaway processes in tokamaks are discussed with regard to hard X radiation measurements. The origin and confinement of runaway electrons, their bremsstrahlung spectra and the influence of lower hybrid current drive on the distribution of high-energy electrons are analyzed for the case of the Castor tokamak. The hard X-ray spectrometer designed for the Castor tokamak is also described and preliminary qualitative results of hard X-ray measurements are presented. The first series of integral measurements made it possible to map the azimuthal dependence of the hard X radiation

  16. RF heating and current drive on NSTX with high harmonic fast waves

    International Nuclear Information System (INIS)

    Ryan, P.M.

    2002-01-01

    NSTX is a small aspect ratio tokamak with a large dielectric constant (50-100); under these conditions high harmonic fast waves (HHFW) will readily damp on electrons via Landau damping and TTMP. The HHFW system is a 30 MHz, 12-element array capable of launching both symmetric and directional wave spectra for plasma heating and non-inductive current drive. It has delivered up to 6 MW for short pulses and has routinely operated at ∼3-4 MW for 100-200 ms pulses. Results include strong, centrally-peaked electron heating in both D and He plasmas, for both high and low phase velocity spectra. H-modes were obtained with application of HHFW power alone, with stored energy doubling after the L-H transition. Beta poloidal as large as unity has been obtained with large fractions (0.4) of bootstrap current. A fast ion tail with energies extending up to 140 keV has been observed when HHFW interacts with 80 keV neutral beams; neutron rate and lost ion measurements, as well as modeling, indicate significant power absorption by the fast ions. Radial power deposition profiles are being calculated with ray tracing and kinetic full-wave codes and benchmarked against measurements. (author)

  17. Survey of heating and current drive for K-DEMO

    Science.gov (United States)

    Mikkelsen, D. R.; Kessel, C. E.; Poli, F. M.; Bertelli, N.; Kim, K.

    2018-03-01

    We present calculations of heating and current drive by neutral injection and by electromagnetic waves in the ion cyclotron, helicon, lower hybrid, and electron cyclotron frequency ranges for the steady state burn conditions in a K-DEMO configuration with I_p=12.3 MA, a  =  2.1 m, R_o=6.8 m, B_o=7.4 T, \

  18. 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

  19. Towards fully non-inductive current drive operation in JET

    International Nuclear Information System (INIS)

    Litaudon, X.; Crisanti, F.; Alper, B.

    2002-01-01

    Quasi steady operation has been achieved at JET in the high confinement regime with Internal Transport Barriers, ITBs. The ITBs' performances are maintained up to 11 s. This duration, much larger than the energy confinement time, is already approaching a current resistive time. The high performance phase is limited only by plant constraints. The radial profiles of the thermal electron and ion pressures have steep gradients typically at mid-plasma radius. A large fraction of non-inductive current (above 80%) is sustained throughout the high performance phase with a poloidal beta exceeding unity. The safety factor profile plays an important role in sustaining the ITB characteristics. In this regime where the self-generated bootstrap current (up to LOMA) represents 50% of the total current, the resistive evolution of the non-monotonic q-profile is slowed down by using off-axis lower hybrid current drive. (authors)

  20. Towards fully non-inductive current drive operation in JET

    Energy Technology Data Exchange (ETDEWEB)

    Litaudon, X. [Association Euratom-CEA Cadarache, Dept. de Recherches sur la Fusion Controlee, 13 - Saint-Paul-lez-Durance (France); Crisanti, F. [Association Euratom-ENEA sulla Fusione, Centro Ricerche Frascati (Italy); Alper, B. [Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon (United Kingdom)] [and others

    2002-01-01

    Quasi steady operation has been achieved at JET in the high confinement regime with Internal Transport Barriers, ITBs. The ITBs' performances are maintained up to 11 s. This duration, much larger than the energy confinement time, is already approaching a current resistive time. The high performance phase is limited only by plant constraints. The radial profiles of the thermal electron and ion pressures have steep gradients typically at mid-plasma radius. A large fraction of non-inductive current (above 80%) is sustained throughout the high performance phase with a poloidal beta exceeding unity. The safety factor profile plays an important role in sustaining the ITB characteristics. In this regime where the self-generated bootstrap current (up to LOMA) represents 50% of the total current, the resistive evolution of the non-monotonic q-profile is slowed down by using off-axis lower hybrid current drive. (authors)

  1. Direct Calculations of Current Drive with a Full Wave Code

    Science.gov (United States)

    Wright, John C.; Phillips, Cynthia K.

    1997-11-01

    We have developed a current drive package that evaluates the current driven by fast magnetosonic waves in arbitrary flux geometry. An expression for the quasilinear flux has been derived which accounts for coupling between modes in the spectrum of waves launched from the antenna. The field amplitudes are calculated in the full wave code, FISIC, and the current response function, \\chi, also known as the Spitzer function, is determined with Charles Karney's Fokker-Planck code, adj.f. Both codes have been modified to incorporate the same numerical equilibria. To model the effects of a trapped particle population, the bounce averaged equations for current and power are used, and the bounce averaged flux is calculated. The computer model is benchmarked against the homogenous equations for a high aspect ratio case in which the expected agreement is confirmed. Results from cases for TFTR, NSTX and CDX-U are contrasted with the predictions of the Ehst-Karney parameterization of current drive for circular equilibria. For theoretical background, please see the authors' archive of papers. (http://w3.pppl.gov/ ~jwright/Publications)

  2. 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.

  3. Progress in the ITER electron cyclotron heating and current drive system design

    International Nuclear Information System (INIS)

    Omori, Toshimichi; Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio; Darbos, Caroline; Denisov, Grigory; Farina, Daniela; Gagliardi, Mario; Gandini, Franco; Gassmann, Thibault; Goodman, Timothy; Hanson, Gregory; Henderson, Mark A.; Kajiwara, Ken; McElhaney, Karen; Nousiainen, Risto; Oda, Yasuhisa; Oustinov, Alexander

    2015-01-01

    Highlights: • EC system is designed with an ability to upgrade in power to 28 MW then 40 MW. • The TL is capable of 3 buildings movements; ±15 mm displacements at the worst. • Improved power deposition access injecting 20 MW across nearly the entire plasma. • Ensured nuclear safety by appropriate definition of confinement boundaries. • Proposed I&C architecture for the overall EC plant was successfully reviewed. - Abstract: An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion.

  4. Electron cyclotron current drive in the Wendelstein 7-AS stellarator

    International Nuclear Information System (INIS)

    Maassberg, H; Rome, M; Erckmann, V; Geiger, J; Laqua, H P; Marushchenko, N B

    2005-01-01

    High power electron cyclotron current drive (ECCD) experiments in the W7-AS stellarator are analysed. In these net-current-free discharges, the ECCD and the bootstrap current are feedback controlled by an inductive current. Based on the measured density and temperature profiles, the neoclassical predictions of the bootstrap (with the ambipolar radial electric field taken into account) and the inductive current densities as well as the ECCD from the linear adjoint approach with trapped particles included are calculated. For stationary conditions, the current balance is checked. Launch-angle scans at fixed density as well as density scans at fixed launch-angle are described. Low-frequency MHD mode activity is obtained for strong co-ECCD, and for counter-ECCD a ' ι-bar approx.= 0 feature' with complete loss of the central confinement is found. The linear ECCD prediction is in reasonable agreement with the current balance except for low-density discharges with highly peaked on-axis deposition, where the ECCD predicted from linear theory exceeds by a factor of about 2 the one from the current balance. Since the bootstrap current is well balanced by the inductive current without ECCD, the linear ECCD overestimate is compared with nonlinear Fokker-Planck (FP) simulations, where two different power loss models are used to reach steady state. These volume-averaged FP simulations cannot describe the ECCD degradation at the low densities

  5. Steady-state experiments on high performance, current profile control and long sustainment of LHCD plasmas on the superconducting tokamak TRIAM-1M

    International Nuclear Information System (INIS)

    Zushi, H.; Itoh, S.; Nakamura, K.; Sakamoto, M.; Hanada, K.; Jotaki, E.; Hasegawa, M.; Kawasaki, S.; Nakashima, H.; Pan, Y.D.

    2001-01-01

    TRIAM-1M (R 0 =0.8m, axb=0.12mx0.18m and B=8T) has the main mission to study the route toward a high field compact steady state fusion reactor. We have advanced steady state operation (SSO) programme in tokamaks, studied a heating mechanism for the high ion temperature (HIT) mode with an internal transport barrier, obtained an enhanced current drive (ECD) mode in an extended (higher power and higher density) operation regime, performed current density profile control experiments using multi-current drive systems and investigated effects of wall recycling, wall pumping and wall saturation on particle control. In HIT mode a hysteresis relation between T i and n e is found to be ascribed to different timescales for T i and n e to change. Excitation of plasma waves corresponding to ion heating is studied by both measurements of electromagnetic and electrostatic waves and their analysis. Achieved plasma parameters in ECD are as follows; n e is 4.3x10 19 m -3 , I LHCD is ∼70kA, T e and T i are 0.8 keV and 0.5 keV, respectively, and the stored energy is 1.9 kJ. The energy confinement time τ E of 8-10 ms, H ITER89-P ∼ 1.4, is achieved and the current drive efficiency η CD =n-bar e I CD R 0 /P LH reaches 1x10 19 Am -2 /W at B=7 T under the fully non-inductive condition. Power threshold and hysteresis nature are studied. Bi-directional current drive and superposed current drive experiments have been carried out. In the former steady current reduction and peaking of j(r) are observed, but it is noticed that self-organization of j(r) occurs above a certain power ratio. In the latter broadening of j(r) can be obtained by increasing superposed RF power, however, self-organization of j(r) also occurs again at a certain power. Temporal behaviour of the recycling coefficient with two different time constants (∼3 s and ∼30 s) is analysed. The wall pumping rates are evaluated to be ∼1.5x10 16 atoms/s/m 2 for low n e and ∼4x10 17 atoms/s/m 2 for high n e , respectively

  6. Experimental study of coupling between eddy currents and deflections in cantilevered beams as models of tokamak limiters

    International Nuclear Information System (INIS)

    Turner, L.R.; Hua, T.Q.

    1987-01-01

    The coupling between eddy current and motion in a cantilevered beam is examined. The beam, which provides a simple model for the limiter blades of a tokamak fusion reactor, was subjected to simultaneous orthogonal time-varying and constant magnetic fields. The dynamic deformation of the beam includes two different modes: a bending mode and a torsional mode. Interaction of current with each mode and with the combined modes of vibration is described. Experimental verification for the case without torsional motion was performed with the FELIX facility at ANL. The peak deflection and stresses are much less than those predicted without consideration given to the coupling. (Auth.)

  7. Experimental study of coupling between eddy currents and deflections in cantilevered beams as models of tokamak limiters

    International Nuclear Information System (INIS)

    Turner, L.R.; Hua, T.Q.

    1986-01-01

    The coupling between eddy current and motion in a cantilevered beam is examined. The beam, which provides a simple model for the limiter blades of a tokamak fusion reactor, was subjected to simultaneous orthogonal time-varying and constant magnetic fields. The dynamic deformation of the beam includes two different modes: a bending mode and a torsional mode. Interaction of current with each mode and with the combined modes of vibration in described. Experimental verification of the case without torsional motion was performed with the FELIX facility at ANL. The peak deflection and stresses are much less than those predicted without consideration given to the coupling

  8. 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

  9. Modeling of finite aspect ratio effects on current drive

    International Nuclear Information System (INIS)

    Wright, J.C.; Phillips, C.K.

    1996-01-01

    Most 2D RF modeling codes use a parameterization of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by a full wave code. This eliminates the need to use the approximation inherent in the parameterization. Current profiles are then calculated using the adjoint formulation. This approach has been implemented in the FISIC code. The accuracy of the parameterization of the current drive efficiency, η, is judged by a comparison with a direct calculation: where χ is the adjoint function, ε is the kinetic energy, and rvec Γ is the quasilinear flux. It is shown that for large aspect ratio devices (ε → 0), the parameterization is nearly identical to the direct calculation. As the aspect ratio approaches unity, visible differences between the two calculations appear

  10. Modeling of finite aspect ratio effects on current drive

    Energy Technology Data Exchange (ETDEWEB)

    Wright, J.C.; Phillips, C.K. [Princeton Plasma Physics Lab., NJ (United States)

    1996-12-31

    Most 2D RF modeling codes use a parameterization of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by a full wave code. This eliminates the need to use the approximation inherent in the parameterization. Current profiles are then calculated using the adjoint formulation. This approach has been implemented in the FISIC code. The accuracy of the parameterization of the current drive efficiency, {eta}, is judged by a comparison with a direct calculation: where {chi} is the adjoint function, {epsilon} is the kinetic energy, and {rvec {Gamma}} is the quasilinear flux. It is shown that for large aspect ratio devices ({epsilon} {r_arrow} 0), the parameterization is nearly identical to the direct calculation. As the aspect ratio approaches unity, visible differences between the two calculations appear.

  11. 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

  12. Linear servomotor probe drive system with real-time self-adaptive position control for the Alcator C-Mod tokamak.

    Science.gov (United States)

    Brunner, D; Kuang, A Q; LaBombard, B; Burke, W

    2017-07-01

    A new servomotor drive system has been developed for the horizontal reciprocating probe on the Alcator C-Mod tokamak. Real-time measurements of plasma temperature and density-through use of a mirror Langmuir probe bias system-combined with a commercial linear servomotor and controller enable self-adaptive position control. Probe surface temperature and its rate of change are computed in real time and used to control probe insertion depth. It is found that a universal trigger threshold can be defined in terms of these two parameters; if the probe is triggered to retract when crossing the trigger threshold, it will reach the same ultimate surface temperature, independent of velocity, acceleration, or scrape-off layer heat flux scale length. In addition to controlling the probe motion, the controller is used to monitor and control all aspects of the integrated probe drive system.

  13. 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

  14. 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)

  15. On the merits of heating and current drive for tearing mode stabilization

    NARCIS (Netherlands)

    De Lazzari, D.; Westerhof, E.

    2009-01-01

    Neoclassical tearing modes (NTMs) are magnetohydrodynamic modes that can limit the performance of high beta discharges in a tokamak, leading eventually to a plasma disruption. A NTM is sustained by the perturbation of the 'bootstrap' current, which is a consequence of the pressure

  16. On the minimum circulating power of steady state tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, K.; Itoh, S.; Fukuyama, A.; Yagi, M.

    1995-07-01

    Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author).

  17. 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.

  18. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options

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

    International Nuclear Information System (INIS)

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

    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)

  20. 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.)

  1. 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

  2. Runaway electron studies with hard x-ray and microwave diagnostics in the FT-2 lower hybrid current drive discharges

    Science.gov (United States)

    Shevelev, A. E.; Khilkevitch, E. M.; Lashkul, S. I.; Rozhdestvensky, V. V.; Pandya, S. P.; Plyusnin, V. V.; Altukhov, A. B.; Kouprienko, D. V.; Chugunov, I. N.; Doinikov, D. N.; Esipov, L. A.; Gin, D. B.; Iliasova, M. V.; Naidenov, V. O.; Polunovsky, I. A.; Sidorov, A. V.; Kiptily, V. G.

    2018-01-01

    Studies of the super-thermal and runaway electron behavior in ohmic and lower hybrid current drive FT-2 tokamak plasmas have been carried out using information obtained from measurements of hard x-ray spectra and non-thermal microwave radiation intensity at the frequency of 10 GHz and in the range of (53 ÷ 78) GHz. A gamma-ray spectrometer based on a scintillation detector with a LaBr3(Ce) crystal was used, which provides measurements at counting rates up to 107 s-1. Reconstruction of the energy distribution of RE interacting with the poloidal limiter of the tokamak chamber was made with application of the DeGaSum code. Super-thermal electrons accelerated up to 2 MeV by the LH waves at the high-frequency pumping of the plasma with low density ≤ft ~ 2  ×  1013 cm-3 and then up to 7 MeV by vortex electric field have been found. Experimental analysis of the runaway electron beam generation and evolution of their energy distribution in the FT-2 plasmas is presented in the article and compared with the numerical calculation of the maximum energy gained by runaway electrons for given plasma parameters. In addition, possible mechanisms for limiting the maximum energy gained by the runaway electrons are also calculated and described for a FT-2 plasma discharge.

  3. 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)

  4. Lower hybrid current drive in Tore Supra and JET

    International Nuclear Information System (INIS)

    Moreau, D.; Gormezano, C.

    1991-01-01

    Recent Lower Hybrid Current Drive (LHCD) experiments in TORE SUPRA and JET are reported. Large multijunction launchers have allowed the coupling of 5MW to the plasma for several seconds with a maximum of 3.8 kW/cm 2 . Measurements of the scattering matrices of the antennae show good agreement with theory. The current drive efficiency in TORE SUPRA is about 0.2 x 10 20 Am -2 /W with LH power alone and reaches 0.4 x 10 20 Am -2 /W in JET thanks to a high volume-averaged electron temperature (1.9 keV) and also to a synergy between Lower Hybrid and Fast Magnetosonic Waves. At n e = 1.5 x 10 19 m -3 in TORE SUPRA, sawteeth are suppressed and m = 1MHD oscillations the frequency of which clearly depends on the amount of LH power are observed on soft X-rays, and also on non-thermal ECE. In Jet ICRH produced sawtooth free periods are extended by the application of LHCD and current profile broadening has been clearly observed consistent with off-axis fast electron populations. LH power modulation experiments performed in TORE SUPRA at n e = 4 x 10 19 m -3 show a delayed central electron heating despite the off-axis creation of suprathermal electrons, thus ruling out the possibility of a direct heating through central wave absorption. A possible explanation in terms of anomalous fast electron transport and classical slowing down would yield a diffusion coefficient of the order of 10 m 2 /s for the fast electrons. Finally, successful pellet fuelling of a partially LH driven plasma was obtained in TORE SUPRA, 28 successive pellets allowing the density to rise to n e = 4 x 10 19 m -3 . This could be achieved by switching the LH power off for 90 ms before each pellet injection, i.e. without modifying significantly the current density profile

  5. Plasma current start-up experiments without a central solenoid in the iron core STOR-M tokamak

    Science.gov (United States)

    Mitarai, O.; Tomney, G.; Rohollohi, A.; Lewis, E.; McColl, D.; Xiao, C.; Hirose, A.

    2015-06-01

    Reproducible plasma current start-up without a central solenoid (CS) has been demonstrated using the outer ohmic heating (OH) coils in the iron core STOR-M tokamak (Mitarai et al 2014 Fusion Eng. Des. 89 2467-71). Although the outer OH coil current saturates the iron core eventually, it has been demonstrated that the plasma current can be maintained during the iron core saturation phase. In this work, further studies have been conducted to investigate the effects of the turn number of the outer OH coils (N = 4 or N = 6) in the CS-less discharges and to evaluate the plasma stability with respect to the n-decay index of the vertical magnetic field. For the loose coupling of the iron core with N = 4 turns, the plasma current can be sustained after the additional third capacitor bank is applied near the iron core saturation phase, showing the slow transition from the unsaturated to the partially saturated phase. For the case of stronger coupling of N = 6 turns, the plasma current is increased at the same fast bank voltage, but the main discharge is shortened from 35 to 20 ms. As the magnetizing current is smaller due to stronger coupling between the OH coils and the plasma current, the transition from the unsaturated to the saturated phase is slightly difficult at present. The present experimental results suggest a feasible operation scenario in a future spherical tokamak (ST) at least using loose iron core coupling for smoother transition from the unsaturated to the saturated iron core phase. Thus, a reliable plasma current start-up by the outer OH coils and the current ramp-up to a steady state by additional heating power and vertical field coils could be considered as an operation scenario for future ST reactors with an iron core transformer.

  6. Observation of poloidal current flow to the vacuum vessel wall during vertical instabilities in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Strait, E.J.; Lao, L.L.; Luxon, J.L.; Reis, E.E.

    1991-01-01

    An attached poloidal current, which flows in a circuit lying partly in the vacuum vessel wall and partly in the scrape-off layer of the plasma, is observed during vertical instabilities in the DIII-D tokamak. A direct measurement of the current, using Rogowski loops on several protective tiles at locations where the plasma contacts the wall, is in good agreement with the value determined from MHD equilibrium reconstructions using measured values of magnetic field and flux. This attached current, which can reach transient peaks of several hundred kilo-amperes, interacts with the toroidal magnetic field to create a large vertical force on the vacuum vessel. The predicted motion of the vessel resulting from the measured currents agrees well with the observed displacement of the vacuum vessel. (author). 14 refs, 5 figs

  7. 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.

  8. Physics design of the HL-1M tokamak

    International Nuclear Information System (INIS)

    Gao Qingdi; Shi Bingren; Liu Yukui; Zhang Jinhua; Xue Siwen; Li Fangzhu

    1999-08-01

    Presented is the physics design of the HL-1M tokamak, which is a machine upgraded from the HL-1 tokamak. Based upon the intensive investigations on the controlled nuclear fusion research in the world, the direction for modifying the HL-1 tokamak was determined, i.e. reconstructing the vacuum chamber without the thick copper shell which is used as an outer vacuum vessel in HL-1, reforming the poloidal magnetic field system and upgrading the power supply so as to be suitable for performing experimental study on high power auxiliary heating and non-inductive current drive. The main physics objectives of HL-1M is to carry out investigations on MWs power auxiliary heating and current drive with lower hybrid wave. Besides this, the other physics objectives are as follows: to perform further experimental study on the ohmic heating plasma with higher parameters so that a database for extrapolating to a larger tokamak device could be obtained, and to accumulate experiences for the construction of next tokamak device, HL-2. By using the extrapolation of the HL 1 experiment results, the tokamak scaling law and numerical computation, the physics parameters of ohmic heating and auxiliary heating plasmas are designed in some details

  9. 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

  10. 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

  11. 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.

  12. 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

  13. Trapped electron effects on ICRF Current Drive Predictions in TFTR

    Science.gov (United States)

    Wright, John C.; Phillips, Cynthia K.; Bonoli, Paul T.

    1996-11-01

    Most 2D RF modeling codes use a parameterization^1 of current drive efficiencies to calculate fast wave driven currents. Because this parameterization is derived from a ray--tracing model, there are difficulties in applying it to a spectrum of waves. In addition, one cannot account for multiple resonances and coherency effects between the electrons and the waves. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient in an inhomogenous geometry coupled with a full wave code for the field polarizations. Current profiles are then calculated using the adjoint formulation^2, with the magnetic equilibrium specified consistently in both the adjoint routine and the full wave code. This approach has been implemented in the FISIC code^3. Results are benchmarked by comparing a power deposition calculation from conductivity to one from the quasilinear expression. It is shown that the two expressions agree. We quantify differences seen based upon aspect ratio and elongation. The largest discrepancies are seen in the regime of small aspect ratio, and little loss in accuracy for moderate aspect ratios ~>3. This work supported by DoE contract No. DE--AC02--76--CH03073. ^1 D. A. Ehst and C. F. F. Karney, Nucl. Fusion 31, 1933 (1991). ^2 C. F. F. Karney, Computer Physics Reports 4, 183 (1986). ^3 M. Brambilla and T. Krücken, Nucl. Fusion 28, 1813 (1988).

  14. Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

    Science.gov (United States)

    Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

    2017-10-01

    The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

  15. Lower hybrid current drive in Tore Supra and jet

    International Nuclear Information System (INIS)

    Moreau, D.; Gormezano, C.

    1991-07-01

    Recent Lower Hybrid Current Drive (LHCD) experiments in TORE SUPRA and JET are reported. Large multijunction launchers have allowed the coupling of 5 MW to the plasma for several seconds with a maximum of 3.8 kW/cm 2 . Measurements of the scattering matrices of the antennae show good agreement with theory. The current drive efficiency in TORE SUPRA is about 0.2 x 10 20 Am -2 /W with LH power alone and reaches 0.4 x 10 20 Am -2 /W in JET thanks to a high volume-averaged electron temperature (1.9 keV) and also to a synergy between Lower Hybrid and Fast Magnetosonic Waves. At average n e = 1.5 x 10 19 m -3 in TORE SUPRA, sawteeth are suppressed and m = 1 MHD oscillations the frequency of which clearly depends on the amount of LH power are observed on soft X-rays, and also on non-thermal ECE. In JET ICRH produced sawtooth-free periods are extended by the application of LHCD (2.9 s. with 4 MW ICRH) and current profile broadening has been clearly observed consistent with off-axis fast electron populations. LH power modulation experiments performed in TORE SUPRA at average n e = 4 x 10 19 m -3 show a delayed central electron heating despite the off-axis creation of suprathermal electrons, thus ruling out the possibility of a direct heating through central wave absorption. A possible explanation in terms of anomalous fast electron transport and classical slowing down would yield a diffusion coefficient of the order of 10 m 2 /s for the fast electrons. Other interpretations such as an anomalous heat pinch or a central confinement enhancement cannot be excluded. Finally, successful pellet fuelling of a partially LH driven plasma was obtained in TORE SUPRA, 28 successive pellets allowing the density to rise to average n e = 4 x 10 19 m -3 . This could be achieved by switching the LH power off for 90 ms before each pellet injection, i.e. without modifying significantly the current density profile

  16. 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

  17. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

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

  18. Conceptual design of a radio-frequency driven compact tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Ludwig, G.O.; Montes, A.; Ueda, M.; Goes, L.C.S.

    1987-09-01

    Preliminary results of the design of a small compact tokamak are presented. The design incorporates advanced concepts as start-up and current drive by electron-cyclotron and lower-hybrid waves; plasma heating by intense ion beams; and achievement of high-β by decreasing the aspect ratio. (author) [pt

  19. 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.

  20. 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.

  1. 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

  2. The microwave Tokamak experiment (MTX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-01-01

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement

  3. 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

  4. Current Behaviours and Attitudes Towards Texting While Driving in Australia

    DEFF Research Database (Denmark)

    Adamsen, Jannie Mia; Beasley, Keiran

    This paper aims to understand the behaviour of texting and driving among the broader driving public in Australia and uncover whether attitudes are congruent with behaviours. Recent studies have generally been focussing on the behaviours of 18-24 year olds suggesting that the practice is mainly...... confined to people in this age bracket. Findings from an anonymous online survey show that the practice of texting and driving is widespread in Australia and not just confined to the younger demographic. Additionally, evidence suggests smart phone users are more likely to engage in texting while driving...

  5. Fokker-Planck modeling of current penetration during electron cyclotron current drive

    International Nuclear Information System (INIS)

    Merkulov, A.; Westerhof, E.; Schueller, F. C.

    2007-01-01

    The current penetration during electron cyclotron current drive (ECCD) on the resistive time scale is studied with a Fokker-Planck simulation, which includes a model for the magnetic diffusion that determines the parallel electric field evolution. The existence of the synergy between the inductive electric field and EC driven current complicates the process of the current penetration and invalidates the standard method of calculation in which Ohm's law is simply approximated by j-j cd =σE. Here it is proposed to obtain at every time step a self-consistent approximation to the plasma resistivity from the Fokker-Planck code, which is then used in a concurrent calculation of the magnetic diffusion equation in order to obtain the inductive electric field at the next time step. A series of Fokker-Planck calculations including a self-consistent evolution of the inductive electric field has been performed. Both the ECCD power and the electron density have been varied, thus varying the well known nonlinearity parameter for ECCD P rf [MW/m -3 ]/n e 2 [10 19 m -3 ] [R. W. Harvey et al., Phys. Rev. Lett 62, 426 (1989)]. This parameter turns out also to be a good predictor of the synergetic effects. The results are then compared with the standard method of calculations of the current penetration using a transport code. At low values of the Harvey parameter, the standard method is in quantitative agreement with Fokker-Planck calculations. However, at high values of the Harvey parameter, synergy between ECCD and E parallel is found. In the case of cocurrent drive, this synergy leads to the generation of large amounts of nonthermal electrons and a concomitant increase of the electrical conductivity and current penetration time. In the case of countercurrent drive, the ECCD efficiency is suppressed by the synergy with E parallel while only a small amount of nonthermal electrons is produced

  6. 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

  7. 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

  8. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    International Nuclear Information System (INIS)

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

  9. 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

  10. A steady state tokamak operation by use of magnetic monopoles

    International Nuclear Information System (INIS)

    Narihara, K.

    1991-12-01

    A steady state tokamak operation based on a magnetic monopole circuit is considered. Circulation of a chain of iron cubes which trap magnetic monopoles generates the needed loop voltage. The monopole circuit is enclosed by a series of solenoid coils in which magnetic field is feedback controlled so that the force on the circuit balance against the mechanical friction. The driving power is supplied through the current sources of poloidal, ohmic and solenoid coils. The current drive efficiency is same as that of the ohmic current drive. (author)

  11. The current status of the psychoanalytic theory of instinctual drives. I: Drive concept, classification, and development.

    Science.gov (United States)

    Compton, A

    1983-07-01

    The evolution of Freud's theory of instinctual drives, with the accompanying models of a mental apparatus, is remarkable for its tenacious adherence to addressing the fundamental problems of human psychology, here phrased as the problems of body-mind-environment relationships. The concept of instinctual drives continues to be one of the most pervasive concepts of psychoanalysis, weathering considerable attack over the last several decades, although losing some clarity in the process. I have cited and discussed as basic issues of the concept of instinctual drives: the relationship of observational data and theoretical constructs in psychology; whether our construct of drives is or should be or can be purely psychological; the problem of conceptualizing the ontogenetic origin of mind; the issues of the "force-meaning conjunction" and the problem of psychic energy in psychoanalytic constructs; and the relation of our concept of instinctual drives to the concept of instincts in general. It seems that progress with these fundamental issues might be made by utilizing models that are more homologous with present knowledge in related fields than is Freud's reflex arc model of the nervous system, in order to build a better drive construct within the framework of psychoanalysis. The classification of instinctual drives remains a problem. Clinically, aggression seems to be a factor in conflict, very much like sexuality. Despite widespread acceptance of the idea of aggression as simply parallel to sexuality in all respects, there are major discrepancies. Perhaps aggression cannot be viewed as a drive after all; perhaps our drive construct needs to be modified to accommodate aggression. Certainly, controversy in this area has interfered with the production of good clinical studies which could begin to increase our understanding of aggression and its place in the human personality. The psychoanalytic theory of drive development has probably undergone less change in the last

  12. 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)

  13. Control systems for ITER diagnostics, heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Simrock, Stefan, E-mail: stefan.simrock@iter.org

    2016-11-15

    The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170 GHz gyrotrons and 5 steerable launchers to deliver 20 MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55 MHz sources and two multi-element launchers to deliver 20 MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5 MW of 1 MeV beams to the plasma. Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design. The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing

  14. 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

  15. The design of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.; Kim, J.; Hwang, S.M.

    1999-01-01

    The Korea superconducting tokamak advanced research (KSTAR) project is the major effort of the Korean national fusion program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 MA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron-cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002. (orig.)

  16. Two-dimensional effects in the problem of tearing modes control by electron cyclotron current drive

    International Nuclear Information System (INIS)

    Comisso, L.; Lazzaro, E.

    2010-01-01

    The design of means to counteract robustly the classical and neoclassical tearing modes in a tokamak by localized injection of an external control current requires an ever growing understanding of the physical process, beyond the Rutherford-type zero-dimensional models. Here a set of extended magnetohydrodynamic nonlinear equations for four continuum fields is used to investigate the two-dimensional effects in the response of the reconnecting modes to specific inputs of the localized external current. New information is gained on the space- and time-dependent effects of the external action on the two-dimensional structure of magnetic islands, which is very important to formulate applicable control strategies.

  17. Tokamak power systems studies at ANL

    International Nuclear Information System (INIS)

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-01-01

    A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

  18. On the NBI system for substantial current drive in a fusion power plant: Status and R and D needs for ion source and laser neutralizer

    Energy Technology Data Exchange (ETDEWEB)

    Franzen, P., E-mail: peter.franzen@ipp.mpg.de; Fantz, U.

    2014-11-15

    Highlights: • NBI is a candidate for a cw tokamak DEMO due to its high current drive efficiency. • The plug-in efficiency must be improved from the present 20–30% to more than 50%. • A suitable candidate is a photo neutralizer with almost 100% neutralization efficiency; basic feasibility studies are underway. • Cw operation with a large availability puts rather high demands on source operation with some safety margins, especially for the components with high power density loads (source back plate and extraction system). • Alternatives to the present use of cesium are under exploitations. - Abstract: The requirements for the heating and current drive systems of a fusion power plant will strongly depend on the DEMO scenario. The paper discusses the R and D needs for a neutral beam injection system — being a candidate due to the highest current drive efficiency — for the most demanding scenario, a steady state tokamak DEMO. Most important issues are the improvement of the wall-plug efficiency from the present ∼25% to the required 50–60% by improving the neutralization efficiency with a laser neutralizer system and the improvement of the reliability of the ion source operation. The demands on and the potential of decreasing the ion source operation pressure, as well as decreasing the amount of co-extracted electrons and backstreaming ions are discussed using the ITER requirements and solutions as basis. A further concern is the necessity of cesium for which either the cesium management must be improved or alternatives to cesium for the production of negative ions have to be identified.

  19. Plasma simulation by macroscale, electromagnetic particle code and its application to current-drive by relativistic electron beam injection

    International Nuclear Information System (INIS)

    Tanaka, M.; Sato, T.

    1985-01-01

    A new implicit macroscale electromagnetic particle simulation code (MARC) which allows a large scale length and a time step in multi-dimensions is described. Finite mass electrons and ions are used with relativistic version of the equation of motion. The electromagnetic fields are solved by using a complete set of Maxwell equations. For time integration of the field equations, a decentered (backward) finite differencing scheme is employed with the predictor - corrector method for small noise and super-stability. It is shown both in analytical and numerical ways that the present scheme efficiently suppresses high frequency electrostatic and electromagnetic waves in a plasma, and that it accurately reproduces low frequency waves such as ion acoustic waves, Alfven waves and fast magnetosonic waves. The present numerical scheme has currently been coded in three dimensions for application to a new tokamak current-drive method by means of relativistic electron beam injection. Some remarks of the proper macroscale code application is presented in this paper

  20. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2005-01-01

    SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

  1. Plea for stellarator funding raps tokamaks

    International Nuclear Information System (INIS)

    Blake, M.

    1992-01-01

    The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

  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. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks. Annual report, November 16, 1991--November 15, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Scharer, J.E.

    1992-12-31

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  4. 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.

  5. Radio-frequency current drive efficiency in the presence of ITBs and a dc electric field

    International Nuclear Information System (INIS)

    Rosa, P.R. da S; Mourao, R.; Ziebell, L.F.

    2009-01-01

    This paper discusses the current drive efficiency by the combined action of EC and LH waves in the presence of a dc electric field and transport, with an internal transport barrier. The transport is assumed to be produced by magnetic fluctuations. The study explores the different barrier parameters and their influence on the current drive efficiency. We study the subject by numerically solving the Fokker-Planck equation. Our main result is that the barrier depth and barrier width are important to determine the correct shape of the current density profile but not to determine the current drive efficiency, which is very little influenced by these parameters. We also found similar results for the influence of the level of magnetic fluctuations on the current density profile and on the current drive efficiency.

  6. 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)

  7. Design and Construction of Variable Direct Current Speed Drive ...

    African Journals Online (AJOL)

    controlled rectifiers from the viewpoint of simplicity and cost effectiveness to act as power converter and controller. Design and construction of constituent circuits such as acceleration/deceleration, speed and current amplifier and the trigger ...

  8. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. 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

  12. 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

  13. Suppression of the Neoclassical Tearing Modes in Tokamaks under Anomalous Transverse Transport Conditions when the Magnetic Well Effect Predominates over the Bootstrap Drive

    International Nuclear Information System (INIS)

    Konovalov, S.V.; Mikhailovskii, A.B.; Shirokov, M.S.; Ozeki, T.; Tsypin, V.S.

    2005-01-01

    A study is made of the suppression of neoclassical tearing modes in tokamaks under anomalous transverse transport conditions when the magnetic well effect predominates over the bootstrap drive. It is stressed that the corresponding effect, which is called the compound suppression effect, depends strongly on the profiles of the electron and ion temperature perturbations. Account is taken of the fact that the temperature profile can be established as a result of the competition between anomalous transverse heat transport, on the one hand, and longitudinal collisional heat transport, longitudinal heat convection, longitudinal inertial transport, and transport due to the rotation of magnetic islands, on the other hand. The role of geodesic effects is discussed. The cases of competition just mentioned are described by the model sets of reduced transport equations, which are called, respectively, collisional, convective, inertial, and rotational plasmophysical models. The magnetic well is calculated with allowance for geodesic effects. It is shown that, for strong anomalous heat transport conditions, the contribution of the magnetic well to the generalized Rutherford equation for the island width W is independent of W not only in the collisional model (which has been investigated earlier) but also in the convective and inertial models and depends very weakly (logarithmically) on W in the rotational model. It is this weak dependence that gives rise to the compound effect, which is the subject of the present study. A criterion for the stabilization of neoclassical tearing modes by the compound effect at an arbitrary level of the transverse heat transport by electrons and ions is derived and is analyzed for two cases: when the electron heat transport and ion heat transport are both strong, and when the electron heat transport is strong and the ion heat transport is weak

  14. Control of plasma position in TEXTOR tokamak applying a method of multipole moments of plasma current

    International Nuclear Information System (INIS)

    Kardon, B.; Soltwitsch, H.; Waidmann, G.

    1987-07-01

    On the basis of a method developed by Zakharov and Shafranov for the determination of the plasma cross section in toroidal geometry two special pick-up coils for the two components of the poloidal magnetic field were built. A cosine coil of higher order and a saddle coil were designed and installed outside the plasma column. The proper sum of these two signals permits the evaluation of the first momentum of the plasma current which is the equilibrium plasma displacement. The horizontal position of the plasma column was controlled on-line by the TEXTOR feedback system using this analog sum signal. (orig.)

  15. The spheric tokamak programme at Culham

    International Nuclear Information System (INIS)

    Sykes, A.

    1999-01-01

    The Spherical Tokamak (ST) is the low aspect ratio limit of the conventional tokamak, and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration, and was operational at Culham from January 1991 to March 1998, obtaining plasma current of up to 300 kA and pulse durations of ∼ 50 ms. Its successor, MAST is scheduled to obtain first plasma in Autumn 1998 and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high-performance central solenoid, 1.5 MW ECRH and 5 MW of Neutral Beam Injection. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose-built STs (such as MAST) are described. (author)

  16. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  17. High-beta tokamak research. Annual progress report, August 1, 1983-July 30, 1984

    International Nuclear Information System (INIS)

    Navratil, G.A.

    1984-08-01

    Our main research objectives during the past year fell into four areas: (1) construction and initial operation of the new tokamak, HBT; (2) further numerical modeling of the Torus II experimental equilibria using the PPPL equilibrium and stability codes; (3) diagnostic development; and (4) ICRF antenna coupling calculation in 2D and rf current drive

  18. Effects of passive components on the input current interharmonics of adjustable-speed drives

    DEFF Research Database (Denmark)

    Soltani, Hamid; Blaabjerg, Frede; Zare, Firuz

    2016-01-01

    Current and voltage source Adjustable Speed Drives (ASDs) exert distortion current into the grid, which may produce some interharmonic components other than the characteristic harmonic components. This paper studies the effects of passive components on the input current interharmonics of adjustable...... speed drives with and/or without motor current imbalance. The investigation is done at different motor operating frequencies and load torque values. It shows that selecting the small filter components (ac choke, dc choke and dc-link capacitor) results in different performances in respect to those...... interharmincs issued by motor current imbalance and other non-characteristic interharmonics. The results are helpful for engineers investigating the effects of drive filters on the input current interharmonic components....

  19. Present status of TCA/BR Tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Galvao, R.M.O.; Tuszel, A.G.

    1997-01-01

    The TCA tokamak is being partially reconstructed and reassembled in the Plasma Laboratory of The University of Sao Paulo, and afterwards it will be named TCA/BR. The first discharges are expected by June/July of next year. The main scientific objectives envisaged for the machine are: Alfven wave heating and current drive, confinement improvement, disruptions and turbulence. In this paper we also describe: (i) the present status of the project; (ii) the diagnostic system; (iii) the control and data acquisition system; (iv) the RF system for the excitation of Alfven waves, that are being developed, and also the results of predictive transport simulations of its performance. (author)

  20. The spherical tokamak fusion power plant

    International Nuclear Information System (INIS)

    Wilson, H.R.; Voss, G.; Ahn, J.W.

    2003-01-01

    The design of a 1GW(e) steady state fusion power plant, based on the spherical tokamak concept, has been further iterated towards a fully self-consistent solution taking account of plasma physics, engineering and neutronics constraints. In particular a plausible solution to exhaust handling is proposed and the steam cycle refined to further improve efficiency. The physics design takes full account of confinement, MHD stability and steady state current drive. It is proposed that such a design may offer a fusion power plant which is easy to maintain: an attractive feature for the power plants following ITER. (author)

  1. The drive to strive: goal generation based on current needs

    Directory of Open Access Journals (Sweden)

    Elisabeth A Murray

    2013-06-01

    Full Text Available Hungry animals are influenced by a multitude of different factors when foraging for sustenance. Much of the work on animal foraging has focused on factors relating to the amount of time and energy animals expend searching for and harvesting foods. Models that emphasize such factors have been invaluable in determining when it is beneficial for an animal to search for pastures new. When foraging, however, animals also have to determine how to direct their search. For what food should they forage? There is no point searching for more of a particular food when you are sated from eating it. Here we review work in macaques and humans that has sought to reveal the neural circuits critical for determining the subjective value of different foods and associated objects in our environment and tracking this value over time. There is mounting evidence that a network composed of the orbitofrontal cortex (OFC, amygdala and medial thalamus is critical for linking objects in the environment with food value and adjusting those valuations in real time based on current biological needs. Temporal inactivation studies have revealed that the amygdala and OFC play distinct, but complementary roles in this valuation process. Such a network for determining the subjective value of different foods and, by extension, associated objects, must interact with systems that determine where and for how long to forage. Only by efficiently incorporating these two factors into their decisions will animals be able to achieve maximal fitness.

  2. GENERATING OF OPTIMAL QUANTIZATION LEVELS OF CONTROL CURRENTS FOR LINEAR STEPPING DRIVES OF PRECISION MOTION SYSTEMS

    Directory of Open Access Journals (Sweden)

    I. V. Dainiak

    2014-01-01

    Full Text Available The paper proposes a method of taking into account accumulated and temperature errors while forming coordinate discrete grid of a linear stepping drive. An algorithm for determination of optimal quantization levels of control currents of drive's phases has been developed in the paper; it minimizes an error of positioning that forms correction files for application of a control system in the software. Investigations on stability of discrete grid nodes coordinates have been carried our with the help of a monitoring station for accurate parameters of linear stepping drive. The investigations have proved an efficiency of the proposed algorithm and methodology for forming coordinate discrete grid.

  3. Lower-hybrid poloidal current drive for fluctuation reduction in a reversed field pinch

    International Nuclear Information System (INIS)

    Uchimoto, E.; Cekic, M.; Harvey, R.W.; Litwin, C.; Prager, S.C.; Sarff, J.S.; Sovinec, C.R.

    1994-06-01

    Current drive using the lower-hybrid slow wave is shown to be a promising candidate for improving confinement properties of a reversed field pinch (RFP). Ray-tracing calculations indicate that the wave will make a few poloidal turns while spiraling radially into a target zone inside the reversal layer. The poloidal antenna wavelength of the lower hybrid wave can be chosen so that efficient parallel current drive will occur mostly in the poloidal direction in this outer region. Three-dimensional resistive magnetohydrodynamic (MHD) computation demonstrates that an additive poloidal current in this region will reduce the magnetic fluctuations and magnetic stochasticity

  4. Modeling of the electron distribution based on bremsstrahlung emission during lower hybrid current drive on PLT

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J.E.; von Goeler, S.; Bernabei, S.; Bitter, M.; Chu, T.K.; Efthimion, P.; Fisch, N.; Hooke, W.; Hosea, J.; Jobes, F.

    1985-03-01

    Lower hybrid current drive requires the generation of a high energy electron tail anisotropic in velocity. Measurements of bremsstrahlung emission produced by this tail are compared with the calculated emission from reasonable model distributions. The physical basis and the sensitivity of this modeling process are described and the plasma properties of current driven discharges which can be derived from the model are discussed.

  5. Modeling of the electron distribution based on bremsstrahlung emission during lower hybrid current drive on PLT

    International Nuclear Information System (INIS)

    Stevens, J.E.; von Goeler, S.; Bernabei, S.

    1985-03-01

    Lower hybrid current drive requires the generation of a high energy electron tail anisotropic in velocity. Measurements of bremsstrahlung emission produced by this tail are compared with the calculated emission from reasonable model distributions. The physical basis and the sensitivity of this modeling process are described and the plasma properties of current driven discharges which can be derived from the model are discussed

  6. Closure of the single fluid magnetohydrodynamic equations in presence of electron cyclotron current drive

    NARCIS (Netherlands)

    Westerhof, E.; Pratt, J.

    2014-01-01

    In the presence of electron cyclotron current drive (ECCD), the Ohm's law of single fluid magnetohydrodynamics is modified as E + v × B = η(J – J EC). This paper presents a new closure relation for the EC driven current density appearing in this modified Ohm's law. The new relation

  7. Control of plasma profiles and stability through localised Electron Cyclotron Current Drive

    NARCIS (Netherlands)

    Merkulov, Oleksiy

    2006-01-01

    The work presented in this thesis addresses several topics from the physics of the magnetically confined plasma inside a tokamak. At the moment, the tokamak is the most successful concept for becoming a future thermonuclear reactor. However, there are plenty of physics and engineering problems to

  8. Analytical calculation of current drive synergy between LH and EC waves

    International Nuclear Information System (INIS)

    Dumont, R.; Giruzzi, G.

    2001-01-01

    An analytical model for the evaluation of electron cyclotron current drive efficiency improvement in lower hybrid current drive regimes is presented. The adjoint equation is written and solved by a perturbation treatment, allowing to derive a response function including both collisional and lower hybrid effects, in the limit where the former still dominate. This allows an analytical demonstration of the current drive synergy effects, previously found by numerical solutions of the kinetic equation. The model is especially useful for the determination of appropriate wave parameters optimizing this synergy effect, such as the EC launching angles suitable for a given LH target plasma. Under these conditions, it is shown that a significant improvement of the ECCD efficiency can be obtained

  9. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...wavelength polariza- tion field produced by the curvature and field gradient drifts [15]. The growth rate is y = Vs[k/R 2 = [T(eV)/X(cm)J 2 3.3 x 105 sec

  10. Recent Improvements to the Control of the CTF3 High-Current Drive Beam

    CERN Document Server

    Constance, B; Gamba, D; Skowronski, P K

    2013-01-01

    In order to demonstrate the feasibility of the CLIC multiTeV linear collider option, the drive beam complex at the CLIC Test Facility (CTF3) at CERN is providing highcurrent electron pulses for a number of related experiments. By means of a system of electron pulse compression and bunch frequency multiplication, a fully loaded, 120 MeV linac is used to generate 140 ns electron pulses of around 28 Amperes. Subsequent deceleration of this high-current drive beam demonstrates principles behind the CLIC acceleration scheme, and produces 12 GHz RF power for experimental purposes. As the facility has progressed toward routine operation, a number of studies aimed at improving the drive beam performance have been carried out. Additional feedbacks, automated steering programs, and improved control of optics and dispersion have contributed to a more stable, reproducible drive beam with consequent benefits for the experiments.

  11. Analysis of JET LCHD/ICRH synergy experiments in terms of relativistic current drive theory

    International Nuclear Information System (INIS)

    Start, D.F.H.; Baranov, Y.; Brusati, M.; Ekedahl, A.; Froissard, P.; Gormezano, C.; Jacquinot, J.; Paquin, L.; Rimini, F.G.; Di Vita, A.

    1994-01-01

    The present analysis shows that the observed efficiency of current drive with synergy between LHCD and ICRH is in good agreement with the relativistic theory of Karney and Fisch for Landau damped waves. The predicted power absorption from the fast wave by the electron tail is within 30% of the measured value. In the presence of significant fast electron diffusion within a slowing down time it would be possible to produce central current drive using multiple ICRF resonances even when the LHCD deposition is at half radius, as in an ITER type device. (authors). 4 refs., 6 figs

  12. A Smart Current Modulation Scheme for Harmonic Reduction in Three- Phase Motor Drive Applications

    DEFF Research Database (Denmark)

    Davari, Pooya; Zare, Firuz; Blaabjerg, Frede

    2015-01-01

    harmonic mitigation methods have been developed over the years, the total cost and complexity has become the main obstacle in employing prior-art methods for motor drive systems. This paper presents a novel current modulation method based on the electronic inductor concept for three-phase ac-dc systems......Electric motor-driven systems consume considerable amount of the global electricity. Majority of three-phase motor drives are equipped with conventional diode rectifier and passive harmonic mitigation, being witnessed as the main source in generating input current harmonics. While many active...

  13. Compensation methods applied in current control schemes for large AC drive systems

    DEFF Research Database (Denmark)

    Rus, D. C.; Preda, N. S.; Teodorescu, Remus

    2012-01-01

    The paper deals with modified PI current control structures for large AC drive systems which use surface mounted permanent magnet synchronous machines or squirrel-cage induction motors supplied with voltage source inverters. In order to reduce the power losses caused by high frequency switching...... of the semiconductor devices, various compensation methods are used and a modified structure for a PI current controller is proposed, to reduce the switching frequency of the inverter for the same operating frequency of the drive. Simulation, experimental development and test results are presented in order...

  14. Start-up of spherical tokamak without a center solenoid

    International Nuclear Information System (INIS)

    Maekawa, Takashi; Nagata, Masayoshi

    2012-01-01

    For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)

  15. Theory and experiments on RF plasma heating, current drive and profile control in TORE SUPRA

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, D.

    1994-01-01

    This paper reviews the main experimental and theoretical achievements related to the study of RF heating and non-inductive current drive and particularly phenomena related to the current density profile control and the potentiality of producing stationary enhanced performance regimes: description of the Lower Hybrid (LH) and Ion Cyclotron Resonant Frequency (ICRF) systems; long pulse coupling performance of the RF systems; observation of the transition to the so-called ``stationary LHEP regime`` in which the (flat) central current density and (peaked) electron temperature profiles are fully decoupled; experiments on ICRF sawtooth stabilization with the combined effect of LHCD modifying the current density profile; diffusion of fast electrons generated by LH waves; ramp-up experiments in which the LH power provided a significant part of the resistive poloidal flux and flux consumption scaling; theory of spectral wave diffusion and multipass absorption; fast wave current drive modelling with the Alcyon full wave code; a reflector LH antenna concept. 18 figs., 48 refs.

  16. Theory and experiments on RF plasma heating, current drive and profile control in TORE SUPRA

    International Nuclear Information System (INIS)

    Moreau, D.

    1994-01-01

    This paper reviews the main experimental and theoretical achievements related to the study of RF heating and non-inductive current drive and particularly phenomena related to the current density profile control and the potentiality of producing stationary enhanced performance regimes: description of the Lower Hybrid (LH) and Ion Cyclotron Resonant Frequency (ICRF) systems; long pulse coupling performance of the RF systems; observation of the transition to the so-called ''stationary LHEP regime'' in which the (flat) central current density and (peaked) electron temperature profiles are fully decoupled; experiments on ICRF sawtooth stabilization with the combined effect of LHCD modifying the current density profile; diffusion of fast electrons generated by LH waves; ramp-up experiments in which the LH power provided a significant part of the resistive poloidal flux and flux consumption scaling; theory of spectral wave diffusion and multipass absorption; fast wave current drive modelling with the Alcyon full wave code; a reflector LH antenna concept. 18 figs., 48 refs

  17. Complex state variable- and disturbance observer-based current controllers for AC drives

    DEFF Research Database (Denmark)

    Dal, Mehmet; Teodorescu, Remus; Blaabjerg, Frede

    2013-01-01

    , extracted by a disturbance observer and then injected into the current controller. In this study, a revised version of a disturbance observer-based controller and a well known complex variable model-based design with a single set of complex pole are compared in terms of design aspects and performance......In vector-controlled AC drives, the design of current controller is usually based on a machine model defined in synchronous frame coordinate, where the drive performance may be degraded by both the variation of the machine parameters and the cross-coupling between the d- and q-axes components...... of the stator current. In order to improve the current control performance an alternative current control strategy was proposed previously aiming to avoid the undesired cross-coupling and non-linearities between the state variables. These effects are assumed as disturbances arisen in the closed-loop path...

  18. Lower hybrid heating and current drive in ignitor shear reversal scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Barbato, E.; Pinaccione, L. [Italian Agengy for New Technologies, Energy and the Environment, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-05-01

    Injection of Lower Hybrid (LH) Wave power at 8 GHz is considered into IGNITOR shear reversal scenarios, characterized by a reduced plasma current and density. Power deposition calculation are performed to establish whether LH waves can be used both as central heating and off axis current drive tool. It turns out that LH waves can be used (a) for central plasma heating purpose during the current vamp phase, to freeze the shear reversed configuration, at the power level of {approx}10 MW. (b) to drive a current in the outer part of the plasma at the power level of 20 MW. In this way around 1/3-1/6 of the total current in the proper plasma position (i.e. where q is minimum) is driven.

  19. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

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

  20. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Gilleland, J.R.

    1988-01-01

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  1. 60 MHz fast wave current drive experiment for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, M.J.; Chiu, S.C.; Porkolab, M.; Chan, V.; Freeman, R.; Harvey, R.; Pinsker, R. (General Atomics, San Diego, CA (USA))

    1989-07-01

    The DIII-D facility provides an opportunity to test fast wave current drive appoach. Efficient FWCD is achieved by direct electron absorption due to Landa damping and transit time magnetic pumping. To avoid competing damping mechamisms we seek to maximize the single-pass asorption of the fast waves by electrons. (AIP)

  2. Advanced launcher design options for electron cyclotron current drive on ITER based on remote steering

    NARCIS (Netherlands)

    Graswinckel, M. R.; Bongers, W. A.; M.R. de Baar,; van den Berg, M. A.; Denisov, G.; Donne, A. J. H.; Elzendoorn, B. S. Q.; Goede, A. P. H.; Heidinger, R.; Kuzikov, S.; Kruijt, O. G.; Kruizinga, B.; Moro, A.; Poli, E.; Ronden, D. M. S.; Saibene, G.; Thoen, D. J.; Verhoeven, A. G. A.

    2008-01-01

    Electron cyclotron current drive will become the main scheme on ITER for the stabilization of neoclassical tearing modes (NTMs) and the control of sawtooth oscillations. The effectiveness of this scheme forms the basis for the requirements of the ITER Upper Port Launcher. These requirements include

  3. Heating and current-drive with high phase velocity compressional Alfven waves

    International Nuclear Information System (INIS)

    Li, Y.M.; Mahajan, S.M.; Ross, D.W.

    1986-12-01

    It is shown that high phase velocity compressional Alfven waves have the desirable features needed for efficient current drive in fusion-reactor-like conditions; the energy deposition is low on the α-particles, and high on the hot electrons in the plasma interior

  4. Globalisation and the foreignisation of space: The seven processes driving the current global land grab.

    NARCIS (Netherlands)

    Zoomers, E.B.

    2010-01-01

    The current global land grab is causing radical changes in the use and ownership of land. The main process driving the land grab, or ‘foreignisation of space’, as highlighted in the media and the emerging literature is the production of food and biofuel for export in the aftermath of recent food

  5. Tokamak power system studies at ANL

    International Nuclear Information System (INIS)

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-06-01

    The following features, in particular, have been examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I ≅ 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

  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. Statistical analysis of first period of operation of FTU Tokamak

    International Nuclear Information System (INIS)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S.

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted

  8. Wave-driver options for low-aspect-ratio steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1981-02-01

    Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value

  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. Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU

    Energy Technology Data Exchange (ETDEWEB)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.

  11. Hysteretic self-oscillating bandpass current mode control for Class D audio amplifiers driving capacitive transducers

    DEFF Research Database (Denmark)

    Nielsen, Dennis; Knott, Arnold; Andersen, Michael A. E.

    2013-01-01

    A hysteretic self-oscillating bandpass current mode control (BPCM) scheme for Class D audio amplifiers driving capacitive transducers are presented. The scheme provides excellent stability margins and low distortion over a wide range of operating conditions. Small-signal behavior of the amplifier...... is analysis through transfer function based linear control methodology. Measurements are performed on a single-ended ± 300 V half-bridge amplifier driving a capacitive load of 100 nF. Total Harmonic Distortion plus noise (THD+N) below 0.1 % are reported. Transducers representing a capacitive load and obeying...... the rules of electrostatics have been known as very interesting alternatives to the traditional inefficient electrodynamic transducers. When driving capacitive transducers from a Class D audio amplifier the high impedance nature of the load represents a key challenge. The BPCM control scheme ensures a flat...

  12. An analysis of JET fast-wave heating and current drive experiments directly related to ITER

    International Nuclear Information System (INIS)

    Bhatnagar, V.P.; Eriksson, L.; Gormezano, C.; Jacquinot, J.; Kaye, A.; Start, D.F.H.

    1994-01-01

    The ITER fast-wave system is required to serve a variety of purposes, in particular, plasma heating to ignition, current profile and burn control and eventually, in conjunction with other schemes, a central non-inductive current drive (CD) for the steady-state operation of ITER. The ICRF heating and current drive data that has been obtained in JET are analyzed in terms of dimensionless parameters, with a view to ascertaining its direct relevance to key ITER requirements. The analysis is then used to identify areas both in physics and technological aspects of ion-cyclotron resonance heating (ICRH) and CD that require further experimentation in ITER-relevant devices such as JET to establish the required data base. (authors). 12 refs., 8 figs

  13. Summary and viewgraphs from the Q-121 US/Japan advanced current drive concepts workshop

    International Nuclear Information System (INIS)

    Bonoli, P.; Porkolab, M.; Chan, V.; Pinsker, R.; Politzer, P.; Darrow, D.; Fukuyama, Atsushi; Imai, Tsuyoshi; Watari, Tetsuo; Itoh, Satoshi; Nakamura, Yukio; James, R.; Logan, G.; Porter, G.; Thomassen, K.; Lyon, J.; Mau, Tak; Tanaka, Hitoshi; Tanaka, Shigetoshi

    1990-01-01

    With the emphasis placed on current drive by ITER, which requires steady state operation in its engineering phase, it is important to bring theory and experiment in agreement for each of the schemes that could be used in that design. Both neutral beam and lower hybrid (LH) schemes are in excellent shape in that regard. Since the projected efficiency of all schemes is marginal it is also important to continue our search for more efficient processes. This workshop featured experimental and theoretical work in each processes. This workshop featured experimental and theoretical work in each of these areas, that is, validation of theory and the search for better ideas. There were a number of notable results to report, the most striking again (as with last year) the long pulse operation of TRIAM-1M. A low current was sustained for over 1 hour with LH waves, using new hall-effect sensors in the equilibrium field circuit to maintain position control. In JT-60, by sharpening the wave spectrum the current drive efficiency was improved to 0.34 x 10 20 m -2 A/W and 1.5 MA of current was driven entirely by the lower hybrid system. Also in that machine, using two different LH frequencies, the H-mode was entered. Finally, by using the LH system for startup they saved 2.5 resistive volt-sec of flux, which if extrapolated to ITER would save 40 volt-sec there. For the first time, and experiment on ECH current drive showed reasonable agreement with theory. Those experiments are reported here by James (LLNL) on the D3-D machine. Substantially lower ECH current drive than expected theoretically was observed on WT-3, but if differed by being in a low absorption regime. Nonetheless, excellent physics results were achieved in the WT-3 experiments, notably in having careful measurements of the parallel velocity distributions

  14. ENHANCING THE OPERATIONAL EFFICIENCY OF DIRECT CURRENT DRIVE BASED ON USE OF SUPERCONDENSER POWER STORAGE UNITS

    Directory of Open Access Journals (Sweden)

    А. M. Mukha

    2017-10-01

    Full Text Available Purpose.The scientific work is intended to analyse the expansion of the load range and the implementation of regeneration braking (RB of the direct current drive by using the supercondenser power storage units. Methodology.To solve the problem, we use the methods of the electric drive theory, impulse electronics and the method of calculation of transient electromagnetic processes in linear electric circuits in the presence of super-condensers therein. Findings.The stiffness of the mechanical and electromechanical characteristics of a series motor is significantly increased, which makes it possible to use a DC drive under load, much smaller than 15…20% of the nominal one. Numerical calculations of the operation process of the supercondenser power storage unit were fulfilled with a sharp decrease in the load of a traction electric motor of a direct current electric locomotive. The possibility of RB of the direct current drive with the series motor is substantiated. The equations of the process of charging and discharging of super-condenser storage unit in RB mode are solved. The authors examined the effect of capacitance on the nature of maintaining the excitation current of an electric motor in the mode of small loads.Originality.The paper developed theoretical approaches for the transformation of soft (mechanical and electromechanical characteristics into hard ones of DC series motors. For the first time a new, combined method of the series motor RB is proposed and substantiated. Further development obtained the methods for evaluating the storage unit parameters, taking into account the criteria for reliable parallel operation of super-condensers with an electric motor field. Practical value.The proposed and substantiated transformation of soft characteristics into stiff ones allows us to use general-purpose electric drives with series motors and at low loads, and in traction electric drives - to reduce the intensity of electric stockwheel

  15. A Phase Current Reconstruction Approach for Three-Phase Permanent-Magnet Synchronous Motor Drive

    Directory of Open Access Journals (Sweden)

    Hao Yan

    2016-10-01

    Full Text Available Three-phase permanent-magnet synchronous motors (PMSMs are widely used in renewable energy applications such as wind power generation, tidal energy and electric vehicles owing to their merits such as high efficiency, high precision and high reliability. To reduce the cost and volume of the drive system, techniques of reconstructing three-phase current using a single current sensor have been reported for three-phase alternating current (AC control system using the power converts. In existing studies, the reconstruction precision is largely influenced by reconstructing dead zones on the Space Vector Pulse Width Modulation (SVPWM plane, which requires other algorithms to compensate either by modifying PWM modulation or by phase-shifting of the PWM signal. In this paper, a novel extended phase current reconstruction approach for PMSM drive is proposed. Six novel installation positions are obtained by analyzing the sampling results of the current paths between each two power switches. By arranging the single current sensor at these positions, the single current sensor is sampled during zero voltage vectors (ZVV without modifying the PWM signals. This proposed method can reconstruct the three-phase currents without any complex algorithms and is available in the sector boundary region and low modulation region. Finally, this method is validated by experiments.

  16. Current Reversal Due to Coupling Between Asymmetrical Driving Force and Ratchet Potential

    International Nuclear Information System (INIS)

    Ai Baoquan; Xie Huizhang; Liu Lianggang

    2006-01-01

    Transport of a Brownian particle moving in a periodic potential is investigated in the presence of an asymmetric unbiased external force. The asymmetry of the external force and the asymmetry of the potential are the two ways of inducing a net current. It is found that the competition of the spatial asymmetry of potential with the temporal asymmetry of the external force leads to the phenomena like current reversal. The competition between the two opposite driving factors is a necessary but not a sufficient condition for current reversals.

  17. Noise and driving induced Brownian heat current in the Frenkel-Kontorova lattices

    Science.gov (United States)

    Chen, Chongyang; Chen, Ruyin; Nie, Linru; Wang, Chaojie; Jia, Youjian

    2018-02-01

    We investigate Brownian heat current induced by noise and driving in the heat conduction of Frenkel-Kontorova (FK) lattices, whose two ends contact respectively with two baths with temporally. Our numerical results indicate that thermal current of the system versus environmental reference temperature exhibits two peaks in symmetric cases, i.e., thermal SR. In asymmetric cases, the phenomenon also takes place, depending on on-site potentials of the atoms exerted on by the periodic signal, but disappears for harmonic lattices. Moreover, its intrinsic physical mechanism is analyzed in detail. The above results play important roles in controlling thermal current of nonlinear lattices.

  18. Application of Electron Bernstein Wave heating and current drive to high beta plasmas

    International Nuclear Information System (INIS)

    Efthimion, P.C.

    2002-01-01

    Electron Bernstein Waves (EBW) can potentially heat and drive current in high-beta plasmas. Electromagnetic waves can convert to EBW via two paths. O-mode heating, demonstrated on W-7AS, requires waves be launched within a narrow k-parallel range. Alternately, in high-beta plasmas, the X-mode cutoff and EBW conversion layers are millimeters apart, so the fast X-mode can tunnel to the EBW branch. We are studying the conversion of EBW to the X-mode by measuring the radiation temperature of the cyclotron emission and comparing it to the electron temperature. In addition, mode conversion has been studied with an approximate kinetic full-wave code. We have enhanced EBW mode conversion to ∼ 100% by encircling the antenna with a limiter that shortens the density scale length at the conversion layer in the scrape off of the CDX-U spherical torus (ST) plasma. Consequently, a limiter in front of a launch antenna achieves efficient X-mode coupling to EBW. Ray tracing and Fokker-Planck codes have been used to develop current drive scenarios in NSTX high-beta (∼ 40%) ST plasmas and a relativistic code will examine the potential synergy of EBW current drive with the bootstrap current. (author)

  19. Simple multijunction launcher with oversized waveguides for lower hybrid current drive on JT-60U

    International Nuclear Information System (INIS)

    Ikeda, Y.; Naito, O.; Seki, M.; Kondoh, T.; Ide, S.; Anno, K.; Fukuda, H.; Ikeda, Y.; Kitai, T.; Kiyono, K.; Sawahata, M.; Shinozaki, S.; Suganuma, K.; Suzuki, N.; Ushigusa, K.

    1994-01-01

    A multijunction technique with oversized waveguides has been developed for the lower hybrid current drive launcher on JT-60U. The launcher consists of 4 (toroidal)x4 (poloidal) multijunction modules. RF power in the module is divided toroidally into 12 sub-waveguides at a junction point through an oversized waveguide. This method simplifies the structure of the multijunction launcher with a large number of subwaveguides. A maximum power density up to 25 MW m -2 has been achieved with a low reflection coefficient of less than 2%. The coupling and current drive efficiency are well explained by the designed wave spectra without taking account of higher modes in the oversize waveguides. Thus, the simple multijunction launcher has been demonstrated to excite expected wave spectra with high power handling capability. ((orig.))

  20. Overview of physics research on the TCV tokamak

    Czech Academy of Sciences Publication Activity Database

    Fasoli, A.; Alberti, S.; Amorim, P.; Angioni, C.; Asp, E.; Behn, R.; Bencze, A.; Berrino, J.; Blanchard, P.; Bortolon, A.; Brunner, S.; Camenen, Y.; Cirant, S.; Coda, S.; Curchod, L.; DeMeijere, K.; Duval, B. P.; Fable, E.; Fasel, D.; Felici, F.; Furno, I.; Garcia, O.E.; Giruzzi, G.; Gnesin, S.; Goodman, T.; Graves, J.; Gudozhnik, A.; Gulejova, B.; Henderson, M.; Hogge, J. Ph.; Horáček, Jan; Joye, B.; Karpushov, A.; Kim, S.-H.; Laqua, H.; Lister, J. B.; Llobet, X.; Madeira, T.; Marinoni, A.; Marki, J.; Martin, Y.; Maslov, M.; Medvedev, S.; Moret, J.-M.; Paley, J.; Pavlov, I.; Piffl, Vojtěch; Piras, F.; Pitts, R.A.; Pitzschke, A.; Pochelon, A.; Porte, L.; Reimerdes, H.; Rossel, J.; Sauter, O.; Scarabosio, A.; Schlatter, C.; Sushkov, A.; Testa, D.; Tonetti, G.; Tskhakaya, D.; Tran, M. Q.; Turco, F.; Turri, G.; Tye, R.; Udintsev, V.; Véres, G.; Villard, L.; Weisen, H.; Zhuchkova, A.; Zucca, C.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104005-104005 ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : overview highlights * fusion research * tokamak TCV * self-generated current * H-mode physics * Electron internal transport barrier * electron cyclotron heating * electron cyclotron current drive physics * density peaking * MHDactivity * edge physics * reciprocating Mach probe * Pfirsch–Schlueter component. Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://stacks.iop.org/NF/49/104005

  1. 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

  2. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    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{sup 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.

  3. Conceptual design of a Tokamak hybrid power reactor (THPR)

    International Nuclear Information System (INIS)

    Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.

    1987-01-01

    A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity

  4. A comparison of burn cycle options for Tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Consideration of burn cycle options for commercial tokamaks shows that there is substantial motivation to achieve steady state operation. This is partly due to longer replacement periods for first wall and impurity control components, but, in addition, large cost savings are found when magnets, power supplies, and the energy transfer system are not frequently pulsed. The hybrid burn cycle, with a combination of ohmic and noninductive current drive, does not significantly improve the economics of ohmically-driven commercial reactors with large major radius. However, an INTOR-class device has a critically small hole in the doughnut, and we find for this size tokamak that the hybrid cycle is preferred over ohmically-driven operation

  5. Coupling of tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Finn, J.M.

    1977-01-01

    The simultaneous presence of tearing modes of different helical pitches leads to the destruction of magnetic surfaces, which has been suggested as the mechanism leading to the onset of the disruptive instability in tokamaks. For current profiles in which the m = 2 mode is unstable, but the m = 3 is stable, the coupling of the m = 3 to the m = 2 through the poloidal variation of the toroidal field can drive the m = 3 amplitude psi 3 to order psi 2 times the inverse aspect ratio. Detailed calculations, both analytical and numerical, have been performed for two models for the equilibrium and m = 2 mode structure. A slab model and incompressible m = 3 perturbations are assumed. The m = 3 amplitude increases with shear, up to a point, showing that as the current channel shrinks, overlap of resonances becomes more likely. The results also apply qualitatively to other m, m +- 1 interactions

  6. 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)

  7. 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.

  8. Load Torque Compensator for Model Predictive Direct Current Control in High Power PMSM Drive Systems

    DEFF Research Database (Denmark)

    Preindl, Matthias; Schaltz, Erik

    2010-01-01

    the use of a current controller which takes into account the discrete states of the inverter, e.g. DTC or a more modern approach: Model Predictive Direct Current Control (MPDCC). Moreover overshoots and oscillations in the speed are not desired in many applications, since they lead to mechanical stress......In drive systems the most used control structure is the cascade control with an inner torque, i.e. current and an outer speed control loop. The fairly small converter switching frequency in high power applications, e.g. wind turbines lead to modest speed control performance. An improvement bring...... behaviour. It compensates the load torque influence on the speed control setting a feed forward torque value, i.e. current reference value. The benefits are twice. The speed controller reaches immediately the speed reference value avoiding offsets which must be compensated by the weak integrator. Moreover...

  9. Method for producing silicon thin-film transistors with enhanced forward current drive

    Science.gov (United States)

    Weiner, Kurt H.

    1998-01-01

    A method for fabricating amorphous silicon thin film transistors (TFTs) with a polycrystalline silicon surface channel region for enhanced forward current drive. The method is particularly adapted for producing top-gate silicon TFTs which have the advantages of both amorphous and polycrystalline silicon TFTs, but without problem of leakage current of polycrystalline silicon TFTs. This is accomplished by selectively crystallizing a selected region of the amorphous silicon, using a pulsed excimer laser, to create a thin polycrystalline silicon layer at the silicon/gate-insulator surface. The thus created polysilicon layer has an increased mobility compared to the amorphous silicon during forward device operation so that increased drive currents are achieved. In reverse operation the polysilicon layer is relatively thin compared to the amorphous silicon, so that the transistor exhibits the low leakage currents inherent to amorphous silicon. A device made by this method can be used, for example, as a pixel switch in an active-matrix liquid crystal display to improve display refresh rates.

  10. Three-wave interaction during electron cyclotron resonance heating and current drive

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Jacobsen, Asger Schou; Hansen, Søren Kjer

    2016-01-01

    Non-linear wave-wave interactions in fusion plasmas, such as the parametric decay instability (PDI) of gyrotron radiation, can potentially hamper the use of microwave diagnostics. Here we report on anomalous scattering in the ASDEX Upgrade tokamak during electron cyclotron resonance heating exper...

  11. 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)

  12. Thyristor-based current-fed drive with direct power control for permanent magnet-assisted synchronous reluctance generator

    Science.gov (United States)

    Baek, J.; Kwak, S.-S.; Toliyat, H. A.

    2015-03-01

    This paper proposes a robust and simple direct power control (DPC) of a thyristor-based current-fed drive for generator applications. A current-fed drive and permanent magnet-assisted synchronous reluctance generator (PMa-SynRG) are investigated to deliver 3 kW power using a combustion engine. The current-fed drive utilises a thyristor-based three-phase rectifier to convert generator power to DC-link power and a single-phase current-fed inverter to supply a single-phase inductive load. In addition, a new control algorithm is developed based on DPC for the current-fed drive. The DC-link voltage-based DPC is proposed in order to directly control the output power. The goal of the DPC is to maintain the DC-link voltage at the required output power operating point. The DPC has advantages such as a simple algorithm for constant speed operation. Another feature of the developed current-fed drive is its inherent capability to provide generating action by making the PMa-SynRG operates as a generator, rectifying the phase voltages by means of the three-phase rectifier and feeding the power into the load. These features make the current-fed drive a good candidate for driving any type of synchronous generators including the proposed PMa-SynRG.

  13. Destabilization of fast particle stabilized sawteeth in ASDEX Upgrade with electron cyclotron current drive

    DEFF Research Database (Denmark)

    Igochine, V.; Chapman, I.T.; Bobkov, V.

    2011-01-01

    It is often observed that large sawteeth trigger the neoclassical tearing mode well below the usual threshold for this instability. At the same time, fast particles in the plasma core stabilize sawteeth and provide these large crashes. The paper presents results of first experiments in ASDEX...... Upgrade for destabilization of fast particle stabilized sawteeth with electron cyclotron current drive (ECCD). It is shown that moderate ECCD from a single gyrotron is able to destabilize the fast particle stabilized sawteeth. A reduction in sawtooth period by about 40% was achieved in first experiments...

  14. What drives the German current account? And how does it affect other EU member states?

    OpenAIRE

    In 'T Veld, Jan; Kollmann, Robert; Ratto, Marco; Roeger, Werner; Vogel, Lukas

    2014-01-01

    We estimate a three-country model using 1995-2013 data for Germany, the Rest of the Euro Area (REA) and the Rest of the World (ROW) to analyze the determinants of Germany’s current account surplus after the launch of the Euro. The most important factors driving the German surplus were positive shocks to the German saving rate and to ROW demand for German exports, as well as German labour market reforms and other positive German aggregate supply shocks. The convergence of REA interest rates to...

  15. Analyzing Operating Behavior of Hot Mill Table Roll Drives using Statistical Methods for Current Values

    Science.gov (United States)

    Kabakov, P. Z.; Kozhevnikov, A. V.; Ilatovsky, I. S.

    2017-12-01

    The article briefly describes failures of table rolls in Hot Rolling Mill 2000, PAO Severstal, and provides research results obtained by applying statistical methods. The statistical analysis showed a possibility of detecting deviations in roll behavior based on electric drive current changes, which cause failures. The statistical data analysis was employed to define roll critical behavior, which is to be checked using additional data. For this purpose, the research will be continued, roll operating behavior will be studied during the entire lifetime with due regard for deviations detected, comprehensive statistical analysis will be carried out, influence of rolling mill practices and roll failure predictability will be assessed.

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

    DEFF Research Database (Denmark)

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

    1985-01-01

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

  17. Commissioning of the long-pulse fast wave current drive antennas for DIII-D

    International Nuclear Information System (INIS)

    Baity, F.W.; Barber, G.C.; Goulding, R.H.; Hoffman, D.J.; DeGrassie, J.S.; Pinsker, R.I.; Petty, C.C.; Cary, W.

    1995-01-01

    Two new four-element fast wave current drive antennas have been installed on DIII-D. These antennas are designed for 10-s pulses at 2 MW each in the frequency range of 30 to 120 MHz. Each element comprises two poloidal segments fed in parallel in order to optimize plasma coupling at the upper end of the frequency range. The antennas are mounted on opposite sides of the vacuum vessel, in ports designated 0 degrees and 180 degrees after their toroidal angle. Each antenna array is fed by a single transmitter. The power is first split two ways by means of a 3-dB hybrid coupler, then each of these lines feeds a resonant loop connecting a pair of array elements. The power transfer during asymmetric phasing is shunted between resonant loops by a decoupler. The resonant loops are fitted with line stretchers so that multiple frequencies of operation are possible without reconfiguring the transmission line. Commissioning of these antennas has been underway since June 1994. Several deficiencies in the transmission line system were uncovered during initial vacuum conditioning, including problems with the transmission line insulators and with the drive rods for the variable elements. The former was solved by replacing the original alumina insulators, and the latter has been avoided during operation to date by positioning the tuners to avoid high voltage appearing on the drive rods. A modified design for the drive rods will be implemented before RF operations resume operation June 1995. New transmitters were procured from ABB for the new antennas and were installed in parallel with the antenna installation. During initial vacuum conditioning of the antenna in the 180 degree port a fast digital oscilloscope was used to try to pinpoint the location of arcing by a time-of-flight technique and to develop an understanding of the typical arc signature in the system

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  19. The impurity transport in HT-6M tokamak

    International Nuclear Information System (INIS)

    Xu Wei; Wan Baonian; Xie Jikang

    2003-01-01

    The space-time profile of impurities has been measured with a multichannel visible spectroscopic detect system and UV rotation-mirror system in the HT-6M tokamak. An ideal impurity transport code has been used to simulate impurities (carbon and oxygen) behaviour during the OHM discharge. The profiles of impurities diffusion and convection coefficient, impurities ion densities in different ionized state, loss power density and effective charge number have been derived. The impurity behaviour during low-hybrid current drive has also been analyzed, the results show that the confinement of particles, impurities and energy has been improved, and emission power and effective charge number have been reduced

  20. Overview on Chinese tokamak experimental progress

    International Nuclear Information System (INIS)

    Xie, J.K.; Li, J.; Liu, Y.; Wen, Y.Z.; Wang, L.

    2001-01-01

    Tokamak experiment research in China has made important progress. The main efforts subjected to quasi-steady state operation, LHCD, plasma heating with ICRF, IBW, NBI, ECRH, fueling with pellet and supersonic molecular beam, first wall conditioning technique. Plasma parameters in experiments were much improved, such as n e =8x10 19 m -3 , plasma pulse >10Sec. ICRF boronization and conditioning made Z eff close to unit. Steady state full LH wave current drive has been achieved for more than 3 seconds. LHCD ramp up and recharge have also been demonstrated. The Best η CD exp ∼0.5(1+0.085 exp(4.8(B T -1.45))n e I CD R p /P LH =10 19 m -2 A/W. Quasi steady state H-mode like plasma with density close to Greenwald limit was obtained by LHCD, in which energy confinement time was nearly 5 times longer than the Ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macro-turbulence has been extensively carried out experimentally. Ac operation of tokamak was successfully demonstrated. (author)

  1. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  2. Compact tokamak reactors part 2 (numerical results)

    International Nuclear Information System (INIS)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-01-01

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced

  3. Compact tokamak reactors part 2 (numerical results)

    Energy Technology Data Exchange (ETDEWEB)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-10-21

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced.

  4. SIMULATION OF A HIGH-POWERED VARIABLE FREQUENCY AC DRIVE BASED ON A MULTIPULSE CURRENT SOURCE INVERTER

    Directory of Open Access Journals (Sweden)

    G.G. Zhemerov

    2013-10-01

    Full Text Available A computer model building technique for an AC drive system with an autonomous 24-pulse current inverter is introduced. On the basis of the technique developed, a MATLAB variable-frequency drive model capable of working in both quasi-stationary and transient modes is built.

  5. Non-equilibrium reactivation of Na+ current drives early afterdepolarizations in mouse ventricle

    Science.gov (United States)

    Edwards, Andrew G.; Grandi, Eleonora; Hake, Johan E.; Patel, Sonia; Li, Pan; Miyamoto, Shigeki; Omens, Jeffrey H.; Brown, Joan Heller; Bers, Donald M.; McCulloch, Andrew D.

    2015-01-01

    Background Early-afterdepolarizations (EADs) are triggers of cardiac arrhythmia driven by L-type Ca2+ current (ICaL) reactivation or sarcoplasmic reticulum (SR) Ca2+ release and Na+/Ca2+ exchange. In large mammals the positive action potential (AP) plateau promotes ICaL reactivation, and the current paradigm holds that cardiac EAD dynamics are dominated by interaction between ICaL and the repolarizing K+ currents. However, EADs are also frequent in the rapidly repolarizing mouse AP, which should not readily permit ICaL reactivation. This suggests that murine EADs exhibit unique dynamics, which are key for interpreting arrhythmia mechanisms in this ubiquitous model organism. We investigated these dynamics in myocytes from arrhythmia-susceptible CaMKIIδC-overexpressing mice (Tg), and via computational simulations. Methods and Results In Tg myocytes, β-adrenergic challenge slowed late repolarization, potentiated SR Ca2+ release, and initiated EADs below the ICaL activation range (−47±0.7 mV). These EADs were abolished by caffeine and tetrodotoxin (but not Ranolazine), suggesting that SR Ca2+ release and Na+ current (INa), but not late INa, are required for EAD initiation. Simulations suggest that potentiated SR Ca2+ release and Na+/Ca2+ exchange triangulate late AP repolarization, which permits non-equilibrium reactivation of INa, and thereby drives the EAD upstroke. AP clamp experiments suggest that lidocaine eliminates virtually all inward current elicited by EADs, and that this effect occurs at concentrations (40-60 μM) for which lidocaine remains specific for inactivated Na+ channels. This strongly suggests that previously inactive channels are recruited during the EAD upstroke, and that non-equilibrium INa dynamics underlie murine EADs. Conclusions Non-equilibrium reactivation of INa drives murine EADs. PMID:25236710

  6. Lower hybrid current drive at ITER-relevant high plasma densities

    International Nuclear Information System (INIS)

    Cesario, R.; Amicucci, L.; Cardinali, A.; Castaldo, C.; Marinucci, M.; Panaccione, L.; Pericoli-Ridolfini, V.; Tuccillo, A. A.; Tudisco, O.; Calabro, G.

    2009-01-01

    Recent experiments indicated that a further non-inductive current, besides bootstrap, should be necessary for developing advanced scenario for ITER. The lower hybrid current drive (LHCD) should provide such tool, but its effectiveness was still not proved in operations with ITER-relevant density of the plasma column periphery. Progress of the LH deposition modelling is presented, performed considering the wave physics of the edge, and different ITER-relevant edge parameters. Operations with relatively high edge electron temperatures are expected to reduce the LH || spectral broadening and, consequently, enabling the LH power to propagate also in high density plasmas ( || is the wavenumber component aligned to the confinement magnetic field). New results of FTU experiments are presented, performed by following the aforementioned modeling: they indicate that, for the first time, the LHCD conditions are established by operating at ITER-relevant high edge densities.

  7. EC + LH current drive efficiency in the presence of an internal transport barrier

    International Nuclear Information System (INIS)

    Rosa, P.R. da S; Ziebell, L.F.

    2002-01-01

    In this paper we study the effects of the presence of an internal transport barrier (ITB) on the current drive efficiency and power deposition profiles in the case of electron cyclotron waves interacting with an extended tail generated by lower hybrid (LH) waves. We study the subject by numerically solving the Fokker-Planck equation, with temperature and density profiles corrected along the time evolution at each collision time, based on the actual time-evolving electron distribution function. The results obtained show that the LH and electron cyclotron (EC) power absorption profiles and the current driven by the combined action of both types of waves are weakly dependent on the depth of the ITB, slightly more dependent on the level of magnetic turbulence and much more dependent on the level of EC wave power. (author)

  8. 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 ...

  9. 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 ...

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

    International Nuclear Information System (INIS)

    Jayakumar, R.J.

    2003-01-01

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

  11. Direct Measurement of the Electron Bernstein Wave Absorption and Current Drive at the WEGA Stellarator

    Czech Academy of Sciences Publication Activity Database

    Laqua, H.; Marsen, S.; Otte, M.; Podoba, Y.; Preinhaelter, Josef; Urban, Jakub

    2007-01-01

    Roč. 52, č. 16 (2007), s. 280-280 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/49th./. Orlando , Florida, 12.11.2007-16.11.2007] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Content/901

  12. Design constraints for rf-driven steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1979-02-01

    Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides

  13. Controlling fluctuations and transport in the reversed field pinch with edge current drive and plasma biasing

    International Nuclear Information System (INIS)

    Craig, D.J.G.

    1998-09-01

    Two techniques are employed in the Madison Symmetric Torus (MST) to test and control different aspects of fluctuation induced transport in the Reversed Field Pinch (RFP). Auxiliary edge currents are driven along the magnetic field to modify magnetic fluctuations, and the particle and energy transport associated with them. In addition, strong edge flows are produced by plasma biasing. Their effect on electrostatic fluctuations and the associated particle losses is studied. Both techniques are accomplished using miniature insertable plasma sources that are biased negatively to inject electrons. This type of emissive electrode is shown to reliably produce intense, directional current without significant contamination by impurities. The two most important conclusions derived from these studies are that the collective modes resonant at the reversal surface play a role in global plasma confinement, and that these modes can be controlled by modifying the parallel current profile outside of the reversal surface. This confirms predictions based on magnetohydrodynamic (MHD) simulations that auxiliary current drive in the sense to flatten the parallel current profile can be successful in controlling magnetic fluctuations in the RFP. However, these studies expand the group of magnetic modes believed to cause transport in MST and suggest that current profile control efforts need to address both the core resonant magnetic modes and those resonant at the reversal surface. The core resonant modes are not significantly altered in these experiments; however, the distribution and/or amplitude of the injected current is probably not optimal for affecting these modes. Plasma biasing generates strong edge flows with shear and particle confinement likely improves in these discharges. These experiments resemble biased H modes in other magnetic configurations in many ways. The similarities are likely due to the common role of electrostatic fluctuations in edge transport

  14. Role of the lower hybrid spectrum in the current drive modeling for DEMO scenarios

    Science.gov (United States)

    Cardinali, A.; Castaldo, C.; Cesario, R.; Santini, F.; Amicucci, L.; Ceccuzzi, S.; Galli, A.; Mirizzi, F.; Napoli, F.; Panaccione, L.; Schettini, G.; Tuccillo, A. A.

    2017-07-01

    The active control of the radial current density profile is one of the major issues of thermonuclear fusion energy research based on magnetic confinement. The lower hybrid current drive could in principle be used as an efficient tool. However, previous understanding considered the electron temperature envisaged in a reactor at the plasma periphery too large to allow penetration of the coupled radio frequency (RF) power due to strong Landau damping. In this work, we present new numerical results based on quasilinear theory, showing that the injection of power spectra with different {n}// widths of the main lobe produce an RF-driven current density profile spanning most of the outer radial half of the plasma ({n}// is the refractive index in a parallel direction to the confinement magnetic field). Plasma kinetic profiles envisaged for the DEMO reactor are used as references. We demonstrate the robustness of the modeling results concerning the key role of the spectral width in determining the lower hybrid-driven current density profile. Scans of plasma parameters are extensively carried out with the aim of excluding the possibility that any artefact of the utilised numerical modeling would produce any novelty. We neglect here the parasitic effect of spectral broadening produced by linear scattering due to plasma density fluctuations, which mainly occurs for low magnetic field devices. This effect will be analyzed in other work that completes the report on the present breakthrough.

  15. 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

  16. Response of a tokamak plasma to particle and momentum sources

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Sigmar, D.J.

    1978-12-01

    The response of an axisymmetric toroidal tokamak plasma to first-order particle and momentum sources is investigated. The momentum sources drive coupled poloidal and toroidal mass flows and electrostatic field evolution which relax to asymptotic values on a time scale that is characteristic of the dominant viscous or external drag mechanism. The asymptotic steady-state momentum balance provides the necessary condition to completely determine the particle fluxes and currents in the flux surfaces, and, hence, to determine transport fluxes across flux surfaces. Transport fluxes are driven across flux surfaces both by interspecies collisional momentum exchange, the usual case, and by momentum exchange between the plasma and external sources and/or drags. A generalized Ohm's law is obtained and used to determine the manner in which particle and momentum sources can drive parallel currents and can alter the evolution of the q-profile. The theory is formulated for arbitrary plasma cross sections, beta, and collision regimes

  17. Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry

    Energy Technology Data Exchange (ETDEWEB)

    Long-Poe Ku and Allen H. Boozer

    2009-06-05

    If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.

  18. 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

  19. Status of tokamak research

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, J.M. (ed.)

    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. (MOW)

  20. Compact ASD Topologies for Single-Phase Integrated Motor Drives with Sinusoidal Input Current

    DEFF Research Database (Denmark)

    Klumpner, Christian; Blaabjerg, Frede; Thoegersen, Paul

    2005-01-01

    A standard configuration of an Adjustable Speed Drive (ASD) consists of two separate units: an AC motor, which runs with fixed speed when it is supplied from a constant frequency grid voltage and a frequency converter, which is used to provide the motor with variable voltage-variable frequency......-density integration of the converter caused by the large size of the passive components (electrolytic capacitors and iron chokes) and vibration of the converter enclosure. This paper analyzes the implementation aspects for obtaining a compact and cost effective single-phase ASD with sinusoidal input current...... for high frequency operation, higher core losses will occur, but outside the converter enclosure. The advantages are: the reduction of the number of active semiconductor devices, the reduction of the ASD size and the better integration potential....

  1. RAMI Analysis of Neutral Beam Heating and Current Drive System for ITER

    International Nuclear Information System (INIS)

    Chang, Doo Hee; Lee, Sang Il

    2009-01-01

    A RAMI (Reliability, Availability, Maintainability, Inspectability) analysis has been performed for the neutral beam (NB) heating and current drive (H and CD) system of ITER (International Thermonuclear Experimental Reactor) device. The objective of this analysis is to implement RAMI engineering requirements for the design and testing to prepare a reliability-centred plan for commissioning, operation, and maintenance of the system in the framework of a technical risk control to support the overall ITER Project. These RAMI requirements will correspond to the RAMI targets for the ITER project and the compensating provisions to reach them as deduced from the necessary actions to decrease the risk level of the function failure modes. The RAMI analyses have to match with the procurement plan of the system

  2. HHFW Heating and Current Drive Studies of NSTX H-Mode Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    G. Taylor, P.T. Bonoli, D.L. Green, R.W. Harvey, J.C. Hosea, E.F. Jaeger, B.P. LeBlanc, R. Maingi, C.K. Phillips, P.M. Ryan, E.J. Valeo, J.R. Wilson, J.C. Wright, and the NSTX Team

    2011-06-08

    30 MHz high-harmonic fast wave (HHFW) heating and current drive are being developed to assist fully non-inductive plasma current (I{sub p}) ramp-up in NSTX. The initial approach to achieving this goal has been to heat I{sub p} = 300 kA inductive plasmas with current drive antenna phasing in order to generate an HHFW H-mode with significant bootstrap and RF-driven current. Recent experiments, using only 1.4 MW of RF power (P{sub RF}), achieved a noninductive current fraction, f{sub NI} {approx} 0.65. Improved antenna conditioning resulted in the generation of I{sub p} = 650 kA HHFW H-mode plasmas, with f{sub NI} {approx} 0.35, when P{sub RF} {ge} 2.5 MW. These plasmas have little or no edge localized mode (ELM) activity during HHFW heating, a substantial increase in stored energy and a sustained central electron temperature of 5-6 keV. Another focus of NSTX HHFW research is to heat an H-mode generated by 90 keV neutral beam injection (NBI). Improved HHFW coupling to NBI-generated H-modes has resulted in a broad increase in electron temperature profile when HHFW heating is applied. Analysis of a closely matched pair of NBI and HHFW+NBI H-mode plasmas revealed that about half of the antenna power is deposited inside the last closed flux surface (LCFS). Of the power damped inside the LCFS about two-thirds is absorbed directly by electrons and one-third accelerates fast-ions that are mostly promptly lost from the plasma. At longer toroidal launch wavelengths, HHFW+NBI H-mode plasmas can have an RF power flow to the divertor outside the LCFS that significantly reduces RF power deposition to the core. ELMs can also reduce RF power deposition to the core and increase power deposition to the edge. Recent full wave modeling of NSTX HHFW+NBI H-mode plasmas, with the model extended to the vessel wall, predicts a coaxial standing mode between the LCFS and the wall that can have large amplitudes at longer launch wavelengths. These simulation results qualitatively agree with HHFW

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

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

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

  4. Integrated tokamak modeling: when physics informs engineering and research planning

    Science.gov (United States)

    Poli, Francesca

    2017-10-01

    Simulations that integrate virtually all the relevant engineering and physics aspects of a real tokamak experiment are a power tool for experimental interpretation, model validation and planning for both present and future devices. This tutorial will guide through the building blocks of an ``integrated'' tokamak simulation, such as magnetic flux diffusion, thermal, momentum and particle transport, external heating and current drive sources, wall particle sources and sinks. Emphasis is given to the connection and interplay between external actuators and plasma response, between the slow time scales of the current diffusion and the fast time scales of transport, and how reduced and high-fidelity models can contribute to simulate a whole device. To illustrate the potential and limitations of integrated tokamak modeling for discharge prediction, a helium plasma scenario for the ITER pre-nuclear phase is taken as an example. This scenario presents challenges because it requires core-edge integration and advanced models for interaction between waves and fast-ions, which are subject to a limited experimental database for validation and guidance. Starting from a scenario obtained by re-scaling parameters from the demonstration inductive ``ITER baseline'', it is shown how self-consistent simulations that encompass both core and edge plasma regions, as well as high-fidelity heating and current drive source models are needed to set constraints on the density, magnetic field and heating scheme. This tutorial aims at demonstrating how integrated modeling, when used with adequate level of criticism, can not only support design of operational scenarios, but also help to asses the limitations and gaps in the available models, thus indicating where improved modeling tools are required and how present experiments can help their validation and inform research planning. Work supported by DOE under DE-AC02-09CH1146.

  5. Direct observation of current in type-I edge-localized-mode filaments on the ASDEX Upgrade tokamak.

    Science.gov (United States)

    Vianello, N; Naulin, V; Schrittwieser, R; Müller, H W; Zuin, M; Ionita, C; Rasmussen, J J; Mehlmann, F; Rohde, V; Cavazzana, R; Maraschek, M

    2011-03-25

    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 in intermediate mode number (n≅10-15) structures (filaments) localized in the perpendicular plane and extended along the field lines. By detailed localized measurements of the magnetic field perturbation associated to type-I ELM filaments, it is shown that these filaments carry a substantial current.

  6. A DEMO relevant fast wave current drive high harmonic antenna exploiting the high impedance technique

    Energy Technology Data Exchange (ETDEWEB)

    Milanesio, D., E-mail: daniele.milanesio@polito.it; Maggiora, R. [Politecnico di Torino, Dipartimento di Elettronica e Telecomunicazioni (DET), Torino (Italy)

    2015-12-10

    Ion Cyclotron (IC) antennas are routinely adopted in most of the existing nuclear fusion experiments, even though their main goal, i.e. to couple high power to the plasma (MW), is often limited by rather severe drawbacks due to high fields on the antenna itself and on the unmatched part of the feeding lines. In addition to the well exploited auxiliary ion heating during the start-up phase, some non-ohmic current drive (CD) at the IC range of frequencies may be explored in view of the DEMO reactor. In this work, we suggest and describe a compact high frequency DEMO relevant antenna, based on the high impedance surfaces concept. High-impedance surfaces are periodic metallic structures (patches) usually displaced on top of a dielectric substrate and grounded by means of vertical posts embedded inside the dielectric, in a mushroom-like shape. These structures present a high impedance, within a given frequency band, such that the image currents are in-phase with the currents of the antenna itself, thus determining a significant efficiency increase. After a general introduction on the properties of high impedance surfaces, we analyze, by means of numerical codes, a dielectric based and a full metal solution optimized to be tested and benchmarked on the FTU experiment fed with generators at 433MHz.

  7. 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)

  8. 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...

  9. 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.)

  10. 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

  11. Canonical profiles in tokamaks

    International Nuclear Information System (INIS)

    Dnestrovskij, Yu.N.

    2002-01-01

    We consider the problem of the canonical profiles for tokamak plasma with arbitrary cross-section, taking into account two principles: 1) the free plasma energy minimum with the constraint of total current conservation and 2) the profile consistency. We deduce the Euler differential equation for the canonical profile of μ=1/q with two types of the boundary conditions: soft and stiff. The soft conditions correspond to the Kadomtsev solution for the circular cylinder. The stiff conditions describe a fast response of the plasma over the whole cross-section on the edge impact. Using the canonical profile of the current density, we calculate the critical gradients for the temperature, and create the transport model for the electron and ion temperatures and density. We show that, when the aspect ratio is diminished, or when the elongation increases, the canonical profiles become flatten. The similar tendency for the real profiles of the electron temperature was found in analysis of JET and START experiments. The obtained critical gradients were used to analysis of the experiments in tokamaks with moderate and tight aspect ratios. (author)

  12. Modeling of Optimization and Control of EBW Heating and Current Drive

    Czech Academy of Sciences Publication Activity Database

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

    2009-01-01

    Roč. 54, č. 15 (2009), s. 100-100 ISSN 0003-0503. [Annual Meeting of the APS Division of Plasma Physics/51st./. Atlanta, 02.11.2009-06.11.2009] R&D Projects: GA MŠk 7G09042; GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Electron Bernstein wave * EBW * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/link/BAPS.2009.DPP.JO4.16

  13. Intense lower-hybrid wave penetration and current drive in reactor-grade plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, R.H.; Rognlien, T.D (Lawrence Livermore National Lab., CA (USA)); Bonoli, P.T.; Porkolab, M. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Plasma Fusion Center)

    1990-01-01

    Apply lower-hybrid power in short, intense pulses can overcome Landau damping, allowing penetration into the core of reactor-grade plasmas. We present a theoretical description of the absorption and parametric stability of the pulses, and show results of ray-tracing calculations which include the absorption calculation. Consideration of the absorption and potential source availability lead to the consideration of 5--10 GW peak power, 30--100 {mu}s pulses for ITER, and {approximately} 2 MW, 20 {mu}s pulses for a proof-of-principle experiment in the Microwave Tokamak Experiment (MTX).

  14. Investigation on synergy of IBW and LHCD for integrated high performance operation in HT-7 tokamak

    International Nuclear Information System (INIS)

    Wan Baonian

    2002-01-01

    Control of the current density profile has been realized with off-axis current drive by LHW in the HT-7 tokamak predicted by a 2D FP code simulation and supported by measurements of a vertical HX array. IBW is explored to improve performance through heating electrons in the selected region. Strong synergy effect on driven current profile and increased driven efficiency was observed. Electron temperature shows an ITB-like profile with a significantly improved performance. Operation of IBW and LHCD synergetic discharges was optimized through moving the IBW resonant layer to maximize the plasma performance and to avoid the MHD activities. A variety of high performance discharges indicated by β N *H89=1∼ 4 was produced for several tens energy confinement times. This operation mode utilizing synergy effect of IBW and LHCD provide a new way to obtain steady-state operation in advanced tokamak scenario. (author)

  15. 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

  16. Development of long-pulse heating & current drive actuators & operational techniques compatible with a high-Z divertor & first wall

    Energy Technology Data Exchange (ETDEWEB)

    Tynan, George [Univ. of California, San Diego, CA (United States)

    2018-01-09

    This was a collaboration between UCSD and MIT to study the effective application of ion-cyclotron heating (ICRH) on the EAST tokamak, located in China. The original goal was for UCSD to develop a diagnostic that would allow measurement of the steady state, or DC, convection pattern that develops on magnetic field lines that attach or connect to the ICRH antenna. This diagnostic would then be used to develop techniques and approaches that minimize or even eliminate such DC convection during application of strong ICRH heating. This was thought to then indicate reduction or elimination of parasitic losses of heating power, and thus be an indicator of effective RF heating. The original plan to use high speed digital gas-puff imaging (GPI) of the antenna-edge plasma region in EAST was ultimately unsuccessful due to limitations in machine and camera operations. We then decided to attempt the same experiment on the ALCATOR C-MOD tokamak at MIT which had a similar instrument already installed. This effort was ultimately successful, and demonstrated that the underlying idea of using GPI as a diagnostic for ICRH antenna physics would, in fact, work. The two-dimensional velocity fields of the turbulent structures, which are advected by RF-induced E x B flows, are obtained via the time-delay estimation (TDE) techniques. Both the magnitude and radial extension of the radial electric field E-r were observed to increase with the toroidal magnetic field strength B and the ICRF power. The TDE estimations of RF-induced plasma potentials are consistent with previous results based on the probe measurements of poloidal phase velocity. The results suggest that effective ICRH heating with reduced impurity production is possible when the antenna/box system is designed so as to reduce the RF-induced image currents that flow in the grounded conducting antenna frame elements that surround the RF antenna current straps.

  17. 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.)

  18. First lower hybrid current drive experiments at 3.7 GHz in Tore Supra

    International Nuclear Information System (INIS)

    Tonon, G.; Goniche, M.; Moreau, D.

    1989-01-01

    The results of electromagnetic waves injection in the Tore Supra plasma, at a frequency of 3.7 GHz, are reported. The process is applied for current generation and plasma heating, through Landau damping on the electron population. The experimental set-up is described. The lower hybrid current drive experiments in Tore Supra are carried out under the following conditions: major and minor radii of the plasma are respectively 2.37 m and 0.77 m and the toroidal magnetic field is 1.8 Teslas. A multijunction-grill composed of 128 waveguides is applied. Up to 1.25 MW of rf power is injected in Tore Supra, after less than 30 plasma shots. The results lead to the conclusion that the coupling, not yet optimized, is good enough for safe klystron operation with no circulator. The measured value RIp P RF -1 (δV L /V L ) obtained on Tore Supra (Bt = 1.8 T) is closed to one observed on PETULA-B (Bt = 2.75 T) at the same frequency and density

  19. Measurement of LHCD antenna position in Aditya tokamak

    International Nuclear Information System (INIS)

    Ambulkar, K K; Sharma, P K; Virani, C G; Parmar, P R; Thakur, A L; Kulkarni, S V

    2010-01-01

    To drive plasma current non-inductively in ADITYA tokamak, 120 kW pulsed Lower Hybrid Current Drive (LHCD) system at 3.7 GHz has been designed, fabricated and installed on ADITYA tokamak. In this system, the antenna consists of a grill structure, having two rows, each row comprising of four sub-waveguides. The coupling of LHCD power to the plasma strongly depends on the plasma density near the mouth of grill antenna. Thus the grill antenna has to be precisely positioned for efficient coupling. The movement of mechanical bellow, which contracts or expands up to 50mm, governs the movement of antenna. In order to monitor the position of the antenna precisely, the reference position of the antenna with respect to the machine/plasma position has to be accurately determined. Further a mechanical system or an electronic system to measure the relative movement of the antenna with respect to the reference position is also desired. Also due to poor accessibility inside the ADITYA machine, it is impossible to measure physically the reference position of the grill antenna with respect to machine wall, taken as reference position and hence an alternative method has to be adopted to establish these measurements reliably. In this paper we report the design and development of a mechanism, using which the antenna position measurements are made. It also describes a unique method employing which the measurements of the reference position of the antenna with respect to the inner edge of the tokamak wall is carried out, which otherwise was impossible due to poor accessibility and physical constraints. The position of the antenna is monitored using an electronic scale, which is developed and installed on the bellow. Once the reference position is derived, the linear potentiometer, attached to the bellow, measures the linear distance using position transmitter. The accuracy of measurement obtained in our setup is within +/- 0.5 % and the linearity, along with repeatability is excellent.

  20. Observation of Abrupt- and Fast-rising SOL Current during Trigger Phase of ELMs in DIII-D Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-06-27

    Extensive studies to date of edge localized modes (ELMs) have sought their origin inside the separatrix, i.e., MHD instability from steep gradients in the plasma edge, and examined their consequences outside the separatrix, i.e., transport of heat and particles in the scrape-off-layer (SOL) and divertors. Recent measurement by a high-speed scrape-off-layer current (SOLC) diagnostic may indicate that the ELM trigger process lies, in part, in the SOL. Thermoelectrically driven SOLC precedes, or co-evolves with, other parameters of the ELM process, and thus can potentially play a causal role: error field generated by non-axisymmetric SOLC, flowing in the immediate vicinity (approximately 1 cm) of the plasma edge, may contribute toward destabilizing MHD modes. The SOLC, observed concurrently with MHD activity, including ELMs, has been reported elsewhere.

  1. 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

  2. Design of test kits for the RF characterization of the PAM antenna of LHCD system for Aditya-upgrade Tokamak

    International Nuclear Information System (INIS)

    Jain, Yogesh M.; Sharma, P.K.; Parmar, P.R.; Ambulkar, K.K.

    2017-01-01

    The Lower Hybrid Current Drive (LHCD) system of the ADITYA-Upgrade tokamak will employ a Passive Active Multijunction (PAM) antenna to launch 250 kW of RF power at 3.7 GHz to drive plasma current non inductively in the tokamak. To evaluate the RF performance of the designed PAM antenna, it is characterized with the help of VNA measurements. The performance of the PAM antenna is mainly decided by the integrated performance of the entire antenna (with a differential phase shift of 270° and equal power distribution between each of the output waveguides) and the performance of mode converter, which transforms input TE 10 mode to TE 30 mode (with a mode purity of 98.5% at the output). This poster thus reports the design and analysis of these testing kits. Also, the test results of PAM antenna obtained by using these test kits would also be presented and discussed in this poster

  3. 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)

  4. 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

  5. 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)

  6. 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

  7. Proposed tokamak poloidal field system development program

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.

    1977-05-01

    A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.

  8. 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.

  9. Compact Torus Injection Experiments on the H.I.T. teststand and the JFT-2M tokamak

    Science.gov (United States)

    Fukumoto, Naoyuki; Fujiwara, Makoto; Kuramoto, Keiji; Ageishi, Masaya; Nagata, Masayoshi; Uyama, Tadao; Ogawa, Hiroaki; Kasai, Satoshi; Hasegawa, Kouichi; Shibata, Takatoshi

    1997-11-01

    A spheromak-type compact torus (CT) acceleration and injection experiment has been carried out using the Himeji Institute of Technology Compact Torus Injector (HIT-CTI). We investigate the possibility of refueling, density control, current drive, and edge electric field control of tokamak plasmas by means of CT injection. The HIT-CTI produces a CT with a speed of 200 km/s and a density of 1× 10^21m-3. We have constructed new electrodes and power supplies, and will install the HIT-CTI on the JFT-2M tokamak at JAERI in Autumn 1997. The outer electrode serves as a common ground for both the formation bank (144μF, 20kV) and the acceleration bank (92.4μF, 40kV). If the external toroidal field of the tokamak is applied across the CT acceleration region, the CT kinetic energy might decrease during penetration into the field lines joining the inner and outer electrode. This could result in the CT not being able to reach the core of the tokamak plasma. Determining the optimum position of the inner electrode is one of the near term goals of this research. We will present magnetic probe, He-Ne interferometer and fast framing camera data from experiments at H.I.T., where a CT was accelerated into a transverse field. We will also present initial results from the operation of the HIT-CTI on the JFT-2M tokamak.

  10. Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

    International Nuclear Information System (INIS)

    Sharma, P.K.; Kazarian, F.; Garibaldi, P.; Gassman, T.; Artaud, J.F.; Bae, Y.S.; Belo, J.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Goniche, M.; Guilhem, D.

    2011-01-01

    In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.

  11. Hard X-ray intensity reduction during lower hybrid current drive experiments

    International Nuclear Information System (INIS)

    Mlynar, J.; Stoeckel, J.; Magula, P.

    1993-01-01

    A strong hard X-ray intensity reduction during a standard LHCD at the CASTOR tokamak was studied. From discussion it followed that the magnetic fluctuations level decrease is likely to be responsible for this effect beside the loop voltage decrease. To verify this idea, the connection between the magnetic fluctuation level and the hard X-ray intensity was studied in a nonstandard LHCD regime with a zero loop voltage reduction. These measurements strongly supported the concept that magnetic fluctuations level substantially influences the runaway electrons cross-field transport. Though, more data and a good code for modelling the anomalous transport and hard X-rays production would be of high value. Similar measurements especially for higher RF power should be carried out soon. Besides, the reduction of hard X-rays was observed in the experiments with edge plasma polarization lately; therefore, the magnetic fluctuations level in these experiments should be studied soon. (author) 6 figs., 6 refs

  12. 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.

  13. Active load current sharing in fuel cell and battery fed DC motor drive for electric vehicle application

    International Nuclear Information System (INIS)

    Pany, Premananda; Singh, R.K.; Tripathi, R.K.

    2016-01-01

    Highlights: • Load current sharing in FC and battery fed dc drive. • Active current sharing control using LabVIEW. • Detail hardware implementation. • Controller performance is verified through MATLAB simulation and experimental results. - Abstract: In order to reduce the stress on fuel cell based hybrid source fed electric drive system the controller design is made through active current sharing (ACS) technique. The effectiveness of the proposed ACS technique is tested on a dc drive system fed from fuel cell and battery energy sources which enables both load current sharing and source power management. High efficiency and reliability of the hybrid system can be achieved by proper energy conversion and management of power to meet the load demand in terms of required voltage and current. To overcome the slow dynamics feature of FC, a battery bank of adequate power capacity has to be incorporated as FC voltage drops heavily during fast load demand. The controller allows fuel cell to operate in normal load region and draw the excess power from battery. In order to demonstrate the performance of the drive using ACS control strategy different modes of operation of the hybrid source with the static and dynamic behavior of the control system is verified through simulation and experimental results. This control scheme is implemented digitally in LabVIEW with PCI 6251 DAQ I/O interface card. The efficacy of the controller performance is demonstrated in system changing condition supplemented by experimental validation.

  14. 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.

  15. 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

  16. Prospects for stabilization of neoclassical tearing modes by electron cyclotron current drive in ITER

    International Nuclear Information System (INIS)

    La Haye, R.J.; Isayama, A.; Maraschek, M.

    2009-01-01

    The system planned for electron cyclotron current drive (ECCD) in ITER can mitigate the deleterious effects of neoclassical tearing modes (NTMs) provided that either adequate alignment of the ECCD to the rational surface is maintained or too large a misalignment is corrected on a time scale shorter than the deleterious plasma response to 'large' islands. Resistive neoclassical tearing modes will be the principal limit on stability and performance in the ITER standard scenario as the drag from rotating island induced eddy current in the resistive wall (particularly from the m/n = 2/1 mode) can slow the plasma rotation, produce locking to the wall and cause loss of high-confinement H-mode and disruption. Continuous wave (cw) ECCD at the island rational surface is successful in stabilization and/or prevention of NTMs in ASDEX Upgrade, DIII-D and JT-60U. Modulating the ECCD so that it is absorbed only on the rotating island O-point is proving successful in recovering effectiveness in ASDEX Upgrade when the ECCD is configured for wider deposition as expected in ITER. The models for the effect of misalignment on ECCD effectiveness are applied to ITER. Tolerances for misalignment are presented to establish criteria for both the alignment (by moving mirrors in ITER) in the presence of an island, and for the accuracy of real-time ITER MHD equilibrium reconstruction in the absence of an island, i.e. alignment to the mode or to the rational surface in the absence of the mode. The narrower ECCD with front steering makes precise alignment more necessary for the most effective stabilization even though the ECCD is still relatively broad, with current density deposition (full width half maximum) almost twice the marginal island width. This places strict requirements on ECCD alignment with the expected ECCD effectiveness dropping to zero for misalignments as small as 1.7 cm. The system response time for growing islands and slowing rotation without and with ECCD (at different

  17. 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)

  18. 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

  19. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    Science.gov (United States)

    Greenfield, C. M.; DIII-D Team

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available.

  20. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    International Nuclear Information System (INIS)

    Greenfield, C M

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available

  1. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  2. 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.)

  3. Concept study of the Steady State Tokamak Reactor (SSTR)

    International Nuclear Information System (INIS)

    1991-06-01

    The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

  4. Analysis and Suppression of Zero Sequence Circulating Current in Open Winding PMSM Drives With Common DC Bus

    OpenAIRE

    Zhan, H.; Zhu, Z.Q.; Odavic, M.

    2017-01-01

    In this paper, the zero sequence circulating current in open winding permanent magnet synchronous machine (OW-PMSM) drives with common dc bus is systematically analyzed for the first time. It is revealed that the zero sequence circulating current is affected by zero sequence back-electromotive force, cross coupling voltages in zero sequence from the machine side, pulse-width modulation induced zero sequence voltage, and inverter nonlinearity from the inverter side. Particularly, the influence...

  5. 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.

  6. Configuration and engineering design of the ARIES-RS tokamak power plant

    International Nuclear Information System (INIS)

    Tillack, M.S.; Malang, S.; Waganer, L.; Wang, X.R.; Sze, D.K.; El-Guebaly, L.; Wong, C.P.C.; Crowell, J.A.; Mau, T.K.; Bromberg, L.

    1997-01-01

    ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems. (orig.)

  7. Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

    Science.gov (United States)

    Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the energy, vacuum, gas fueling, and security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

  8. Suprathermal electron studies in the TCV tokamak: Design of a tomographic hard-x-ray spectrometer

    International Nuclear Information System (INIS)

    Gnesin, S.; Coda, S.; Decker, J.; Peysson, Y.

    2008-01-01

    Electron cyclotron resonance heating and electron cyclotron current drive, disruptive events, and sawtooth activity are all known to produce suprathermal electrons in fusion devices, motivating increasingly detailed studies of the generation and dynamics of this suprathermal population. Measurements have been performed in the past years in the tokamak a configuration variable (TCV) tokamak using a single pinhole hard-x-ray (HXR) camera and electron-cyclotron-emission radiometers, leading, in particular, to the identification of the crucial role of spatial transport in the physics of ECCD. The observation of a poloidal asymmetry in the emitted suprathermal bremsstrahlung radiation motivates the design of a proposed new tomographic HXR spectrometer reported in this paper. The design, which is based on a compact modified Soller collimator concept, is being aided by simulations of tomographic reconstruction. Quantitative criteria have been developed to optimize the design for the greatly variable shapes and positions of TCV plasmas.

  9. Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Hennen, B.A.; Westerhof, E.; De Baar, M.R.; Nuij, P.W.J.M.; Steinbuch, M.

    2012-01-01

    Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron cyclotron resonance heating and current drive (ECRH/ECCD) with a tearing mode and the stabilization of a mode at a specific width. In order to simulate these control tasks, the time evolution of a tearing mode subject to suppression by ECRH/ECCD and destabilization by a magnetic perturbation field is modelled using the generalized Rutherford equation. The model includes an equilibrium model and an ECRH/ECCD launcher model. The dynamics and static equilibria of this model are analysed. The model is linearized and based on the linearized model, linear feedback controllers are designed and simulated, demonstrating both alignment and width control of tearing modes in TEXTOR. (paper)

  10. When do we think it is Safe to Drive after Hand Surgery? – Current Practice and Legal Perspective

    LENUS (Irish Health Repository)

    Murphy, SF

    2016-11-01

    Patients recovering from hand surgery frequently ask when it is safe to drive and it is unclear where the responsibility lies; the surgeon, the patient or the insurance company. An eight-question survey looking at various aspects of clinical practice was circulated to consultant and trainee plastic and orthopaedic surgeons in Ireland and the UK. Of the 89 surgeons who replied, (53%) felt the decision when to drive was the patient’s compared with the insurance company (40%) and the surgeon (7%). 80% advised patients to contact their insurance company. 87% were unaware of current regulations or guidelines. National guidelines were vague and left the decision with the treating doctor. Similarly, major insurers advise patients to contact their doctor for advice. From a legal standpoint, the patient has a duty of care to other road users to be in full control of his vehicle prior to driving, regardless of any advice received.

  11. Gate length scaling trends of drive current enhancement in CMOSFETs with dual stress overlayers and embedded-SiGe

    International Nuclear Information System (INIS)

    Flachowsky, S.; Wei, A.; Herrmann, T.; Illgen, R.; Horstmann, M.; Richter, R.; Salz, H.; Klix, W.; Stenzel, R.

    2008-01-01

    Strain engineering in MOSFETs using tensile nitride overlayer (TOL) films, compressive nitride overlayer (COL) films, and embedded-SiGe (eSiGe) is studied by extensive device experiments and numerical simulations. The scaling behavior was analyzed by gate length reduction down to 40 nm and it was found that drive current strongly depends on the device dimensions. The reduction of drain-current enhancement for short-channel devices can be attributed to two competing factors: shorter gate length devices have increased longitudinal and vertical stress components which should result in improved drain-currents. However, there is a larger degradation from external resistance as the gate length decreases, due to a larger voltage dropped across the external resistance. Adding an eSiGe stressor reduces the external resistance in the p-MOSFET, to the extent that the drive current improvement from COL continues to increase even down the shortest gate length studied. This is due to the reduced resistivity of SiGe itself and the SiGe valence band offset relative to Si, leading to a smaller silicide-active contact resistance. It demonstrates the advantage of combining eSiGe and COL, not only for increased stress, but also for parasitic resistance reduction to enable better COL drive current benefit

  12. The Effects of Transcranial Direct Current Stimulation (tDCS on Psychomotor and Visual Perception Functions Related to Driving Skills

    Directory of Open Access Journals (Sweden)

    Alexander Brunnauer

    2018-01-01

    Full Text Available Objective: It could be demonstrated that anodal transcranial direct current stimulation (tDCS of the left dorsolateral prefrontal cortex (DLPFC enhances accuracy in working memory tasks and reaction time in healthy adults and thus may also have an influence on complex everyday tasks like driving a car. However, no studies have applied tDCS to psychomotor skills related to a standard driving test so far.Methods: 10 female and 5 male healthy adults without any medication and history of psychiatric or neurological illness were randomly assigned to two groups receiving active and sham stimulation in a double blind, cross-over study design. Standardized computerized psychomotor tests according to the German guidelines for road and traffic safety were administered at baseline. Then they performed the same tests during an anodal or sham tDCS of the left DLPFC in two separated sessions.Results: No significant improvements in skills related to driving performance like visual perception, stress tolerance, concentration, and vigilance could be shown after left anodal prefrontal tDCS. Side effects were low and did not differ between active and sham stimulation.Conclusions: The findings of our study indicate that left prefrontal tDCS may not alter driving skills affording more automated action patterns but as shown in previous studies may have an influence on driving behavior requiring executive control processes. This however has to be proved in future studies and within greater samples.

  13. Physics design of the HL-2A tokamak

    International Nuclear Information System (INIS)

    Gao Qingdi; Li Fangzhu; Zhang Jinhua; Pan Yudong; Jiao Yiming

    2005-10-01

    An overview report for the physics design of the HL-2A tokamak is presented. By numerically analyzing the plasma shaping and the vertical instability due to plasma elongation, the requirements for the currents of poloidal magnetic field coils and the control system are put forward. Controlling the plasma profile by using NBI (neutral beam injection) and LHCD (lower hybrid current drive) is investigated, and the high performance modes of operation in HL-2A and modeled and designed. The magnetohydrodynamic instabilities in improved confinement configuration (RS configuration) are studied so as to point out the way of plasma control to perform stationary high performance discharges in HL-2A. In order to offer data for updating the HL-2A divertor, performances of the divertor plasma are simulated. (authors)

  14. Compact ASD Topologies for Single-Phase Integrated Motor Drives with Sinusoidal Input Current

    DEFF Research Database (Denmark)

    Klumpner, Christian; Blaabjerg, Frede; Thoegersen, Paul

    2005-01-01

    A standard configuration of an Adjustable Speed Drive (ASD) consists of two separate units: an AC motor, which runs with fixed speed when it is supplied from a constant frequency grid voltage and a frequency converter, which is used to provide the motor with variable voltage-variable frequency ne...

  15. Input current interharmonics in adjustable speed drives caused by fixed-frequency modulation techniques

    DEFF Research Database (Denmark)

    Soltani, Hamid; Davari, Pooya; Loh, Poh Chiang

    2016-01-01

    Adjustable Speed Drives (ASDs) based on double-stage conversion systems may inject interharmonics distortion into the grid, other than the well-known characteristic harmonic components. The problems created by interharmonics make it necessary to find their precise sources, and, to adopt an approp...

  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. Optimization of steady-state beam-driven tokamak reactors

    International Nuclear Information System (INIS)

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

    1983-01-01

    Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed that includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures that minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in a reactor the size of the International Tokamak Reactor, then the optimum temperatures are typically T /SUB e/ approx. = 12 to 15 keV and T /SUB i/ approx. = 17 to 21 keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E /SUB b/ less than or equal to 400 keV, but rises only slowly above E /SUB b/ about 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors

  19. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    1982-09-01

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

  20. ICRF experiments and synergy with LHCD on HT-6M tokamak

    International Nuclear Information System (INIS)

    Li, J.; Yin, F.X.; Wan, B.N.

    1997-01-01

    The successful ion cyclotron heating (ICRH) experiment with high power density of nearly 1MW/m 3 was carried out in HT-6M tokamak. The good heating efficiency was achieved by using different wall conditioning techniques, such as He GDC, Ti gettering and boronization. With 300kW injected RF power, the ion temperature reach about 750eV and Te increases from 700eV to about 1keV. Synergy effects between lower hybrid current drive (LHCD) and ICRH have some unique features. The current driven efficiency improved in full current drive case from 0.8x10 19 AW -1 M -2 (without ICRH) to 1.75x10 19 AW -1 M -2 (with ICRH). The reason for this high current driven efficiency may because the mode conversion at ion-ion hybrid resonance to an Ion Bernstein Wave (IBW) which is damped on the fast electron. (author)

  1. Development of 4.6 GHz lower hybrid current drive system for steady state and high performance plasma in EAST

    Energy Technology Data Exchange (ETDEWEB)

    Liu, F.K.; Li, J.G.; Shan, J.F.; Wang, M.; Liu, L.; Zhao, L.M.; Hu, H.C.; Feng, J.Q.; Yang, Y.; Jia, H.; Wang, X.J.; Wu, Z.G.; Ma, W.D.; Huang, Y.Y.; Xu, H.D.; Zhang, J.; Cheng, M.; Xu, L.; Li, M.H.; Li, Y.C.; and others

    2016-12-15

    In order to achieve steady state and high performance plasma in EAST, a new lower hybrid current drive system at a frequency of 4.6 GHz has been built. The system is composed of 24 continuous wave (CW) klystron amplifiers to generate 6MW/CW microwave, 24 standard rectangle waveguide transmission lines with water-cooling plate, a multi-junction grill composed of 576 active (in groups of 8) and 84 passive sub-waveguides arranged in 12 rows and 6 columns, and four sets of high voltage power supplies. The power value and the spectrum of the launched microwave from the antenna can be controlled by the low-power microwave circuits in front of the klystrons. The new LHCD system has been applied to the experiments on EAST tokmak since 2014, and the obtained results suggest that it is effective to couple the wave into plasma and drive plasma current.

  2. Tokamak-like confinement at high beta and low field in the reversed field pinch

    International Nuclear Information System (INIS)

    Sarff, J S; Anderson, J K; Biewer, T M; Brower, D L; Chapman, B E; Chattopadhyay, P K; Craig, D; Deng, B; Hartog, D J Den; Ding, W X; Fiksel, G; Forest, C B; Goetz, J A; O'Connell, R; Prager, S C; Thomas, M A

    2003-01-01

    For several reasons, improved-confinement achieved in the reversed field pinch (RFP) during the last few years can be characterized as 'tokamak-like'. Historically, RFP plasmas have had relatively poor confinement due to tearing instability which causes magnetic stochasticity and enhanced transport. Tearing reduction is achieved through modification of the inductive current drive, which dramatically improves confinement. The electron temperature increases to >1 keV and the electron heat diffusivity decreases to approx. 5 m 2 s -1 , comparable with the transport level expected in a tokamak plasma of the same size and current. This corresponds to a 10-fold increase in global energy confinement. Runaway electrons are confined, and Fokker-Planck modelling of the electron distribution reveals that the diffusion at high energy is independent of the parallel velocity, uncharacteristic of stochastic transport. Improved-confinement occurs simultaneously with increased beta approx. 15%, while maintaining a magnetic field strength ten times weaker than a comparable tokamak. Measurements of the current, magnetic, and electric field profiles show that a simple Ohm's Law applies to this RFP sustained without dynamo relaxation

  3. The ARIES-ST study: Assessment of the spherical tokamak concept as fusion power plants

    International Nuclear Information System (INIS)

    Najmabadi, F.; Tillack, M.; Miller, R.; Mau, T.K.; Jardin, S.; Stambaugh, R.; Steiner, D.; Waganer, L.

    2001-01-01

    Recent experimental achievements and theoretical studies have generated substantial interest in the spherical tokamak concept. The ARIES-ST study was undertaken as a national U.S. effort to investigate the potential of the spherical tokamak concept as a fusion power plant and as a vehicle for fusion development. The 1000-MWe ARIES-ST power plant has an aspect ratio of 1.6, a major radius of 3.2 m, a plasma elongation (at 95% flux surface) of 3.4 and triangularity of 0.64. This configuration attains a β of 54% (which is 90% of the maximum theoretical β). While the plasma current is 31 MA, the almost perfect alignment of bootstrap and equilibrium current density profiles results in a current-drive power of only 31 MW. The on-axis toroidal field is 2.1 T and the peak field at the TF coil is 7.6 T, which leads to 288 MW of Joule losses in the normal-conducting TF system. The ARIES-ST study has highlighted many areas where tradeoffs among physics and engineering systems are critical in determining the optimum regime of operation for spherical tokamaks. Many critical issues also have been identified which must be resolved in R and D programs. (author)

  4. ORIENTED FOR CAD UNIVERSAL MATHEMATICAL MODEL OF THE EDDY CURRENT PROBE. PART 1. THE RESULTING FIELD OF THE DRIVE WINDING

    Directory of Open Access Journals (Sweden)

    A. G. Dedegkaev

    2015-01-01

    Full Text Available The work relates to the field of CAD methods and means of nondestructive electromagnetic testing of conductive flat products. The problem of constructing mathematical model of the drive winding eddy current probe linearly extended form is polygon in cross section. This allows you to create a universal mathematical model, which includes a model of a simpler shape in cross-section, as a special case. 

  5. 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.

  6. Dipole Map For Divertor Tokamaks

    International Nuclear Information System (INIS)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2003-01-01

    Heat flux impinging on the collector plates of divertor tokamaks can be prodigious. Therefore, the problem of spreading the heat flux on plates is a crucial issue for divertor tokamaks such as ITER. Here we use method of maps /1,2/ to investigate this problem. Magnetic field lines in non-axisymmetric divertor tokamaks are a one and a half degree of freedom Hamiltonian system /1-3/. We represent the unperturbed magnetic topology by the Symmetric Simple Map (SSM) /4/ given by yn+1 = yn + 2kxn - 2k2yn (1 - yn), xn+1 = xn - kyn (1 - yn) - 2k2yn+1 (1 - yn+1). The effects of a current carrying coil placed externally across from X-point is represented by Dipole Map (DP) /4,5/ given by x n+1 = x n + 2δs 3 x n+1 (y n - y s + s/[x n+1 2 + (y n - y s + s) 2 ] 2 ), y n+1 = y n + δs 3 x n+1 ((y n - y s + s) 2 - x n+1 2 /[x n+1 2 + (y n - y s + s) 2 ] 2 ) δ is amplitude of high MN magnetic perturbation, s is the distance of coil from last good surface across from X point, and is the y coordinate of last good surface where it crosses the axis joining X point and O point across from X point. We fix k=0.3 and s = (1/2)|y s |. We calculate the increase in width of stochastic layer and area of footprint of field lines on divertor plate as δ is increased. We also calculate how connection length, toroidal and poloidal circuits and their fractal structures, the number, location and density of hot spots change with δ. Finally, we make conclusions about how the heat flux can be possibly controlled and reduced by applying external magnetic perturbation in divertor tokamaks

  7. Overview of spherical tokamak research in Japan

    Science.gov (United States)

    Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.

    2017-10-01

    Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

  8. 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)

  9. New D3-D tokamak plasma control system

    Science.gov (United States)

    Campbell, G. L.; Ferron, J. R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C. M.; Pinsker, R. I.; Lazarus, E. A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and divertor power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the U.S. Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter.

  10. Overview of the STARFIRE reference commercial tokamak fusion power reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

    1980-01-01

    The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

  11. HIGH PERFORMANCE ADVANCED TOKAMAK REGIMES FOR NEXT-STEP EXPERIMENTS

    International Nuclear Information System (INIS)

    GREENFIELD, C.M.; MURAKAMI, M.; FERRON, J.R.; WADE, M.R.; LUCE, T.C.; PETTY, C.C.; MENARD, J.E; PETRIE, T.W.; ALLEN, S.L.; BURRELL, K.H.; CASPER, T.A; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M; GORELOV, Y.A; GROEBNER, R.J.; HOBIRK, J.; HYATT, A.W; JAYAKUMAR, R.J; KESSEL, C.E; LA HAYE, R.J; JACKSON, G.L; LOHR, J.; MAKOWSKI, M.A.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; TAYLOR, T.S; WEST, W.P.

    2003-01-01

    OAK-B135 Advanced Tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic (MHD) stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half-radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding (ELMing) H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts

  12. Wildcat: A commercial deuterium-deuterium tokamak reactor

    International Nuclear Information System (INIS)

    Evans, K.; Baker, C.C.; Barry, K.M.

    1983-01-01

    WILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design

  13. STARFIRE: a commercial tokamak fusion power plant study

    International Nuclear Information System (INIS)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations

  14. Neutral beam injection system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B.H.; Lee, K.W.; Chung, K.S.; Oh, B.H.; Cho, Y.S.; Bae, Y.D.; Han, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    The NBI system for KSTAR (Korean Superconducting Tokamak Advanced Research) has been designed based on conventional positive ion beam technology. One beam line consists of three ion sources, three neutralizers, one bending magnet, and one drift tube. This system will deliver 8 MW deuterium beam to KSTAR plasma in normal operation to support the advanced experiments on heating, current drive and profile control. The key technical issues in this design were high power ion source(120 kV, 65 A), long pulse operation (300 seconds; world record is 30 sec), and beam rotation from vertical to horizontal direction. The suggested important R and D points on ion source and beam line components are also included. (author). 7 refs., 27 figs., 1 tab.

  15. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  16. Long pulse FRC sustainment with enhanced edge driven rotating magnetic field current drive

    International Nuclear Information System (INIS)

    Hoffman, A.L.; Guo, H.Y.; Miller, K.E.; Milroy, R.D.

    2005-01-01

    FRCs have been formed and sustained for up to 50 normal flux decay times by Rotating Magnetic Fields (RMF) in the TCS experiment. For these longer pulse times a new phenomenon has been observed: switching to a higher performance mode delineated by shallower RMF penetration, higher ratios of generated poloidal to RMF drive field, and lower overall plasma resistivity. This global data is not explainable by previous RMF theory based on uniform electron rotational velocities or by numerical calculations based on uniform plasma resistivity, but agrees in many respects with new calculations made using strongly varying resistivity profiles. In order to more realistically model RMF driven FRCs with such non-uniform resistivity profiles, a double rigid rotor model has been developed with separate inner and outer electron rotational velocities and resistivities. The results of this modeling suggest that the RMF drive results in very high resistivity in a narrow edge layer, and that the higher performance mode is characterized by a sharp reduction in resistivity over the bulk of the FRC. (author)

  17. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

  18. A comparison of pulsed and steady-state tokamak reactor burn cycles. Pt. 2

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.; Kim, S.

    1985-01-01

    Pulsed operation of a tokamak reactor imposes cost penalties due to such problems as mechanical fatigue and the need to periodically transfer large amounts of energy to various reactor components. This study focuses on lifetime limitations and capital costs of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas include: fatigue in pulsed poloidal field coils; out-of-plane bending fatigue in toroidal field coils; electric power supply costs; and noninductive current driver costs. A capital cost comparison is made for tokamak reactors operating under the four distinct operating cycles which have been proposed. Since high availability and a low cost of energy will be mandatory for a commercial fusion reactor, we can characterize improvements in physics and technology which will help achieve these goals for different burn cycles. A key conclusion is that steady-state operation is likely to result in the least expensive tokamak reactor (perhaps 20% cheaper than the best pulsed reactor), provided noninductive current drive efficiency can be increased roughly four-fold over present-day experimental results. (orig.)

  19. Current drive drift waves as a possible mechanism for dynamo effect and transport in reversed field pinches

    International Nuclear Information System (INIS)

    Briguglio, S.; Romanelli, F.; Vlad, G.

    1986-01-01

    The possibility that a current driven drift wave turbulence may be responsible for the outward ion flux observed in Reversed Field Pinches (RFPs) is investigated; the latter flux was recently proposed as the driving mechanism of the dynamo sustaining the poloidal current in the external region of an RFP discharge. It is shown that this possibility can be supported by the linear theory of current driven drift waves. Finally, on the assumption that the transport is dominated by these instabilities, a scaling law for the temperature in RFPs is derived, which shows an approximately linear dependence on the current and a weak dependence on the size of the machine, in agreement with the experimental results. (author)

  20. Research using small tokamaks

    International Nuclear Information System (INIS)

    1989-07-01

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  1. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

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

  2. 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

  3. A mechanism for driving the gross Birkeland current configuration in the auroral oval

    International Nuclear Information System (INIS)

    Rostoker, G.; Bostrom, R.

    1976-01-01

    Birkeland (field-aligned) sheet currents flowing into and out of the auroral oval as reported by Zmuda and Armstrong (1974) are integrally associated with convective motion of plasma in the magnetotail. It is demonstrated that these currents can be driven by energy supplied by the braking of this convective motion of the plasma sheet particles as they drift toward the flanks of the magnetosphere. In the ionosphere the sheet currents close as Pedersen currents, resulting in the dissipation of power, while far from the earth the closure currents, which provide the braking force for the plasma, flow in the plasma sheet approximately normal to the neutral sheet out to radial distances of about 80 R/subE/. During periods of moderate magnetospheric activity the Birkeland currents result in a rate of dissipation of convective energy of the order of 10 GW

  4. Study on Rotor Current Waveforms in an Inverter-fed Induction Motor Drive During Overmodulation

    OpenAIRE

    Hari, Pavan Kumar VSS; Narayanan, G

    2011-01-01

    Overmodulation introduces low-order harmonics in the output voltage of a voltage source inverter. This paper presents the effects of low-order harmonics in the stator voltage on the rotor currents of an induction motor. Rotor current waveforms are presented for various operating zones in overmodulation, including six-step mode. Harmonic spectra of stator and rotor currents are compared in six-step mode of operation. Pulsating torque is evaluated at various depths of modulation during ov...

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. Lower hybrid current drive and heating experiments at the 1 MW rf power level on Alcator C

    International Nuclear Information System (INIS)

    Porkolab, M.; Lloyd, B.; Schuss, J.J.

    1983-07-01

    Lower hybrid current drive experiments were carried out in the density range 1.0 x 10 13 less than or equal to anti n(cm -3 ) less than or equal to 1.0 x 10 14 , at magnetic fields 6.0 less than or equal to B(T) less than or equal to 10. Using one 16 waveguide array, plasma currents of 150 to 200 kA have been driven by rf powers up to 600 kW for times greater than 100 msec at anti n/sub e/ up to 5 x 10 13 cm -3 . With two arrays at anti n/sub e/ approx. = 4.3 x 10 13 cm -3 at B/sub T/ = 10 T, plasma currents of 160 kA have been maintained by the rf power for 300 msec with zero loop voltage and constant internal inductance

  11. Current understanding of the driving mechanisms for spatiotemporal variations of atmospheric speciated mercury: a review

    Directory of Open Access Journals (Sweden)

    H. Mao

    2016-10-01

    Full Text Available Atmospheric mercury (Hg is a global pollutant and thought to be the main source of mercury in oceanic and remote terrestrial systems, where it becomes methylated and bioavailable; hence, atmospheric mercury pollution has global consequences for both human and ecosystem health. Understanding of spatial and temporal variations of atmospheric speciated mercury can advance our knowledge of mercury cycling in various environments. This review summarized spatiotemporal variations of total gaseous mercury or gaseous elemental mercury (TGM/GEM, gaseous oxidized mercury (GOM, and particulate-bound mercury (PBM in various environments including oceans, continents, high elevation, the free troposphere, and low to high latitudes. In the marine boundary layer (MBL, the oxidation of GEM was generally thought to drive the diurnal and seasonal variations of TGM/GEM and GOM in most oceanic regions, leading to lower GEM and higher GOM from noon to afternoon and higher GEM during winter and higher GOM during spring–summer. At continental sites, the driving mechanisms of TGM/GEM diurnal patterns included surface and local emissions, boundary layer dynamics, GEM oxidation, and for high-elevation sites mountain–valley winds, while oxidation of GEM and entrainment of free tropospheric air appeared to control the diurnal patterns of GOM. No pronounced diurnal variation was found for Tekran measured PBM at MBL and continental sites. Seasonal variations in TGM/GEM at continental sites were attributed to increased winter combustion and summertime surface emissions, and monsoons in Asia, while those in GOM were controlled by GEM oxidation, free tropospheric transport, anthropogenic emissions, and wet deposition. Increased PBM at continental sites during winter was primarily due to local/regional coal and wood combustion emissions. Long-term TGM measurements from the MBL and continental sites indicated an overall declining trend. Limited measurements suggested TGM

  12. Effect of resonant-to-bulk electron momentum transfer on the efficiency of electron-cyclotron current drive

    International Nuclear Information System (INIS)

    Matsuda, Y.; Smith, G.R.; Cohen, R.H.

    1988-01-01

    Efficiency of current drive by electron-cyclotron waves is investigated numerically by a bounce-average Fokker-Planck code to elucidate the effects of momentum transfer from resonant to bulk electrons, finite bulk temperature relative to the energy of resonant electrons, and trapped electrons. Comparisons are made with existing theories to assess their validity and quantitative difference between theory and code results. Difference of nearly a factor of 2 was found in efficiency between some theory and code results. 4 refs., 4 figs

  13. A model for the advantage of early electron cyclotron current drive in the suppression of neoclassical tearing modes

    International Nuclear Information System (INIS)

    Lazaros, Avrilios; Maraschek, Marc; Zohm, Hartmut

    2007-01-01

    An analytic model for the advantage of the early application of electron cyclotron current drive (ECCD) in the suppression of neoclassical tearing modes (NTMs) is presented. The improved performance of early ECCD is attributed to the second (smaller) saturation island width, which appears for sufficiently small (relative to the ECCD deposition width) critical island widths, in the strongly nonlinear growth rate profile. The operational range for the advantage of early ECCD is obtained, and it is shown that it is favored by broad deposition profiles. The preliminary experimental results in ASDEX Upgrade [H. Zohm et al., Nucl. Fusion 41, 197 (2001)] are consistent with the present model

  14. Behavior of impurity ion velocities during the pulsed poloidal current drive in the Madison symmetric torus reversed-field pinch

    International Nuclear Information System (INIS)

    Sakakita, Hajime; Craig, Darren; Anderson, Jay K.; Chapman, Brett E.; Den-Hartog, Daniel J.; Prager, Stewart C.; Biewer, Ted M.; Terry, Stephen D.

    2003-01-01

    We report on passive measurements of impurity ion velocities during the pulsed poloidal current drive (PPCD) in the Madison Symmetric Torus reversed-field pinch. During PPCD, the electron temperature increased and a sudden reduction of magnetic fluctuations was observed. For this change, we have studied whether plasma velocity is affected. Plasma rotation is observed to decrease during PPCD. From measurements of line intensities for several impurities at 10 poloidal chords, it is found that the impurity line emission shifts outward. The ion temperature of impurities is reasonably connected to that measured by charge exchange recombination spectroscopy from core to edge. (author)

  15. A Review of Voltage and Current Signature Diagnosis in Industrial Drives

    OpenAIRE

    K. Vinoth Kumar

    2011-01-01

    This paper presents the review of identify the different types of faults in the induction motor during online condition by using current and voltage signature analysis. Special attention is focused on the effect of both space distribution of rotor breakage and rotor dis-symmetry on the mechanism of generation of diagnosis signatures with the consideration of voltage supply unbalance and speed ripples. A comparison is made between the voltage signature analysis and current signature analysis. ...

  16. 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

  17. 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)

  18. Cyclotron radiation as Tokamak diagnostics

    International Nuclear Information System (INIS)

    Fiedler-Ferrari, N.

    1985-01-01

    A brief introduction to the use of Electron Cyclotron Emission as diagnostics in tokamaks is made. The utilization feasibility of this dignostics in the TBR-1 and TTF2A tokamaks is discussed. (L.C.) [pt

  19. The rationale driving the evolution of deep brain stimulation to constant-current devices.

    Science.gov (United States)

    Bronstein, Jeff M; Tagliati, Michele; McIntyre, Cameron; Chen, Robert; Cheung, Tyler; Hargreaves, Eric L; Israel, Zvi; Moffitt, Michael; Montgomery, Erwin B; Stypulkowski, Paul; Shils, Jay; Denison, Timothy; Vitek, Jerrold; Volkman, Jens; Wertheimer, Jeffrey; Okun, Michael S

    2015-02-01

    Deep brain stimulation (DBS) is an effective therapy for the treatment of a number of movement and neuropsychiatric disorders. The effectiveness of DBS is dependent on the density and location of stimulation in a given brain area. Adjustments are made to optimize clinical benefits and minimize side effects. Until recently, clinicians would adjust DBS settings using a voltage mode, where the delivered voltage remained constant. More recently, a constant-current mode has become available where the programmer sets the current and the stimulator automatically adjusts the voltage as impedance changes. We held an expert consensus meeting to evaluate the current state of the literature and field on constant-current mode versus voltage mode in clinical brain-related applications. There has been little reporting of the use of constant-current DBS devices in movement and neuropsychiatric disorders. However, as impedance varies considerably between patients and over time, it makes sense that all new devices will likely use constant current. © 2014 International Neuromodulation Society.

  20. 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