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Sample records for high field tokamaks

  1. Super high field ohmically heated tokamak operation

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Leclaire, R.J.; Potok, R.E.; Jassby, D.L.

    1986-01-01

    The authors discuss a super high field mode of tokamak operation that uses ohmic heating or near ohmic heating to ignition. The super high field mode of operation uses very high values of Β/sup 2/α, where Β is the magnetic field and a is the minor radius (Β/sup 2/α > 100 T/sup 2/m). We analyze copper magnet devices with major radii from 1.7 to 3.0 meters. Minimizing or eliminating the need for auxiliary heating has the potential advantages of reducing uncertainty in extrapolating the energy confinement time of current tokamak devices, and reducing engineering problems associated with large auxiliary heating requirements. It may be possible to heat relatively short pulse, inertially cooled tokamaks to ignition with ohmic power alone. However, there may be advantages in using a very small amount of auxiliary power (less than the ohmic heating power) to boost the ohmic heating and provide a faster start-up, expecially in relatively compact devices

  2. 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

  3. High-field, high-density tokamak power reactor

    International Nuclear Information System (INIS)

    Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

    1977-11-01

    A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

  4. Report on the high magnetic field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

    Ito, T; Kawai, Y; Toi, K; Hiraki, N; Nakamure, K [Kyushu Univ., Fukuoke (Japan). Research Inst. for Applied Mechanics

    1981-02-01

    A high magnetic field tokamak has been constructed at Kyushu University to study the confinement of high magnetic field tokamak plasma and turbulent heating. The tokamak device consists of toroidal field coils, vertical field coils, horizontal field coils, primary windings, a transformer iron core, turbulent heating coils, and a vacuum chamber. For the observation of plasma, plasma monitors, a micro-wave interferometer, a laser scattering system, a neutral particle energy analyzer, a soft X-ray detector, and a visible spectrometer were installed on the vacuum chamber. The experimental results showed that the central electron temperature was about 640 eV, the central ion temperature 280 eV and mean electron density 2.2 x 10/sup 14//cm/sup 3/. It was found that the proportionality law of electron density and confinement time was valid for this small plasma system. By the turbulent heating, the central ion temperature increased from 170 eV to 580 eV.

  5. High field Tokamaks: The why's and how's

    International Nuclear Information System (INIS)

    Sestero, A.

    1989-01-01

    After expressing the input power into the plasma through the energy confinement time τ E (by invoking thermal equilibrium), and in turn expressing τ E through the LAWSON parameter L, power-law type scalings can be manipulated into the form G(L,T,m)=M(α,B,A,q,β) (1) where G and M are monomial functions of their arguments. The function G can be labelled goal factor in that by assigning numerical values to its arguments L (Lawson's parameter), T (temperature) and m (ion mass), one is typically identifying specific goals on the way to fusion. Conversely, the function M can be labelled merit factor in that, whichever merit exists in the discharge parameters appearing as its arguments, it is expressed by the particular combination of values that M represents. Please note that, of the usual variables appearing in scaling laws, the minor radius α and the magnetic field B are maintained as such in relation (1), while the major radius R, the plasma current I and the plasma density n are respectively expressed in terms of the aspect ratio A, the safety factor q and the plasma β: consequently, arguments of M are one extensive variable (namely α) and one intensive variable (namely B), plus a set of three dimensionless variables (namely A,q,β). (author) 6 refs., 2 figs

  6. High gain requirements and high field Tokamak experiments

    International Nuclear Information System (INIS)

    Cohn, D.R.

    1994-01-01

    Operation at sufficiently high gain (ratio of fusion power to external heating power) is a fundamental requirement for tokamak power reactors. For typical reactor concepts, the gain is greater than 25. Self-heating from alpha particles in deuterium-tritium plasmas can greatly reduce ητ/temperature requirements for high gain. A range of high gain operating conditions is possible with different values of alpha-particle efficiency (fraction of alpha-particle power that actually heats the plasma) and with different ratios of self heating to external heating. At one extreme, there is ignited operation, where all of the required plasma heating is provided by alpha particles and the alpha-particle efficiency is 100%. At the other extreme, there is the case of no heating contribution from alpha particles. ητ/temperature requirements for high gain are determined as a function of alpha-particle heating efficiency. Possibilities for high gain experiments in deuterium-tritium, deuterium, and hydrogen plasmas are discussed

  7. Development of high field superconducting Tokamak 'TRIAM-1M'

    International Nuclear Information System (INIS)

    Ito, Satoshi; Suzuki, Takao; Suzuki, Shohei; Nishi, Masatsugu; Kawasaki, Takahide.

    1984-01-01

    The tokamak nuclear fusion apparatus ''TRIAM-1M'' which is constructed in the Research Institute for Applied Mechanics, Kyushu University, has a number of distinctive features as compared with other tokamak projects, that is, the toroidal field coils are made of superconductors for the first time in Japan, and the apparatus is small and has strong magnetic field. Hitachi Ltd. designed and has forwarded the manufacture of the TRIAM-1M. In this paper, the total constitution of the apparatus and the design and manufacture of the plasma vacuum vessel, superconducting toroidal coils and others are reported. The objectives of research are the containment of strong field tokamak plasma and the establishment of the law of proportion, the development of turbulent flow heating method, the adoption of mixed wave current driving method and the practical use of Nb 3 Sn superconducting coils. The apparatus is composed of the vacuum vessel containing plasma, toroidal field coils, poloidal field coils, current transformer coils and turbulent flow heating coils for plasma heating, heat insulating vacuum vessel and supporting structures. The evacuating facility, helium liquefying refrigerator and cooling water facility are installed around the main body. (Kako, I.)

  8. 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

  9. DEALS magnet concept and its applcations to high density, high field tokamak systems

    International Nuclear Information System (INIS)

    Hsieh, S.Y.; Powell, J.; Lehner, J.; Bezler, P.; Laverick, C.; Finkelman, M.; Brown, T.; Bundy, J.

    1977-01-01

    The goal of the DEALS program is to develop a demountable TF magnet system concept that will reduce construction and life cycle costs, enhance the accessibility of components inside the coil system, and increase the chances for being able to use large high-field magnet systems in post TFTR reactor experiments. These experiments are projected to occur during the mid 1980's, with conceptual designs beginning in two or three years. A number of recent studies have highlighted the need for Tokamak fusion reactor systems with reasonable down time for maintenance and repair and realistic operating capacity factors, as well as the need for smaller, lower cost reactors. Two scoping studies were carried out of recent Tokamak system concepts incorporating conventionally wound coils to assess the possibilities of using demountable coils of rectangular section with an active support system and a third more intensive study using a passive support with slight movement of the joints. These studies are described briefly

  10. A commercial tokamak reactor using super high field superconducting magnets

    International Nuclear Information System (INIS)

    Schwartz, J.; Bromberg, L.; Cohn, D.R.; Williams, J.E.C.

    1988-01-01

    This paper explores the range of possibilities for producing super high fields with advanced superconducting magnets. Obtaining magnetic fields greater than about 18 T at the coil in a large superconducting magnet system will require advances in many areas of magnet technology. These needs are discussed and potential solutions (advanced superconductors, structural materials and design methods) evaluated. A point design for a commercial reactor with magnetic field at the coil of 24 T and fusion power of 1800 MW is presented. Critical issues and parameters for magnet design are identified. 20 refs., 9 figs., 4 tabs

  11. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  12. Updated tokamak systems code and applications to high-field ignition devices

    International Nuclear Information System (INIS)

    Reid, R.L.; Galambos, J.D.; Peng, Y-K.M.; Strickler, D.J.; Selcow, E.C.

    1985-01-01

    This paper describes revisions made to the Tokamak Systems Code to more accurately model high-field copper ignition devices. The major areas of revision were in the plasma physics model, the toroidal field (TF) coil model, and the poloidal field (PF) coil/MHD model. Also included in this paper are results obtained from applying the revised code to a study for a high-field copper ignition device to determine the impact of magnetic field on axis, (at the major radius), on performance, and on cost

  13. High current superconductors for tokamak toroidal field coils

    International Nuclear Information System (INIS)

    Fietz, W.A.

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

  15. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

  16. Data processing system with a micro-computer for high magnetic field tokamak, TRIAM-1

    International Nuclear Information System (INIS)

    Kawasaki, Shoji; Nakamura, Kazuo; Nakamura, Yukio; Hiraki, Naoharu; Toi, Kazuo

    1981-01-01

    A data processing system was designed and constructed for the purpose of analyzing the data of the high magnetic field tokamak TRIAM-1. The system consists of a 10-channel A-D converter, a 20 K byte memory (RAM), an address bus control circuit, a data bus control circuit, a timing pulse and control signal generator, a D-A converter, a micro-computer, and a power source. The memory can be used as a CPU memory except at the time of sampling and data output. The out-put devices of the system are an X-Y recorder and an oscilloscope. The computer is composed of a CPU, a memory and an I/O part. The memory size can be extended. A cassette tape recorder is provided to keep the programs of the computer. An interface circuit between the computer and the tape recorder was designed and constructed. An electric discharge printer as an I/O device can be connected. From TRIAM-1, the signals of magnetic probes, plasma current, vertical field coil current, and one-turn loop voltage are fed into the processing system. The plasma displacement calculated from these signals is shown by one of I/O devices. The results of test run showed good performance. (Kato, T.)

  17. Data processing system with a micro-computer for high magnetic field tokamak, TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

    A data processing system was designed and constructed for the purpose of analyzing the data of the high magnetic field tokamak TRIAM-1. The system consists of a 10-channel A-D converter, a 20 K byte memory (RAM), an address bus control circuit, a data bus control circuit, a timing pulse and control signal generator, a D-A converter, a micro-computer, and a power source. The memory can be used as a CPU memory except at the time of sampling and data output. The out-put devices of the system are an X-Y recorder and an oscilloscope. The computer is composed of a CPU, a memory and an I/O part. The memory size can be extended. A cassette tape recorder is provided to keep the programs of the computer. An interface circuit between the computer and the tape recorder was designed and constructed. An electric discharge printer as an I/O device can be connected. From TRIAM-1, the signals of magnetic probes, plasma current, vertical field coil current, and one-turn loop voltage are fed into the processing system. The plasma displacement calculated from these signals is shown by one of I/O devices. The results of test run showed good performance.

  18. Ultra-long pulse operation using lower hybrid waves on the superconducting high field tokamak TRIAM-1M

    International Nuclear Information System (INIS)

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

    1990-01-01

    Ultra-long pulse operation (>3 min) was achieved on the superconducting high field tokamak TRIAM-1M. In this operation, the plasma current was maintained with a relatively peaked current distribution by the 2.45 GHz radiofrequency power (P RF ≤ 35 kW) alone. A stationary plasma with a driven current of up to 35 kA and a line averaged electron density of up to 3x10 12 cm -3 was produced by precise plasma position and gas feed control. The extremely long discharge showed the interesting characteristics that the high temperatures of about 1 keV for the electrons and about 0.5 keV for the ions were kept almost constant during steady state current drive and that there was no impurity accumulation which could have a fatally adverse effect on steady state tokamak operation. (author). 16 refs, 17 figs

  19. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1978-01-01

    A new type of superconducting magnet system for large fusion reactors is described. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil. (author)

  20. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1977-01-01

    A new type of superconducting magnet system for large fusion reactors is described in this report. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil

  1. The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Miller, R.L.

    1989-01-01

    The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

  2. Spherical tokamak without external toroidal fields

    International Nuclear Information System (INIS)

    Kaw, P.K.; Avinash, K.; Srinivasan, R.

    2001-01-01

    A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

  3. Stability of highly shifted equilibria in a large aspect ratio low-field tokamak

    International Nuclear Information System (INIS)

    Gourdain, P.-A.; Leboeuf, J.-N.; Neches, R. Y.

    2007-01-01

    In the long run, the economics of fusion will dictate that reactors confine large plasma pressure rather efficiently. A possible route manifests itself as equilibria with large shift of the plasma magnetic axis. This shift compresses the flux surfaces on the outer part of the plasma, hereby increasing the allowable plasma pressure a machine can confine for a given toroidal magnetic field, which is the main cost of the device. As a first step toward a reactor, we propose investigating the stability of such configurations in a low magnetic field high aspect ratio machine. By focusing our arguments solely on the shape of the toroidal plasma current density profile we discuss the stability of highly shifted equilibria and their robustness to current profile variations that could occur in actual experiments. The evolution of the plasma parameters, as the beta poloidal is increased, is also examined to give a better understanding of the difference in performance between the various regimes

  4. 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

  5. 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

  6. Development of high-mechanical strength electrical insulations for tokamak toroidal field coils

    International Nuclear Information System (INIS)

    Burke, C.

    1977-01-01

    The electrical insulation for the TF (Toroidal Field) coils is subjected to a high interlaminar shear, tensile and compressive stresses. Two candidate epoxy/glass fiber systems using prepreg and vacuum impregnation techniques were evaluated. Specimens were prepared and processed under controlled conditions to simulate specification manufacturing procedures. The strengths of the insulation were measured in interlaminar shear, tension, compression, and combined shear and compression statically. Shear modulus determinations were also made. Various techniques of surface treatments to increase bond strengths with three resin primers were tested

  7. Ion temperature measurement by neutral energy analyzer in high-field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-02-01

    The measurement of the ion temperature of the TRIAM-1 tokamak plasma is carried out by using a seven-channel neutral energy analyzer. The temporal and spatial variations of the ion temperature have been obtained with the spatial resolution of +-4.3 mm and the temporal resolution of 100 ..mu..sec. The energy range of the analyzed neutral particles is from 0.2 to 8 keV. The energy spectrum in the TRIAM-1 plasma without the strong gas puffing usually consists of two-component Maxwellian; the one represents the thermal part which is a superposition of the contribution from a hot region (T sub(i) = 100 - 300 eV) and that from an edge region (T sub(i) asymptotically equals 50 eV), and the other represents the superthermal part (T sub(i) asymptotically equals 1 keV). The neutral particle energy spectra at several vertical positions are obtained by scanning the analyzer in the vertical direction. From those spectra, the radial profile of the ion temperature is derived by means of the nonlinear optimization method.

  8. Wall conditioning with a high magnetic field in HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Li Jiangang; Gu Xuemao; Gao Xiang; Zhang Souying; Jie Yingxian; Yang Xiaokang

    2000-01-01

    ICRF wall conditioning techniques, which includes the hydrogen removal, impurity cleaning, boronization and siliconization, were described in this paper. This new technique has been demonstrated to be very effective for wall conditioning, recycling, isotopic control and used daily during experiments. The RF plasma parameters were measured as T e =3-8 eV, T i =0.5-2 keV, n e =0.3-5 x 10 17 m -3 by different diagnostics. The nontoxic and nonexplosive solid carborane powder was used for the RF boronization. Energetic ions cracked the carborane molecule and the boron ions impacted and deposited onto first wall. Comparing with GDC boronization, the B/C coating film shows the higher adhesion, better uniformity and longer lifetime to the plasma discharges. Siliconization was carried out by using a high field side long RF antenna, which made the discharge more uniform. The ratio of SiH 4 to helium is about 5:95 at the pressure range of P v =0.8-8 x 10 -2 Pa. Compare with boronization, it showed quicker recovery from a bad wall condition due to leakage of air to good wall condition. Plasma density could be easily controlled after siliconization. But the lifetime is much shorter than that obtained by boronization. Plasma performance has been improved after RF boronization and siliconization. (author)

  9. 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.

  10. Mercier criterion for hightokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

    An expression, for the application of the Mercier criterion to numerical studies of diffuse hightokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the hightokamak approximation is derived. (L.C.) [pt

  11. Resonant helical fields in the TBR tokamak

    International Nuclear Information System (INIS)

    Bender, O.W.

    1986-01-01

    The influence of external resonant helical fields (RHF) in the tokamak TBR plasma discharges was investigated. These fields were created by helical windings wounded on the TBR vessel with the same helicity of rational magnetic surfaces, producing resonant efects on these surfaces. The characteristics of the MHZ activity (amplitude, frequency and poloidal and toroidal wave numbers, m=2,3,4 and n=1, respectively) during the plasma discharges were modified by eletrical winding currents of the order of 2% of the plasma current. These characterisitics were measured for diferent discharges safety factors at the limiter (q) between 3 and 4, with and without the RHF, with the atenuation of the oscillation amplitudes and the increasing of their frequencies. The existente of expontaneous and induced magnetic islands were investigated. The data were compared with results obtained in other tokamaks. (author) [pt

  12. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof

  13. Accessibility of high β tokamak states

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1978-05-01

    Encouraging results with neutral beam heating and adiabatic compression of tokamak plasmas have prompted new experiments which will study the approach to high β states. As projected tokamak β values become nonnegligible (average β of 4% is the goal), the models previously used for transport calculations will become inadequate. These models will be required to account for the evolution of the magnetic geometry, along with the change in plasma parameters. We present an axisymmetric transport model which should be useful for studying the approach to higher β values in tokamak experiments. Results from transport calculations with this model allow us to draw a parallel between observed behavior in seemingly unrelated experiments: electron heating by neutral injection in the ORMAK device and adiabatic compression in the ATC experiment. Finally, we find that the nature of cross-field transport may be expected to change as significant β values are reached. Enhanced transport from ballooning instabilities is likely to play a role as important as that now played by sawtooth (m = 1) and saturated (m = 2) instabilities. New techniques for describing this transport are required

  14. Resonant helical fields in tokamaks

    International Nuclear Information System (INIS)

    Okano, V.

    1990-01-01

    Poincare maps of magnetic field lines of a toroidal helical system were made. The magnetic field is a linear superposition of the magnetic fields produced by a toroidal plasma in equilibrium and by external helical currents. Analytical expression for the Poincare maps was no obtained since the magnetic field do not have symmetry. In order to obtain the maps, the equation minus derivative of l vector times B vector = 0 was numerically integrated. In the Poincare maps, the principal and the secondary magnetic island were observed. (author)

  15. The theory of the quasi-optical grill: A lower hybrid wave launcher in the 4 - 10 GHz range for high field tokamaks

    International Nuclear Information System (INIS)

    Preinhaelter, J.; Vahala, L.; Vahala, G.

    1996-01-01

    Lower hybrid (LH) waves have been utilized for plasma heating and current drive in tokamaks. LH current drive has good efficiency in low to moderate plasma temperatures and is an excellent tool for attaining the reversed shear regions of much interest in advanced steady state tokamak scenarios. For high field tokamaks, the waveguides of the standard multifunction grills would become very narrow and the walls separating the waveguides would need to be very thin. As a result, the cooling of such structures becomes very difficult. Moreover, there are concerns that the classical grill launcher could not withstand the conditions at the reactor first wall. The Quasi-Optical Grill (QOG) was first proposed by Petelin ampersand Suvorov to overcome some of these difficulties. QOG attempts to couple the RF power to the plasma slow wave by means of the diffraction of the incident wave on an array of rods. However, these original calculations are based on certain idealized assumptions and lead to poor coupling to the plasma. Preinhaelter has suggested a new QOG in which the rods are placed in one oversized waveguide (open-quotes hyperguideclose quotes) and irradiated obliquely by the wave emerging as a higher order mode from an auxiliary oversized waveguide. The confining walls are now an intrinsic part of the structure and thus one avoids the need for mirrors and the introduction of open-quote point-like close-quote structures. This new QOG is compact - with several orders of magnitude less construction elements than the classical LH launcher - and the problem of wave diffraction can be readily solved using the full wave method. Here we consider the optimization of a large scale QOG at a given frequency. The irradiation of either a single row or double set of rows of rods are considered as well as their optimal separation. One can achieve transmissivity and directivity comparable to those of the multifunction grill. Design of a QOG for TORE-SUPRA will also be discussed

  16. Equilibrium vertical field in the TBR Tokamak

    International Nuclear Information System (INIS)

    Ueta, A.Y.

    1985-01-01

    An experimental study on the influence of the vertical magnetic field of the TBR tokamak on the stability and equilibrium of plasma column, was done. Magnetic pick-up coils were built to measure plasma current and position, together with active networks, necessary fo the electronic processing of signals. Some measurements were on the space configuration of the vertical field, and on the influence due to the toroidal vessel. From the data obtained it was possible to discuss the influence of the currents induced on the vessel surface, on plasma equilibrium. Theoretical and experimental results of the vertica field, as a function of plasma current were compared, and allowed an evaluation of the plasma kinetic pressure and temperature. (Author) [pt

  17. Impact of magnetic perturbation fields on tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  18. Mode Conversion of High-Field-Side-Launched Fast Waves at the Second Harmonic of Minority Hydrogen in Advanced Tokamak Reactors

    International Nuclear Information System (INIS)

    Sund, R.; Scharer, J.

    2003-01-01

    Under advanced tokamak reactor conditions, the Ion-Bernstein wave (IBW) can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side on the second harmonic resonance of a minority hydrogen component, with near 100% efficiency. IBWs have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. The relatively high frequency (around 200 MHz) minimizes parasitic electron absorption and permits the converted IBW to approach the 5th tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. The scheme is applicable to reactors with aspect ratios < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Large machine size and adequate separation of the mode conversion layer from the magnetic axis minimize poloidal field effects in the conversion zone and permit a 1-D full-wave analysis. 2-D ray tracing of the IBW indicates a slightly bean-shaped equilibrium allows access to the tritium resonance

  19. Ballooning stable high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

    1981-04-01

    The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

  20. Diffusion of chaotic field lines in tokamaks

    Science.gov (United States)

    Ali, Halima; Punjabi, Alkesh

    2006-10-01

    An important instability for the destruction of magnetic surfaces in tokamaks due to island overlapping is the tearing modes. Magnetic fields perturbed by tearing modes are given by the sinusoidal form Br=-1rR∑m,nbm^n ( mθ-n ) . The sinusoidal nature of perturbation creates islands structure near resonant surfaces. In this work, we consider two modes, ( m1,n1 )and ( m2,n2 )that interact with each other, leading to two chains of islands, called primary islands. We use a previously derived Hamiltonian map, the ψ-θ map, with and without higher order control terms to study the diffusion of chaotic field lines. We will present and discuss the results of this work, and discuss its implications with regard to magnetic transport barriers for a fixed q-profile and increasing strength of magnetic perturbations. This work is done under the DOE grant number DE-FG02-01ER54624. 1.A. Punjabi et al, Phys. Rev. lett., 69, 3322 (1992). 2. H. Ali, A. Punjabi, and A. Boozer, Int. J. Comp. Num. Ana. Applications 6, 17 (2005).

  1. Power supply for coil of poloidal field in a tokamak

    International Nuclear Information System (INIS)

    Kirpichev, I.V.; Spevakova, F.M.

    1984-01-01

    The invention refers to power supply systems of poloidal field winding in tokamaks. The purpose of the invention is the extension of functional capabilities of the circuit by means of use of thyristor and diode keys mounted between convertor sections

  2. System design of toroidal field power supply of CDD tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zheng Zhi

    1996-12-01

    This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs.

  3. System design of toroidal field power supply of CDD tokamak

    International Nuclear Information System (INIS)

    Liu, Zheng Zhi.

    1996-12-01

    This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs

  4. [High beta tokamak research and plasma theory

    International Nuclear Information System (INIS)

    1990-01-01

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

  5. Energy confinement of high-density tokamaks

    NARCIS (Netherlands)

    Schüller, F.C.; Schram, D.C.; Coppi, B.; Sadowski, W.

    1977-01-01

    Neoclassical ion heat conduction is the major energy loss mechanism in the center of an ohmically heated high-d. tokamak discharge (n>3 * 1020 m-3). This fixes the mutual dependence of plasma quantities on the axis and leads to scaling laws for the poloidal b and energy confinement time, given the

  6. HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet

    International Nuclear Information System (INIS)

    Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.

    1985-01-01

    The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1

  7. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  8. Propagation Dynamics Associated with Resonant Magnetic Perturbation Fields in High-Confinement Mode Plasmas inside the KSTAR Tokamak.

    Science.gov (United States)

    Xiao, W W; Evans, T E; Tynan, G R; Yoon, S W; Jeon, Y M; Ko, W H; Nam, Y U; Oh, Y K

    2017-11-17

    The propagation dynamics of resonant magnetic perturbation fields in KSTAR H-mode plasmas with injection of small edge perturbations produced by a supersonic molecular beam injection is reported for the first time. The results show that the perturbation field first excites a plasma response on the q=3 magnetic surface and then propagates inward to the q=2 surface with a radially averaged propagation velocity of resonant magnetic perturbations field equal to 32.5  m/ s. As a result, the perturbation field brakes the toroidal rotation on the q=3 surface first causing a momentum transport perturbation that propagates both inward and outward. A higher density fluctuation level is observed. The propagation velocity of the resonant magnetic perturbations field is larger than the radial propagation velocity of the perturbation in the toroidal rotation.

  9. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Wang Zhongtian; Le Clair, G.

    1996-07-01

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

  11. Fusion-product ash buildup in tokamak with radial electric field

    International Nuclear Information System (INIS)

    Downum, W.B.; Choi, C.K.; Miley, G.H.

    1979-01-01

    The buildup of thermalized fusion products (ash) in a tokamak can seriously limit burn times. Prior studies have concentrated on deposition profile effects on alpha particle transport in tokamaks but have not considered the effect on ash of radial electric fields (either created internally, e.g. due to high-energy alpha leakage, or generated externally). The present study focuses on this issue since it appears that electric fields might offer one approach to control of the ash. Approximate field and source profiles are used, based on prior calculations

  12. Tokamak poloidal-field systems. Progress report, January 1-December 31, 1982

    International Nuclear Information System (INIS)

    Rogers, J.D.

    1983-05-01

    The work performed in support of the FED and INTOR tokamak studies is reported at length and covers almost all the aspects of poloidal field (PF) design that were considered. The design work included magnetics, forces and fields, superconductor design, superconductor loss calculations, high field tokamak central solenoid parametric analysis, helium vapor release with bubble clearing and entrainment analysis, eddy current losses in dewars, structural support design for internally cooled cable superconductor (ICCS), research and technology development and manufacturing plans and milestones for poloidal field (PF) coils, fault conditions for shorted PF coils, design of 50 kA vapor cooled leads, and structural design of PF ring coils box frame dewars. Eddy current calculations in tokamak structure are being calculated. A computer code to perform stability analysis of ICCS is being written. Two water cooled switches, a vacuum interrupter and a bypass switch, were tested to develop improved higher current carrying capacities

  13. Commercial tokamak reactors with resistive toroidal field magnets

    International Nuclear Information System (INIS)

    Bombery, L.; Cohn, D.R.; Jassby, D.L.

    1984-01-01

    Scaling relations and design concepts are developed for commercial tokamak reactors that use watercooled copper toroidal field (TF) magnets. Illustrative parameters are developed for reactors that are scaled up in size from LITE test reactor designs, which use quasi-continuous copper plate magnets. Acceptably low magnet power requirements may be attainable in a moderate beta (β = 0.065) commercial reactor with a major radius of 6.2 m. The shielding thickness and magnet size are substantially reduced relative to values in commercial reactors with superconducting magnets. Operation at high beta (β = 0.14) leads to a reduction in reactor size, magnet-stored energy, and recirculating power. Reactors using resistive TF magnets could provide advantages of physically smaller devices, improved maintenance features, and increased ruggedness and reliability

  14. Toroidal inhomogeneity of the vertical field in a tokamak apparatus

    International Nuclear Information System (INIS)

    Sometani, Taro; Takashima, Hidekazu

    1977-01-01

    An experiment with a model device has been made on the toroidal inhomogeneity of the vertical field in a Tokamak with an iron core. The D.C. vertical field is increased near the yokes of the iron core, while the gross plasma image field (consisting of the components due to the plasma current, the primary current, and its image) is reduced there. These two vertical fields, when superposed, exert force on the plasma as a less inhomogeneous external vertical field. The vertical field can be homogenized satisfactorily by using a compensation winding wound at a proper position on the iron core even if the shielding plates, which are mounted on some Tokamaks, are dispensed with. (auth.)

  15. Resistive demountable toroidal-field coils for tokamak reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

    1981-07-01

    Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments

  16. Scaling results for the magnetic field line trajectories in the stochastic layer near the separatrix in divertor tokamaks with high magnetic shear using the higher shear map

    International Nuclear Information System (INIS)

    Punjabi, Alkesh; Ali, Halima; Farhat, Hamidullah

    2009-01-01

    Extra terms are added to the generating function of the simple map (Punjabi et al 1992 Phys. Rev. Lett. 69 3322) to adjust shear of magnetic field lines in divertor tokamaks. From this new generating function, a higher shear map is derived from a canonical transformation. A continuous analog of the higher shear map is also derived. The method of maps (Punjabi et al 1994 J. Plasma Phys. 52 91) is used to calculate the average shear, stochastic broadening of the ideal separatrix near the X-point in the principal plane of the tokamak, loss of poloidal magnetic flux from inside the ideal separatrix, magnetic footprint on the collector plate, and its area, and the radial diffusion coefficient of magnetic field lines near the X-point. It is found that the width of the stochastic layer near the X-point and the loss of poloidal flux from inside the ideal separatrix scale linearly with average shear. The area of magnetic footprints scales roughly linearly with average shear. Linear scaling of the area is quite good when the average shear is greater than or equal to 1.25. When the average shear is in the range 1.1-1.25, the area of the footprint fluctuates (as a function of average shear) and scales faster than linear scaling. Radial diffusion of field lines near the X-point increases very rapidly by about four orders of magnitude as average shear increases from about 1.15 to 1.5. For higher values of average shear, diffusion increases linearly, and comparatively very slowly. The very slow scaling of the radial diffusion of the field can flatten the plasma pressure gradient near the separatrix, and lead to the elimination of type-I edge localized modes.

  17. On the choice of toroidal magnetic field for thermonuclear tokamaks

    International Nuclear Information System (INIS)

    Segre, S.E.

    1981-01-01

    The value of the magnetic field chosen for tokamak experiments is the result of a compromise between physics requirements, technological limits and financial constraints. The consequences of some physics requirements and limitations, in the light of recent results on the scaling of energy confinement and on limits of density are examined. (author)

  18. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. A neoclassical drift-magnetohydrodynamical fluid model of the interaction of a magnetic island chain with a resonant error-field in a high temperature tokamak plasma

    Science.gov (United States)

    Fitzpatrick, Richard

    2018-04-01

    A two-fluid, neoclassical theory of the interaction of a single magnetic island chain with a resonant error-field in a quasi-cylindrical, low-β, tokamak plasma is presented. The plasmas typically found in large hot tokamaks lie in the so-called weak neoclassical flow-damping regime in which the neoclassical ion stress tensor is not the dominant term in the ion parallel equation of motion. Nevertheless, flow-damping in such plasmas dominates ion perpendicular viscosity, and is largely responsible for determining the phase velocity of a freely rotating island chain (which is in the ion diamagnetic direction relative to the local E × B frame at the rational surface). The critical vacuum island width required to lock the island chain is mostly determined by the ion neoclassical poloidal flow damping rate at the rational surface. The stabilizing effect of the average field-line curvature, as well as the destabilizing effect of the perturbed bootstrap current, is the same for a freely rotating, a non-uniformly rotating, and a locked island chain. The destabilizing effect of the error-field averages to zero when the chain is rotating and only manifests itself when the chain locks. The perturbed ion polarization current has a small destabilizing effect on a freely rotating island chain, but a large destabilizing effect on both a non-uniformly rotating and a locked island chain. This behavior may account for the experimentally observed fact that locked island chains are much more unstable than corresponding freely rotating chains.

  1. On the simulation of the tokamak longitUdinal field

    International Nuclear Information System (INIS)

    Simakov, A.S.

    1978-01-01

    The problem of imitation of tokamak longitudinal field with a limited number of coils of a toroidal solenoid is considered in connection with construction of the bench-mark facility for the tokamak superconductive magnetic system. These coils should satisfactorily imitate the fields of the absent twenty three coils in the region of the twenty fourth. Fields and forces are calculated by the Tokat program. The analysis of the variants considered showed that with refuse from limitations on the cryostat sizes with acceptable accuracy the longitudinal field by means of 7-8 coils is possible. With the given sizes of the cryostat (d=4.1 m) it is hardly possible to obtain acceptable field imitation because of great current densities in the neighbouring coils. But on the bench of four coils one can obtain data, which, probably, will be useful during the evaluation of attaining the project parameters of the toroidal solenoid

  2. High-beta tokamak research. Annual progress report, 1 August 1982-1 August 1983

    International Nuclear Information System (INIS)

    Navratil, G.A.

    1983-08-01

    The main research objectives during the past year fell into four areas: (1) detailed observations over a range of high-beta tokamak equilibria; (2) fabrication of an improved and more flexible high-beta tokamak based on our understanding of the present Torus II; (3) extension of the pulse length to 100 usec with power crowbar operation of the equilibrium field coil sets; and (4) comparison of our equilibrium and stability observations with computational models of MHD equilibrium and stability

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

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2002-01-01

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

  4. Toroidal microinstability studies of high temperature tokamaks

    International Nuclear Information System (INIS)

    Rewoldt, G.; Tang, W.M.

    1989-07-01

    Results from comprehensive kinetic microinstability calculations are presented showing the effects of toroidicity on the ion temperature gradient mode and its relationship to the trapped-electron mode in high-temperature tokamak plasmas. The corresponding particle and energy fluxes have also been computed. It is found that, although drift-type microinstabilities persist over a wide range of values of the ion temperature gradient parameter η i ≡ (dlnT i /dr)/(dlnn i /dr), the characteristic features of the dominant mode are those of the η i -type instability when η i > η ic ∼1.2 to 1.4 and of the trapped-electron mode when η i ic . 16 refs., 7 figs

  5. Start-effect measurement of high FEL [free-electron laser] electric fields in MTX [Microwave Tokamak Experiment] by laser-aided particle-probe spectroscopy

    International Nuclear Information System (INIS)

    Oda, T.; Takiyama, K.; Odajima, K.; Ohasa, K.; Shiho, M.; Mizuno, K.; Foote, J.H.; Nilson, D.G.

    1990-01-01

    We are constructing a diagnostic system to measure the electric field (>100 kV/cm) of a free-electron laser (FEL) beam when injected into the plasma of the Microwave Tokamak Experiment (MTX). The apparatus allows a crossed-beam measurement, with 2-cm spatial resolution in the plasma, involving the FEL beam (with 140-GHz, ∼1-GW ECH pulses), a neutral-helium beam, and a dye-laser beam. After the laser beam pumps metastable helium atoms to higher excited states, their decay light is detected by an efficient optical system. Because of the Stark effect arising from the FEL electric field (rvec E), a forbidden transition can be strongly induced. The intensity of emitted light resulting from the forbidden transition is proportional to E 2 . Because photon counting rates are estimated to be low, extra effort is made to minimize background and noise levels. It is possible that the lower rvec E of an MTX gyrotron-produced ECH beam with its longer-duration pulses can also be measured using this method. Other applications of the apparatus described here may include measurements of ion temperature (using charge-exchange recombination), edge-density fluctuations, and core impurity concentrations

  6. Poloidal field distribution studies in tokamak reactor

    International Nuclear Information System (INIS)

    Ueda, Kojyu; Nishio, Satoshi; Fujisawa, Noboru; Sugihara, Masayoshi; Saito, Seiji

    1983-01-01

    On the design studies with the INTOR plasma equilibrium and poloidal field coil configuration (PFCC) from the Phase I to the Phase II A have been obtained the following main results. Three optimized PFCCs have been obtained: the INTOR-J ''Universal'' with the optimized PFCC for the divertor configuration, the optimized PFCC for the pump limiter, and the INTOR ''Universal'' with the PFCC defined as the INTOR reference. These PFCCs satisfy with the requirements for the porthole size for the remote assembly and maintenance of the device, and the maximum flux swing and current densities of the solenoidal coils. The INTOR-J ''Universal'' will be almost the same as the INTOR ''Universal'' in the reactor size. But the optimized PFCC for the pump limiter will be a little larger than the above two configuration because of being in need of slightly larger radii on the two large coils if the plasma with 1.5 in elongation is unconditionally necessary. The total sum of absolute currents with PFCC, which is used as a parameter for its figure of merit, is found to be given in a range of 80 -- 90 MAT at high beta for the divertor configuration for both of the INTOR-J ''Universal'' and the INTOR ''Universal''. The optimized PFCC for pump limiter has 70 -- 80 MAT in its range. The INTOR-J ''Universal'' and the INTOR ''Universal'' for the pump limiter will have its larger sum than one optimized for pump limiter by several MAT. The ''EF only'' method, where the flux, psi sub(P), necessary for maintaining the plasma current on high beta is provided only by EF coils, seems to give the total sum a little less than the ''EF + OH'' method using EF and OH coils for psi sub(P). (J.P.N.)

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

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

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

  8. ICRF [Ion Cyclotron Range of Frequencies] heating and antenna coupling in a high beta tokamak

    International Nuclear Information System (INIS)

    Elet, R.S.

    1988-01-01

    Maxwell's Equations are solved in two-dimensions for the electromagnetic fields in a toroidal cavity using the cold plasma fluid dielectric tensor in the Ion Cyclotron Range of Frequencies (ICRF). The Vector Wave Equation is transformed to a set of two, coupled second-order partial differential equations with inhomogeneous forcing functions which model a wave launcher. The resulting equations are finite differenced and solved numerically with a complex banded matrix algorithm on a Cray-2 computer using a code described in this report. This code is used to study power coupling characteristics of a wave launcher for low and high beta tokamaks. The low and high beta equilibrium tokamak magnetic fields applied in this model are determined from analytic solutions to the Grad-Shafranov equation. The code shows good correspondence with the results of low field side ICRF heating experiments performed on the Tokamak of Fontenay-Aux-Roses (TFR). Low field side and high field side antenna coupling properties for ICRF heating in the Columbia High Beta Tokamak (HBT) experiment are calculated with this code. Variations of antenna position in the tokamak, ionic concentration and plasma density, and volume-averaged beta have been analyzed for HBT. It is found that the location of the antenna with respect to the plasma has the dominant role in the design of an ICRF heating experiment in HBT. 10 refs., 52 figs., 13 tabs

  9. Reclosing of field lines and disruptive instability in tokamaks

    International Nuclear Information System (INIS)

    Kadomtsev, B.B.

    The mechanism of field line reclosing is proposed as the most natural explanation of disruptive instability in tokamaks. This mechanism adequately accounts for the internal disruptive instability, assuming that only mode m = 1 develops. It is extended to the presence of two or several modes. When there is a large number of allowed modes, one can speak of free reclosing, which leads to a force-free magnetic field in a diffusion discharge. In a tokamak, B/sub Z/ much greater than B/sub theta/, free reclosing leads to a uniform distribution of the current over the column cross section and to ejection of part of the poloidal flux beyond the confines of the diaphragm. It may be stated that the disruptive instability in a tokamak is an MHD activity that flares up for a short time and is permanently present in a diffusion column. The geometry of magnetic surfaces during reclosing has been analyzed, and qualitative arguments are given to show that disruptive instability begins to develop as a result of the interaction of the mode m = 2 with the inner mode m = 1

  10. Reclosing of field lines and disruptive instability in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Kadomtsev, B. B.

    1976-07-01

    The mechanism of field line reclosing is proposed as the most natural explanation of disruptive instability in tokamaks. This mechanism adequately accounts for the internal disruptive instability, assuming that only mode m = 1 develops. It is extended to the presence of two or several modes. When there is a large number of allowed modes, one can speak of free reclosing, which leads to a force-free magnetic field in a diffusion discharge. In a tokamak, B/sub Z/ much greater than B/sub theta/, free reclosing leads to a uniform distribution of the current over the column cross section and to ejection of part of the poloidal flux beyond the confines of the diaphragm. It may be stated that the disruptive instability in a tokamak is an MHD activity that flares up for a short time and is permanently present in a diffusion column. The geometry of magnetic surfaces during reclosing has been analyzed, and qualitative arguments are given to show that disruptive instability begins to develop as a result of the interaction of the mode m = 2 with the inner mode m = 1.

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

    Science.gov (United States)

    Peng, X. D.; Xu, J. Q.

    2016-08-01

    A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

  12. Toroidal field ripple effects in large tokamaks

    International Nuclear Information System (INIS)

    Uckan, N.A.; Tsang, K.T.; Callen, J.D.

    1975-01-01

    In an experimental power reactor, the ripple produced by the finite number of toroidal field coils destroys the ideal axisymmetry of the configuration and is responsible for additional particle trapping, loss regions and plasma transport. The effects of toroidal field ripple on the plasma transport coefficient, the loss of alpha particles and energetic injection ions, and the relaxation of toroidal flows are investigated in a new and systematic way. The relevant results are applied to the ORNL-EPR reference design; the maximum ripple there of about 2.2 percent at the outer edge of the plasma column is found to be tolerable from plasma physics considerations

  13. Structure of magnetic field in Tokamaks

    International Nuclear Information System (INIS)

    Heller, M.V.A.P.; Caldas, I.L.

    1990-01-01

    Magnetic surfaces, necessary to plasma confinement, can be extinguished by resonant helical perturbations with small intensities due to plasma oscillations or external helical currents. The mapping of magnetic field is obtained intergrating numerically the differential equation of its lines. Criteria which evaluate the chaotic distribution of lines between resonant magnetic islands are presented. (M.C.K.) [pt

  14. Simulation experiment on magnetic field reconnection processes in tokamak

    International Nuclear Information System (INIS)

    Kiwamoto, Y.

    1982-01-01

    Two experimental studies on magnetic field line reconnection processes relevant to tokamak physics are going on in Japan. In Yokohama National University, reconnection of poloidal magnetic field lines is studied by the author when reversing the toroidal current of a small toroidal plasma in a short period (typically less than 4 μsec). Interaction of two current carrying plasma (linear) columns is being studied by Kawashima and his coleagues in Institute of Space and Aeronautical Sciences. Mutual attraction and merging of the plasma columns and resulting plasma heating are reported. (author)

  15. First Results from Tests of High Temperature Superconductor Magnets on Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)

    2012-09-15

    Full text: It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak. In the experiment, the two copper vertical field coils of the small tokamak GOLEM were replaced by two coils each with 6 turns of HTS (Re)BCO tape. Liquid nitrogen was used to cool the coils to below the critical temperature at which HTS becomes superconducting. Little effect on the HTS critical current has been observed for perpendicular field up to 0.5 T and superconductivity has been achieved at {approx} 90.5K during bench tests. There had been concerns that the plasma pulses and pulsed magnetic fields might cause a 'quench' in the HTS, i.e., a sudden and potentially damaging transition from superconductor to normal conductor. However, many plasma pulses were fired without any quenches even when disruptions occurred with corresponding induced electrical fields. In addition, experiments without plasma have been performed to study properties of the HTS in a tokamak environment, i.e., critical current and its dependence on magnetic and electrical fields generated in a tokamak both in DC and pulsed operations, maximum current ramp-up speed, performance of the HTS tape after number of artificially induced quenches etc. No quench has been observed at DC currents up to 200 A (1.2 kA-turns through the coil). In short pulses, current up to 1 kA through the tape (6 kA-turns) has been achieved with no subsequent degradation of the HTS performance with a current ramp rate up to 0.6 MA/s. In future experiments, increases in both the plasma current and pulse duration are planned. Considerable experience has been gained during design and fabrication of the cryostat, coils, isolation and insulation, feeds and cryosystems, and GOLEM is now routinely operated with HTS coils. The

  16. Physics issues of high bootstrap current tokamaks

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  17. An investigation of coupling of the internal kink mode to error field correction coils in tokamaks

    International Nuclear Information System (INIS)

    Lazarus, E.A.

    2013-01-01

    The coupling of the internal kink to an external m/n = 1/1 perturbation is studied for profiles that are known to result in a saturated internal kink in the limit of a cylindrical tokamak. It is found from three-dimensional equilibrium calculations that, for A ≈ 30 circular plasmas and A ≈ 3 elliptical shapes, this coupling of the boundary perturbation to the internal kink is strong; i.e., the amplitude of the m/n = 1/1 structure at q = 1 is large compared with the amplitude applied at the plasma boundary. Evidence suggests that this saturated internal kink, resulting from small field errors, is an explanation for the TEXTOR and JET measurements of q 0 remaining well below unity throughout the sawtooth cycle, as well as the distinction between sawtooth effects on the q-profile observed in TEXTOR and DIII-D. It is proposed that this excitation, which could readily be applied with error field correction coils, be explored as a mechanism for controlling sawtooth amplitudes in high-performance tokamak discharges. This result is then combined with other recent tokamak results to propose an L-mode approach to fusion in tokamaks. (paper)

  18. Resistive toroidal-field coils for tokamak reactors

    International Nuclear Information System (INIS)

    Kalnavarns, J.; Jassby, D.L.

    1980-11-01

    This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized

  19. Remote replacement of TF [toroidal field] and PF [poloidal field] coils for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.; Watkin, D.C.; Hollis, M.J.; DePew, R.E.; Kuban, D.P.

    1990-01-01

    The use of deuterium-tritium fuel in the Compact Ignition Tokamak will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion devices auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. In-vessel remote maintenance included replacement of divertor and first wall hardware, faraday shields, and for an in-vessel inspection system. Provision for remote replacement of a vacuum vessel sector, toroidal field coil or poloidal field ring coil was not included in the project baseline. As a result of recent coil failures experienced at a number of facilities, the CIT project decided to reconsider the question of remote recovery from a coil failure and, in January of 1990, initiated a coil replacement study. This study focused on the technical requirements and impact on fusion machine design associated with remote recovery from any coil failure

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Resistive MHD studies of high-β-tokamak plasmas

    International Nuclear Information System (INIS)

    Lynch, V.E.; Carreras, B.A.; Hicks, H.R.; Holmes, J.A.; Garcia, L.

    1981-01-01

    Numerical calculations have been performed to study the MHD activity in hightokamaks such as ISX-B. These initial value calculations built on earlier low β techniques, but the β effects create several new numerical issues. These issues are discussed and resolved. In addition to time-stepping modules, our system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes at low β to predominantly pressure driven modes at high β is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment

  2. Drift-kinetic Alfven modes in high performance tokamaks

    International Nuclear Information System (INIS)

    Jaun, A.; Fasoli, A.F.; Testa, D.; Vaclavik, J.; Villard, L.

    2001-01-01

    The stability of fast-particle driven Alfven eigenmodes is modeled in high performance tokamaks, successively with a conventional shear, an optimized shear and a tight aspect ratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilized by mode conversion in the presence of a divertor. This suggests how conventional reactor scenarii could withstand significant pressure gradients from the fusion products. A large safety factor in the core q 0 >2.5 in deeply shear reversed configurations and a relatively large bulk ion Larmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes that are unstable in high performance JET, NSTX and ITER plasmas. (author)

  3. Safety and deterministic failure analyses in high-beta D-D tokamak reactors

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1984-01-01

    Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-01

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

  5. Stability of high β large aspect ratio tokamaks

    International Nuclear Information System (INIS)

    Cowley, S.C.

    1991-10-01

    High β(β much-gt ε/q 2 ) large aspect ratio (ε much-gt 1) tokamak equilibria are shown to be always stable to ideal M.H.D. modes that are localized about a flux surface. Both the ballooning and interchange modes are shown to be stable. This work uses the analytic high β large aspect ratio tokamak equilibria developed by Cowley et.al., which are valid for arbitrary pressure and safety factor profiles. The stability results make no assumption about these profiles or the shape of the boundary. 14 refs., 4 figs

  6. Poloidal field effects on fundamental minority ion cyclotron resonance heating in a tokamak plasma

    International Nuclear Information System (INIS)

    Jun, S. C.; Imre, Kaya; Stevens, D. C.; Weitzner, Harold; Chang, C. S.

    2000-01-01

    Minority ion fundamental cyclotron resonance is studied in a large tokamak in which the geometrical optics approximation applies off resonance and the minority average speed is less than the wave phase speeds. Poloidal equilibrium magnetic field effects are included, which lead to nontrivially nonlocal integrodifferential equations for the wave fields. Exact reciprocity relation is given as well as explicit analytic solutions for the transmission coefficients for both the high and low field side incidences. Numerical solutions are needed only for the high field side incident reflection coefficient. Numerical schemes are described and numerical results are presented together with a reliable error bound. Typically, energy absorption increases with poloidal field. The energy absorption increases with minority density at low values of minority density. However, it decreases at high minority density. Poloidal field effects weaken the dependence of energy absorption on the toroidal wave number. (c) 2000 American Institute of Physics

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

    International Nuclear Information System (INIS)

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

    1991-07-01

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

  8. The use of scaling laws for the design of high beta tokamaks

    International Nuclear Information System (INIS)

    Mauel, M.E.

    1987-01-01

    Several different empirical scaling laws for the tokamak energy confinement time are used to estimate the auxiliary heating power required for a laboratory experiment capable of testing tokamak confinement at high beta and techniques to access the second stability regime. Since operating experience in the second stability regime does not yet exist, these laws predict a wide range of possible power requirements, especially at large aspect ratios. However, by examining a model DT fusion power reactor with reasonable restrictions on the fusion island weight, neutron loading, and maximum magnetic field of the external coils, only a limited range of operating conditions are found for both first and second regime tokamaks, and only a subset of the scaling laws predict ignition. These particular scaling laws are then used to set confinement goals which if demonstrated by the laboratory experiment would indicate favourable scaling to a reactor. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

  10. Effects of Resonant Helical Field on Toroidal Field Ripple in IR-T1 Tokamak

    Science.gov (United States)

    Mahdavipour, B.; Salar Elahi, A.; Ghoranneviss, M.

    2018-02-01

    The toroidal magnetic field which is created by toroidal coils has the ripple in torus space. This magnetic field ripple has an importance in plasma equilibrium and stability studies in tokamak. In this paper, we present the investigation of the interaction between the toroidal magnetic field ripple and resonant helical field (RHF). We have estimated the amplitude of toroidal field ripples without and with RHF (with different q = m/n) ( m = 2, m = 3, m = 4, m = 5, m = 2 & 3, n = 1) using “Comsol Multiphysics” software. The simulations show that RHF has effects on the toroidal ripples.

  11. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

  12. The Role of an Electric Field in the Formation of a Detached Regime in Tokamak Plasma

    Science.gov (United States)

    Senichenkov, I.; Kaveeva, E.; Rozhansky, V.; Sytova, E.; Veselova, I.; Voskoboynikov, S.; Coster, D.

    2018-03-01

    Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-15

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

  14. Fishbone mode in high-β discharges of spherical tokamaks

    International Nuclear Information System (INIS)

    Kolesnichenko, Ya.I.; Lutsenko, V.V.; Marchenko, V.S.

    2000-01-01

    Using Hamiltonian formalism, it has been shown that well-trapped energetic ions moving outwards consume the energy of MHD perturbations through the precessional resonance provided that the plasma pressure is sufficiently high. This supports the conclusion of recent publication that the fishbone mode is stabilized in high-β discharges of spherical tokamaks. It has also been found that the presence of the velocity anisotropy of energetic ions does not change this conclusion. (author)

  15. Murakami density limit in tokamaks and reversed-field pinches

    International Nuclear Information System (INIS)

    Perkins, F.W.; Hulse, R.A.

    1984-03-01

    A theoretical upper limit for the density in an ohmically heated tokamak discharge follows from the requirement that the ohmic heating power deposited in the central current-carrying channel exceed the impurity radiative cooling in this critical region. A compact summary of our results gives this limit n/sub M/ for the central density as n/sub M/ = [Z/sub e//(Z/sub e/-1]/sup 1/2/n/sub eo/ (B/sub T//1T)(1m/R) where n/sub eo/ depends strongly on the impurity species and is remarkably independent of the central electron temperature T/sub e/(0). For T/sub e/(0) approx. 1 keV, we have n/sub eo/ = 1.5 x 10 14 cm -3 for beryllium, n/sub eo/ = 5 x 10 13 cm -3 for oxygen, n/sub eo/ = 1.0 x 10 13 cm -3 for iron, and n/sub eo/ = 0.5 x 10 13 cm -3 for tungsten. The results agree quantitatively with Murakami's original observations. A similar density limit, known as the I/N limit, exists for reversed-field pinch devices and this limit has also been evaluated for a variety of impurity species

  16. 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.

  17. Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

    1975-01-01

    A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

  18. High power RF heating and nonthermal distributions in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Peeters, A.G.

    1994-12-13

    This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL).

  19. High power RF heating and nonthermal distributions in tokamak plasmas

    International Nuclear Information System (INIS)

    Peeters, A.G.

    1994-01-01

    This thesis discusses the nonthermal effects in the electron population of a tokamak, that are generated by the inductive electric field and electron cyclotron resonant heating. The kinetic description of the plasma is given by a Boltzmann equation for the electron velocity distribution, in which the many small angle scattering Coulomb collisions that occur in the plasma are modelled by a Fokker-Planck collision term. These collisions drive the distribution towards the Maxwellian distribution of thermodynamic equilibrium. The energy absorption from the electron cyclotron waves and the acceleration by the toroidal electric field lead to deviations from the Maxwellian destribution. The interaction of the electron cyclotron waves with the plasma is treated within quasilinear theory. Resonant interaction occurs when the wave frequency matches one of the harmonics of the gyration frequency of the electrons in the static magnetic field. The waves generate a diffusion of resonant electrons in velocity space. The inductive electric field accelerates the electrons in the direction prallel to the magnetic field and leads to a convection in velocity space. The equilibrium that is reached between the driving forces of the electric field and the electron cyclotron waves and the restoring force of the collisions is studied in this thesis. The specific geometry of the tokamak is incorporated in the model through an average of the kinetic equation over the electron orbits. (orig./WL)

  20. Analytic, high β, flux conserving equilibria for cylindrical tokamaks

    International Nuclear Information System (INIS)

    Sigmar, D.J.; Vahala, G.

    1978-09-01

    Using Grad's theory of generalized differential equations, the temporal evolution from low to high β due to ''adiabatic'' and nonadiabatic (i.e., neutral beam injection) heating of a cylindrical tokamak plasma with circular cross section and peaked current profiles is calculated analytically. The influence of shaping the initial safety factor profile and the beam deposition profile and the effect of minor radius compression on the equilibrium is analyzed

  1. Analytic, high β, flux conserving equilibria for cylindrical tokamaks

    International Nuclear Information System (INIS)

    Sigmar, D.J.; Vahala, G.

    1978-01-01

    Using Grad's theory of generalized differential equations, the temporal evolution from low to high β due to ''adiabatic'' and nonadiabatic (i.e., neutral beam injection) heating of a cylindrical tokamak plasma with circular cross section and peaked current profiles is calculated analytically. The influence of shaping the initial safety factor profile and the beam deposition profile and the effect of minor radius compression on the equilibrium is analyzed

  2. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  3. Stability at high performance in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Buttery, R.J.; Akers, R.; Arends, E. =

    2003-01-01

    The development of reliable H-modes on MAST, together with advances in heating power and a range of powerful diagnostics, has provided a platform to enable MAST to address some of he most important issues of tokamak stability. In particular the high β potential of the ST is highlighted with stable operation at β N ∼5-6 , β T ∼ 16% and β p as high as 1.9, confirmed by a range of profile diagnostics. Calculations indicate that β N levels are in the vicinity of no-wall stability limits. Studies have provided the first identification of the Neoclassical Tearing Mode (NTM) in the ST, using its behaviour to quantitatively validate predictions of NTM theory, previously only applied to conventional tokamaks. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs - by avoiding large sawteeth much higher β N can, and has, been reached. Further studies have confirmed the NTM's significance, with large islands observed using the 300 point Thomson diagnostic, and locking of large n=1 modes frequently leading to disruptions. H-mode plasmas are also limited by ELMs, with confinement degraded as ELM frequency rises. However, unlike the conventional tokamak, the ELMs in high performing regimes on MAST (H IPB98Y2 ∼1) appear to be type III in nature. Modelling identifies instability to peeling modes, consistent with a type III interpretation, and shows considerable scope to raise pressure gradients (despite n=∞ ballooning theory predictions of instability) before ballooning type modes (perhaps associated with type I ELMs) occur. Finally sawteeth are shown not to remove the q=1 surface in the ST - other promising models are being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels, and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER. (author)

  4. Electrons of high perpendicular energy in the low-density regime of Tokamaks

    International Nuclear Information System (INIS)

    Bornatici, M.; Engelmann, F.

    1978-01-01

    Effects due to instabilities excited in the low-density regime of tokamaks by runaway electrons via the cyclotron resonance ω+Ω=kV along with the formation of a positive slope in the runaway distribution are considered. Conditions for the production of electrons of high perpendicular energy and their trapping in toroidal field ripples, leading to liner damage, are discussed and found to be rather stringent. Fairly good agreement with the experiments is found

  5. Optical visualization of electric and magnetic field perturbations in tokamak discharges by hydrogen pellet injection

    International Nuclear Information System (INIS)

    Drawin, H.W.; Dubois, M.A.

    1992-04-01

    Two-dimensional intensity distribution mappings of photographs of pellet ablation cloud trajectories in the TFR and TS tokamaks reveal irregular shapes of the luminous striations. The observed features are not well understood, but can be described and interpreted as the first direct visual observation of pre-existing electric and/or magnetic field perturbations in the hot core of tokamak plasmas. It is suggested to use pellet injection as a diagnostic tool for the study of plasma structures and transport phenomena

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

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2000-01-01

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

  7. The roles of electric field shear and Shafranov shift in sustaining high confinement in enhanced reversed shear plasmas on the TFTR tokamak

    International Nuclear Information System (INIS)

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

    1997-02-01

    The relaxation of core transport barriers in TFTR Enhanced Reversed Shear plasmas has been studied by varying the radial electric field using different applied torques from neutral beam injection. Transport rates and fluctuations remain low over a wide range of radial electric field shear, but increase when the local E x B shearing rates are driven below a threshold comparable to the fastest linear growth rates of the dominant instabilities. Shafranov-shift-induced stabilization alone is not able to sustain enhanced confinement

  8. Optimization of magnetic field system for glass spherical tokamak GLAST-III

    International Nuclear Information System (INIS)

    Ahmad, Zahoor; Ahmad, S; Naveed, M A; Deeba, F; Javeed, M Aqib; Batool, S; Hussain, S; Vorobyov, G M

    2017-01-01

    GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/ R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA −1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils. (paper)

  9. Spectroscopic study of turbulent heating in the high beta tokamak - Torus II

    International Nuclear Information System (INIS)

    Georgiou, G.E.

    1979-01-01

    Visible spectroscopy, involving line profile and line intensity measurements, was used to study the turbulent heating of the rectangular cross-section high-beta tokamak Torus II. The spectroscopy was done in the visible wave-length region using a six channel polychrometer having 0.2 A resolution, which is capable of radial scans of the plasma. The plasma, obtained by ionizing helium, is heated by poloidal skin currents, induced by a rapid (tau/sub R/ approx. = 1.7 μsec) change of the toroidal magnetic field either parallel or anti-parallel to the initial toroidal bias magnetic field, which converts a cold toroidal Z-pinch plasma into a hot tokamak plasma

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-02-01

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

  12. Study of the behaviour of chaotic magnetic field lines in a tokamak

    International Nuclear Information System (INIS)

    Hugon, M.; Mendonca, J.T.; Rebut, P.H.

    1989-01-01

    One plausible explanation for the observed anomalous energy losses in tokamaks is the existence of a confinement region in the plasma, where magnetic islands coexist with stochastic magnetic field lines. It is the purpose of the present work to elucidate the statistical behaviour of the stochastic field lines. This is done by numerically integrating the field line equations. (author) 4 refs., 2 figs

  13. LLNL high-field coil program

    International Nuclear Information System (INIS)

    Miller, J.R.

    1986-01-01

    An overview is presented of the LLNL High-Field Superconducting Magnet Development Program wherein the technology is being developed for producing fields in the range of 15 T and higher for both mirror and tokamak applications. Applications requiring less field will also benefit from this program. In addition, recent results on the thermomechanical performance of cable-in-conduit conductor systems are presented and their importance to high-field coil design discussed

  14. Impact of maximum TF magnetic field on performance and cost of an advanced physics tokamak

    International Nuclear Information System (INIS)

    Reid, R.L.

    1983-01-01

    Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of variation in the maximum value of the field at the toroidal field (TF) coils on the performance and cost of a low q/sub psi/, quasi-steady-state tokamak. Marginal ignition, inductive current startup plus 100 s of inductive burn, and a constant value of epsilon (inverse aspect ratio) times beta poloidal were global conditions imposed on this study. A maximum TF field of approximately 10 T was found to be appropriate for this device

  15. Quick profile-reorganization driven by helical field perturbation for suppressing tokamak major disruptions

    International Nuclear Information System (INIS)

    Yamazaki, K.; Kawahata, K.; Ando, R.

    1986-09-01

    Disruptive behavior of magnetic field configuration leading to tokamak major disruption is found to be controlled by a mild ''mini-disruption'' which is induced by the compact external modular multipole-field coils with m = 3/n = 2 dominant helical field component in the JIPP T-IIU tokamak. This mini-disruption ergodizes the m = 2/n = 1 magnetic island quickly but mildly and then prevents the profile of electron temperature from flattening. This quick profile-reorganization is effective to avoid the two-step disruption (pre- and major disuptions) responsible for the chatastrophic current termination. (author)

  16. Limits on the field of ohmic heating solenoids, applied to a tokamak TNS

    International Nuclear Information System (INIS)

    Turner, L.R.

    1977-01-01

    If the ohmic heating solenoid for the TNS or other large tokamak is an ungraded cryostable superconducting solenoid, with NbTi at 4.2 K as the superconductor, then the smallest outer diameter is not achieved at the highest attainable field. There is a lower optimum field which minimizes the outer diameter for a given volt-second requirement. At higher fields the mean diameter decreases; but the high fields require more superconductor, more copper stabilizer, more stainless steel for support, and more liquid helium coolant. For the GA-ANL design for TNS, the optimum field is 7.55 T and the minimum outside diameter for the solenoid is 2.15 m. If, on the other hand, the solenoid is graded, with more NbTi, copper, and stainless steel on the inner turns where the field is higher, than the volt-seconds can always be increased, for a given outer diameter, by adding more turns at a higher field inside until either the critical field is reached or the solenoid bore is filled. However, the material and money required to add a few more volt-seconds increases rapidly with field

  17. Limits on the field of ohmic heating solenoids, applied to a tokamak TNS

    International Nuclear Information System (INIS)

    Turner, L.R.

    1978-01-01

    If the ohmic heating solenoid for the TNS or other large tokamak is an ungraded cryostable superconducting solenoid, with NbTi at 4.2 K as the superconductor, then the smallest outer diameter is not achieved at the highest attainable field. There is a lower optimum field which minimizes the outer diameter for a given volt-second requirement. At higher fields the mean diameter decreases; but the high fields require more superconductor, more copper stabilizer, more stainless steel for support, and more liquid helium coolant. For the GA-ANL design for TNS, the optimum field is 7.55 T and the minimum outside diameter for the solenoid is 2.15 m. If, on the other hand, the solenoid is graded, with more NbTi, copper, and stainless steel on the inner turns where the field is higher, then the volt-seconds can always be increased, for a given outer diameter, by adding more turns at a higher field inside until either the critical field is reached or the solenoid bore is filled. However, the material and money required to add a few more volt-seconds increases rapidly with field

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    1998-01-01

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

  20. Development of FEMAG. Calculation code of magnetic field generated by ferritic plates in the tokamak devices

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2003-03-01

    In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)

  1. Tore-Supra: a Tokamak with superconducting toroidal field coils

    International Nuclear Information System (INIS)

    Turck, B.

    1987-07-01

    Tore Supra is a tokamak under construction on the site of Cen Cadarache by the Euratom-CEA Association. The machine technology integrates all problems related to the fabrication and the operation of large superconducting coils and of the associated cryogenic system. Tore Supra will provide a significant experience to prepare the next generation of machines for plasma physics and controlled fusion. Tore Supra is specially designed to implement a large physics program. The superconducting coils make possible the study of plasma confinement in long pulses (more than 60s), the impurities and the stability, and the efficiency of additional heating sources (neutral particle beams and radio frequency heating). The opportunity is taken to recall the particular features and requirements of the superconducting coils of the large future tokamaks in order to point out the problems that have to be faced by any new material (superconducting or not)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. High density plasma heating in the Tokamak à configuration variable

    International Nuclear Information System (INIS)

    Curchod, L.

    2011-04-01

    The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2 nd harmonic EC power in ordinary (O 2 ) or extraordinary (X 2 ) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3 rd harmonic EC power (X 3 ) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X 2 and X 3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O 2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation

  4. Stabilization of the Vertical Mode in Tokamaks by Localized Nonaxisymmetric Fields

    International Nuclear Information System (INIS)

    Reiman, A.

    2007-01-01

    Vertical instability of a tokamak plasma can be controlled by nonaxisymmetric magnetic fields localized near the plasma edge at the bottom and top of the torus. The required magnetic fields can be produced by a relatively simple set of parallelogram-shaped coils.

  5. Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks

    Science.gov (United States)

    Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

    2017-10-01

    The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, W

    1979-01-01

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

  7. Tokamak ion temperature and poloidal field diagnostics using 3 MeV protons

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Strachan, J.D.

    1984-10-01

    The 3 MeV protons created by d(d,p)t fusion reactions in a moderately sized tokamak leave the plasma on trajectories determined by the position of their birth and by the poloidal magnetic field. Pitch-angle resolution of the escaping 3 MeV protons can separately resolve the spatial distribution of the d(d,p)t fusion reactions and the poloidal field distribution inside the tokamak. These diagnostic techniques have been demonstrated on PLT with an array of collimated surface barrier detectors

  8. A comprehensive theory of the equilibria in a tokamak and a reversed field pinch

    International Nuclear Information System (INIS)

    Chiyoda, Katsuji

    1996-01-01

    The equilibrium configuration of a tokamak is analysed by the equilibrium equations derived for analysing a reversed field pinch (RFP). The expressions of the magnetic field and the toroidal shift in the internal plasma region and the external vacuum region are obtained. The expressions in the vacuum region become the Shafranov's expressions, when the plasma-center coordinates is used. Discontinuities of the equilibrium quantities are considered. It is concluded that the equilibrium equations are applicable also to the tokamak plasma and that the difference of the equilibria between the tokamak and the RFP stems from the choices of the pressure and the toroidal current function. A feature of our theory is that any ordering to the safety factor is not imposed. (author)

  9. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    International Nuclear Information System (INIS)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

  10. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  11. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

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

  12. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

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

  13. Effects of q and high beta on tokamak stability

    International Nuclear Information System (INIS)

    Brickhouse, N.S.; Callen, J.D.; Dexter, R.N.

    1984-08-01

    In the Columbia University Torus II tokamak plasmas have been studied with volume averaged toroidal beta values as high as 15%. Experimental equilibria have been compared with a 2D free boundary MHD equilibrium code PSEC. The stability of these equilibria has been computed using PEST, the predictions of which are compatible with an observed instability in Torus II which may be characterized as a high toroidal mode number ballooning fluctuation. In the University of Wisconsin Tokapole II tokamak disruptive instability behavior is investigated, with plasma able to be confined on closed magnetic surfaces in the scrape-off region, as the cylindrical edge safety factor is varied from q approx. 3 to q approx. 0.5. It is observed that at q/sub a/ approx. 3 major disruption activity occurs without current terminations, at q/sub a/ less than or equal to 2 well-confined plasmas are obtained without major disruption, and at q/sub a/ approx. 0.5 only partial reconnection accompanies minor disruptions

  14. Implementation of vertically asymmetric toroidal-field ripple for beam heating of tokamak reactor plasmas

    International Nuclear Information System (INIS)

    Jassby, D.L.; Sheffield, G.V.; Towner, H.H.; Weissenburger, D.W.

    1976-10-01

    The neutral-beam energy required for adequate penetration of tokamak plasmas of high opacity can be reduced by a large factor if the beam is injected vertically into a region of large TF (toroidal-field) ripple. Energetic ions are trapped in local magnetic wells and drift vertically toward the midplane (z = 0). If the ripple is made very small on the opposite side of the midplane, drifting ions are detrapped and thermalized in the central plasma region. This paper discusses design considerations for establishing the required vertically asymmetric ripple. Examples are given of special TF-coil configurations, and of the use of auxiliary coil windings to create the prescribed ripple profiles

  15. Eigenmode field structure of the fast magnetosonic wave in a Tokamak and loading impedance of coupling structures

    International Nuclear Information System (INIS)

    Adam, J.; Jacquinot, J.

    1977-04-01

    Detailed calculations concerning the field structure and excitation of the fast magnetosonic wave are presented keeping in mind RF heating of a Tokamak near the ion cyclotron harmonic. The new contributions are - a discussion of the cylindrical problem in an inhomogeneous plasma including surface waves and the splitting of the eigenmodes by the poloidal field - a calculation of the field structure in the toroidal cavity resonator and the application to mode tracking - a formulation of the loading impedance of various coupling structures: array of coils in the low frequency limit or transmission lines in the high frequency case

  16. 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

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

    Shimada, Ryuichi

    1982-01-01

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

  20. HYFIRE: a tokamak/high-temperature electrolysis system

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.P.; Benenati, R.; Varljen, T.C.; Chi, J.W.H.; Karbowski, J.S.

    1981-01-01

    The HYFIRE studies to date have investigated a number of technical approaches for using the thermal energy produced in a high-temperature Tokamak blanket to provide the electrical and thermal energy required to drive a high-temperature (> 1000 0 C) water electrolysis process. Current emphasis is on two design points, one consistent with electrolyzer peak inlet temperatures of 1400 0 C, which is an extrapolation of present experience, and one consistent with a peak electrolyzer temperature of 1100 0 C. This latter condition is based on current laboratory experience with high-temperature solid electrolyte fuel cells. Our major conclusion to date is that the technical integration of fusion and high-temperature electrolysis appears to be feasible and that overall hydrogen production efficiencies of 50 to 55% seem possible

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

    International Nuclear Information System (INIS)

    Ueda, Yoshinobu; Ono, Yasushi

    1999-01-01

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

  2. βp-collapse-induced vertical displacement event in high βp tokamak disruption

    International Nuclear Information System (INIS)

    Nakamura, Y.; Yoshino, R.; Pomphrey, N.; Jardin, S.C.

    1996-01-01

    Extremely fast vertical displacement events (VDEs) induced by a strong β p collapse were found in a vertically elongated (κ ∼ 1.5), high β p (β p ∼ 1.7) tokamak with a resistive shell through computer simulations using the tokamak simulation code. Although the plasma current quench which has been shown to be the prime cause of VDEs in a relatively low β p tokamak (β p ∼ 0.2) (Nakamura Y et al 1996 Nucl. Fusion 36 643), was not observed during the VDE evolution, the observed growth rate of VDEs was almost five times (γ ∼ 655 s -1 ) faster than the growth rate of the usual positional instability (γ ∼ 149 s -1 ). The essential mechanism of the β p -collapse-induced VDE was clarified to be the intense enhancement of positional instability due to a large and sudden degradation of the magnetic field decay n-index in addition to the significant destabilization due to a reduction in the stability index n s . The radial shift of the magnetic axis caused by the β p collapse induces eddy currents on the resistive shell, and these eddy currents produce a large degradation of the n-index. (author)

  3. Sustained high βN plasmas on EAST tokamak

    Science.gov (United States)

    Gao, Xiang; the EAST team

    2018-05-01

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

  4. Dynamical analysis of the magnetic field line evolution in tokamaks with ergodic limiters

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, Kai; Caldas, Ibere L. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    1997-12-31

    Full text. Magnetic ergodic limiters are commonly used to control chaos in the tokamak border and several models have been developed to study the influence of these limiters on the magnetic field line evolution in the tokamak vessel. In this work we derive a bidimensional symplectic mapping describing this evolution with toroidal corrections. Poincare plots presenting typical Hamiltonian behaviour, such as island chains and hetero clinic and homo clinic orbits are obtained. Then we perform the dynamical analysis of these Poincare plots using standard algorithms such as calculation of Lyapunov exponents, safety factors, FFT spectra and parameters space plots to perform the dynamical analysis. (author)

  5. Development of superconducting poloidal field coils for medium and large size tokamaks

    International Nuclear Information System (INIS)

    Dittrich, H.-G.; Forster, S.; Hofmann, A.

    1983-01-01

    Large long pulse tokamak fusion experiments require the use of superconducting poloidal field (PF) coils. In the past not much attention has been paid to the development of such coils. Therefore a development programme has been initiated recently at KfK. In this report start with summarizing the relevant PF coil parameters of some medium and large size tokamaks presently under construction or design, respectively. The most important areas of research and development work are deduced from these parameters. Design considerations and first experimental results concerning low loss conductors, cooling concepts and structural components are given

  6. On the applicability of the standard approaches for evaluating a neoclassical radial electric field in a tokamak edge region

    Energy Technology Data Exchange (ETDEWEB)

    Dorf, M. A.; Cohen, R. H.; Joseph, I. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Simakov, A. N. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)

    2013-08-15

    The use of the standard approaches for evaluating a neoclassical radial electric field E{sub r}, i.e., the Ampere (or gyro-Poisson) equation, requires accurate calculation of the difference between the gyroaveraged electron and ion particle fluxes (or densities). In the core of a tokamak, the nontrivial difference appears only in high-order corrections to a local Maxwellian distribution due to the intrinsic ambipolarity of particle transport. The evaluation of such high-order corrections may be inconsistent with the accuracy of the standard long wavelength gyrokinetic equation (GKE), thus imposing limitations on the applicability of the standard approaches. However, in the edge of a tokamak, charge-exchange collisions with neutrals and prompt ion orbit losses can drive non-intrinsically ambipolar particle fluxes for which a nontrivial (E{sub r}-dependent) difference between the electron and ion fluxes appears already in a low order and can be accurately predicted by the long wavelength GKE. The parameter regimes, where the radial electric field dynamics in the tokamak edge region is dominated by the non-intrinsically ambipolar processes, thus allowing for the use of the standard approaches, are discussed.

  7. Theoretical considerations and preparatory experiments for poloidal field measurements in tokamaks by far-infrared polarimetry

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, W; Dodel, G [Stuttgart Univ. (TH) (Germany, F.R.). Inst. fuer Plasmaforschung

    1978-12-01

    Numerical calculations give an optimum wavelength and show the precision requirements for determining poloidal field profiles in tokamaks on the basis of the Faraday effect. The required precision of the polarimetric measurements can be achieved in the far-infrared as is verified in a model experiment using a ferrite modulated HCN laser beam.

  8. Tokamak edge electron diffusion and distribution function in the lower hybrid antenna electric field

    Czech Academy of Sciences Publication Activity Database

    Fuchs, Vladimír; Gunn, J. P.; Goniche, M.; Petržílka, Václav

    2003-01-01

    Roč. 43, č. 5 (2003), s. 341-351 ISSN 0029-5515 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, grill electric field Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.390, year: 2003

  9. Linear and Nonlinear Response of a Rotating Tokamak Plasma to a Resonant Error-Field

    Science.gov (United States)

    Fitzpatrick, Richard

    2014-10-01

    An in-depth investigation of the effect of a resonant error-field on a rotating, quasi-cylindrical, tokamak plasma is preformed within the context of resistive-MHD theory. General expressions for the response of the plasma at the rational surface to the error-field are derived in both the linear and nonlinear regimes, and the extents of these regimes mapped out in parameter space. Torque-balance equations are also obtained in both regimes. These equations are used to determine the steady-state plasma rotation at the rational surface in the presence of the error-field. It is found that, provided the intrinsic plasma rotation is sufficiently large, the torque-balance equations possess dynamically stable low-rotation and high-rotation solution branches, separated by a forbidden band of dynamically unstable solutions. Moreover, bifurcations between the two stable solution branches are triggered as the amplitude of the error-field is varied. A low- to high-rotation bifurcation is invariably associated with a significant reduction in the width of the magnetic island chain driven at the rational surface, and vice versa. General expressions for the bifurcation thresholds are derived, and their domains of validity mapped out in parameter space. This research was funded by the U.S. Department of Energy under Contract DE-FG02-04ER-54742.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  13. Magnetic shielding for FEL microwave electric field diagnostic in MTX tokamak

    International Nuclear Information System (INIS)

    Yamada, Shinichi; Odajima, Kazuo; Ishida, Hiroyasu

    1991-07-01

    A diagnostic system for measurement of microwave electric field from free electron laser (FEL) is in preparation at JAERI under JAERI-DOE collaborative program in the Microwave Tokamak Experiment (MTX) being held at Lawrence Livermore National Laboratory in U.S.A.. That is called LAPPS (Laser Aided Particle Probe Spectroscopy). This is consist of helium neutral beam source, a dye laser and viewing optics. It is required that 1000 gauss of the magnetic field must be shielded to less than 1 gauss in order to operate these LAPPS components. New high performance soft ferrous magnetic material 'FERROPERM' and PERMALLOY are used on this purpose. This paper proposes a new method to estimate a required thickness of the magnetic shielding in a saturated region of B-H curve, that is, 'magnetic shielding calculation by Virtual Divided Layers Method (VDLM)', where the shielding layer is virtually divided in many layers in the calculation. The results are compared with a computer simulation using 'three dimensional static magnetic field code' and with experimental results in a uniform static field. (author)

  14. ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-12-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

  15. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Coenen, Jan Willem

    2009-11-06

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E{sub r}. The ergodic zone causes an electron loss, and subsequently a (vector)j x (vector)B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q{sub a}, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time {tau}{sub p} (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E{sub r}. Transport is reduced and the E{sub r} shear is increased locally at q=5/2 up to 1.5 . 10{sup 5}s{sup -1}, while the E{sub r} becomes more positive. (orig.)

  16. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    International Nuclear Information System (INIS)

    Coenen, Jan Willem

    2009-01-01

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E r . The ergodic zone causes an electron loss, and subsequently a vector j x vector B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q a , either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time τ p (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E r . Transport is reduced and the E r shear is increased locally at q=5/2 up to 1.5 . 10 5 s -1 , while the E r becomes more positive. (orig.)

  17. High n ballooning modes in highly elongated tokamaks

    International Nuclear Information System (INIS)

    An, C.H.; Bateman, G.

    1980-02-01

    An analytic study of stability against high n ballooning modes in highly elongated axisymmetric plasmas is presented and compared with computational results. From the equation for the marginal pressure gradient, it is found that the local shear plays an important role on the stability of elongated and shifted plasma, and that high elongation deteriorates the stability by decreasing the stabilizing effects of field line bending and local shear. The net contribution of the local shear to stability decreases with elongation and shift for strongly ballooning modes (eigenfunctions strongly localized near the outer edge of the toroidal flux surfaces) but increases for interchange modes (eigenfunctions more uniform along the flux surfaces). The computational study of high n ballooning modes in a highly elongated plasma reveals that lowering the aspect ratio and broadening the pressure profile enhance the marginal beta for β/sub p/ less than unity but severely reduce the marginal beta for β/sub p/ larger than unity

  18. Experience with high heat flux components in large tokamaks

    International Nuclear Information System (INIS)

    Chappuis, P.; Dietz, K.J.; Ulrickson, M.

    1991-01-01

    The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

  19. High performance operational limits of tokamak and helical systems

    International Nuclear Information System (INIS)

    Yamazaki, Kozo; Kikuchi, Mitsuru

    2003-01-01

    The plasma operational boundaries of tokamak and helical systems are surveyed and compared with each other. Global confinement scaling laws are similar and gyro-Bohm like, however, local transport process is different due to sawtooth oscillations in tokamaks and ripple transport loss in helical systems. As for stability limits, achievable tokamak beta is explained by ideal or resistive MHD theories. On the other hand, beta values obtained so far in helical system are beyond ideal Mercier mode limits. Density limits in tokamak are often related to the coupling between radiation collapse and disruptive MHD instabilities, but the slow radiation collapse is dominant in the helical system. The pulse length of both tokamak and helical systems is on the order of hours in small machines, and the longer-pulsed good-confinement plasma operations compatible with radiative divertors are anticipated in both systems in the future. (author)

  20. Error field and its correction strategy in tokamaks

    International Nuclear Information System (INIS)

    In, Yongkyoon

    2014-01-01

    While error field correction (EFC) is to minimize the unwanted kink-resonant non-axisymmetric components, resonant magnetic perturbation (RMP) application is to maximize the benefits of pitch-resonant non-axisymmetric components. As the plasma response against non-axisymmetric field increases with beta increase, feedback-controlled EFC is a more promising EFC strategy in reactor-relevant high-beta regimes. Nonetheless, various physical aspects and uncertainties associated with EFC should be taken into account and clarified in the terms of multiple low-n EFC and multiple MHD modes, in addition to the compatibility issue with RMP application. Such a multi-faceted view of EFC strategy is briefly discussed. (author)

  1. High resolution detection and excitation of resonant magnetic perturbations in a wall-stabilized tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Maurer, David A. [Physics Department, Auburn University, Auburn, Alabama 36849 (United States); Shiraki, Daisuke; Levesque, Jeffrey P.; Bialek, James; Angelini, Sarah; Byrne, Patrick; DeBono, Bryan; Hughes, Paul; Mauel, Michael E.; Navratil, Gerald A.; Peng Qian; Rhodes, Dov; Rath, Nickolaus; Stoafer, Christopher [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2012-05-15

    We report high-resolution detection of the 3D plasma magnetic response of wall-stabilized tokamak discharges in the High Beta Tokamak-Extended Pulse [T. H. Ivers et al., Phys. Plasmas 3, 1926 (1996)] device. A new adjustable conducting wall has been installed on HBT-EP made up of 20 independent, movable, wall segments instrumented with three distinct sets of 40 modular coils that can be independently driven to generate a wide variety of magnetic perturbations. High-resolution detection of the plasma response is made with 216 poloidal and radial magnetic sensors that have been located and calibrated with high-accuracy. Static and dynamic plasma responses to resonant and non-resonant magnetic perturbations are observed through measurement of the step-response following a rapid change in the toroidal phase of the applied perturbations. Biorthogonal decomposition of the full set of magnetic sensors clearly defines the structures of naturally occurring external kinks as being composed of independent m/n = 3/1 and 6/2 modes. Resonant magnetic perturbations were applied to discharges with pre-existing, saturated m/n = 3/1 external kink mode activity. This m/n = 3/1 kink mode was observed to lock to the applied perturbation field. During this kink mode locked period, the plasma resonant response is characterized by a linear, a saturated, and a disruptive plasma regime dependent on the magnitude of the applied field and value of the edge safety factor and plasma rotation.

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

    Science.gov (United States)

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

    2007-08-10

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

  3. Neoclassical Drift of Circulating Orbits Due toToroidal Electric Field in Tokamaks

    International Nuclear Information System (INIS)

    Qin, Hong; Guan, Xiaoyin; Fisch, Nathaniel J.

    2011-01-01

    In tokamaks, Ware pinch is a well known neoclassical effect for trapped particles in response to a toroidal electric field. It is generally believed that there exists no similar neoclassical effect for circulating particles without collisions. However, this belief is erroneous, and misses an important effect. We show both analytically and numerically that under the influence of a toroidal electric field parallel to the current, the circulating orbits drift outward toward the outer wall with a characteristic velocity O ((var e psilon) -1 ) larger than the E x B velocity, where (var e psilon) is the inverse aspect-ratio of a tokamak. During a RF overdrive, the toroidal electric field is anti-parallel to the current. As a consequence, all charged particles, including backward runaway electrons, will drift inward towards the inner wall.

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

    International Nuclear Information System (INIS)

    Payan, J.

    1994-05-01

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

  5. Measurement of the radial electric field in the ASDEX tokamak

    International Nuclear Information System (INIS)

    Field, A.R.; Fussmann, G.; Hofmann, J.V.

    1990-12-01

    The radial electric field (E Τ ) at the plasma periphery is determined by measuring the drift velocities of low-Z impurities ions (BIV, CIII and HeII). The measurements are performed with a scannable mirror system which allows the determination of the poloidal, perpendicular (to B vector) and toroidal components of the drift velocities from the differential Doppler shift of visible line emission observed along opposing viewing directions. The principle of the measurement is investigated in detail. In particular, it is shown that for radially localised emission shells there exits a line of sight oriented perpendicular to B vector along which E Τ may be inferred directly from the observed Doppler shift of the line emission. Along such a line of sight the net contribution to the shift from the diamagnetic drift and the radial gradient of the excitation probability is negligible. During the Ohmic- and L-phases the perpendicular drift velocity of the BIV ions measured approximately 2 cm inside the separatrix is small (≤ 2 kms -1 ) and in the ion diamagnetic drift direction. However, at the L → H-Mode transition it changes sign and begins to increase on the time-scale of the edge pressure gradients reaching the highest values at the end of the H * -phase. From these high perpendicular drift velocities it is infered that, in the H-mode, there exists a strong negative radial electric field (vertical strokeE τ vertical stroke ≤ kVm -1 ) just inside the separatrix. The dependence of the drift velocity of the BIV ions and E Τ on the NBI-heating power and the magnitude and direction of the plasma current and the magnetic field is investigated. (orig.)

  6. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  7. HYFIRE: a tokamak-high-temperature electrolysis system

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.; Steinberg, M.; Benenati, R.; Horn, F.; Isaacs, H.; Lazareth, O.W.; Makowitz, H.; Usher, J.

    1980-01-01

    Brookhaven National Laboratory (BNL) is carrying out a comprehensive conceptual design study called HYFIRE of a commercial fusion Tokamak reactor, high-temperature electrolysis system. The study is placing particular emphasis on the adaptability of the STARFIRE power reactor to a synfuel application. The HYFIRE blanket must perform three functions: (a) provide high-temperature (approx. 1400 0 C) process steam at moderate pressures (in the range of 10 to 30 atm) to the high-temperature electrolysis (HTE) units; (b) provide high-temperature (approx. 700 0 to 800 0 C) heat to a thermal power cycle for generation of electricity to the HTE units; and (c) breed enough tritium to sustain the D-T fuel cycle. In addition to thermal energy for the decomposition of steam into its constituents, H 2 and O 2 , electrical input is required. Fourteen hundred degree steam coupled with 40% power efficiency results in a process efficiency (conversion of fusion energy to hydrogen chemical energy) of 50%

  8. HYFIRE: a tokamak-high-temperature electrolysis system

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.; Steinberg, M.; Benenati, R.; Horn, F.; Isaacs, H.; Lazareth, O.W.; Makowitz, H.; Usher, J.

    1980-01-01

    Brookhaven National Laboratory (BNL) is carrying out a comprehensive conceptual design study called HYFIRE of a commercial fusion Tokamak reactor, high-temperature electrolysis system. The study is placing particular emphasis on the adaptability of the STARFIRE power reactor to a synfuel application. The HYFIRE blanket must perform three functions: (a) provide high-temperature (approx. 1400 0 C) process steam at moderate pressures (in the range of 10 to 30 atm) to the high-temperature electrolysis (HTE) units; (b) provide high-temperature (approx. 700 0 to 800 0 C) heat to a thermal power cycle for generation of electricity to the HTE units; and (c) breed enough tritium to sustain the D-T fuel cycle. In addition to thermal energy for the decomposition of steam into its constituents, H 2 and O 2 , electrical input is required. Fourteen hundred degree steam coupled with 40% power cycle efficiency results in a process efficiency (conversion of fusion energy to hydrogen chemical energy) of 50%

  9. Gyrokinetic full f analysis of electric field dynamics and poloidal velocity in the FT2-tokamak configuration

    International Nuclear Information System (INIS)

    Leerink, S.; Heikkinen, J. A.; Janhunen, S. J.; Kiviniemi, T. P.; Nora, M.; Ogando, F.

    2008-01-01

    The ELMFIRE gyrokinetic simulation code has been used to perform full f simulations of the FT-2 tokamak. The dynamics of the radial electric field and the creation of poloidal velocity in the presence of turbulence are presented.

  10. Collective processes in a tokamak with high-energy particles: general problems of the linear theory of Alfven instabilities of a tokamak with high-energy ions

    International Nuclear Information System (INIS)

    Mikhailovskii, A.B.

    1986-01-01

    Some general problems of the theory of Alfven instabilities of a tokamak with high-energy ions are considered. It is assumed that such ions are due to either ionization of fast neutral atoms, injected into the tokamak, or production of them under thermo-nuclear conditions. Small-oscillation equations are derived for the Alfven-type waves, which allow for both destabilizing effects, associated with the high-energy particles, and stabilizing ones, such as effects of shear and bulk-plasm dissipation. A high-energy ion contribution is calculated into the growth rate of the Alfven waves. The author considers the role of trapped-electron collisional dissipation

  11. Test particle calculations for the Texas experimental tokamak with resonant magnetic fields

    International Nuclear Information System (INIS)

    Wootton, A.J.; McCool, S.C.; Zheng, S.

    1991-01-01

    This paper presents a simple test particle model that attempts to describe particle motion in the presence of intrinsic electrostatic fluctuations in a prescribed tokamak magnetic field. In particular, magnetic field configurations that include externally produced magnetic islands and stochastic regions are considered. The resulting test particle transport is compared with the predictions of analytic models and with the experimentally measured electron heat and particle transport on the Texas Experimental Tokamak (TEXT). Agreement between the test particle results and applicable analytic theories is found. However, there is only partial agreement with the experimental results, and possible reasons for the discrepancies are explored. Good agreement is found between predicted and measured spatially asymmetric particle distributions. The particle collection efficiency of an apertured limiter inside a magnetic island (an intra-island pump limiter) is discussed

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Wang, Jian-Hua.

    1990-01-01

    Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n e ∼ 1 - 5 x 10 14 (cm -3 ), T e ∼ 4 - 10 (eV), B t ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of C II (4267A, emitted by the C + ions) and C III (4647A, emitted by the C ++ ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ions is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is ''classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the H α emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time τ p is comparable with the plasma energy confinement time τ E ; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy

  14. High-energy ion tail formation due to ion acoustic turbulence in the TRIAM-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Kazuo; Hiraki, Naoji; Nakamura, Yukio; Itoh, Satoshi [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

    1982-02-01

    The two-component ion energy spectra observed in the TRIAM-1 tokamak are explained as a result of the high-energy ion tail formation due to ion acoustic turbulence driven by a toroidal current pulse for turbulent heating.

  15. Measurement of poloidal field distributions in Tokamaks by far-infrared polarimetry

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, W [Stuttgart Univ. (TH) (Germany, F.R.). Inst. fuer Physikalische Elektronik; Dodel, G

    1978-02-01

    A sufficiently precise measurement of small Faraday rotation angles in a Tokamak experiment should be possible. Besides the precision with which the Faraday rotation angles can be measured, the precision with which the poloidal field can finally be deduced from such measurements, depends on other factors such as the measuring precision of the electron density profile, the diameter of the probing beam and the deviation of the plasma from circular symmetry.

  16. Magnetic propulsion of intense lithium streams in a tokamak magnetic field

    International Nuclear Information System (INIS)

    Zakharov, Leonid E.

    2003-01-01

    This paper describes the effect and gives the theory of magnetic propulsion which allows driving free surface plasma facing liquid lithium streams in tokamaks. In the approximation of a thin flowing layer the MHD equations are reduced to one integrodifferential equation which takes into account the propulsion effect, viscosity, and the drag force due to magnetic pumping and other interactions with the magnetic field. A stability criterion is obtained for stabilization of the 'sausage' instability of the streams by centrifugal force

  17. Computational model for superconducting toroidal-field magnets for a tokamak reactor

    International Nuclear Information System (INIS)

    Turner, L.R.; Abdou, M.A.

    1978-01-01

    A computational model for predicting the performance characteristics and cost of superconducting toroidal-field (TF) magnets in tokamak reactors is presented. The model can be used to compare the technical and economic merits of different approaches to the design of TF magnets for a reactor system. The model has been integrated into the ANL Systems Analysis Program. Samples of results obtainable with the model are presented

  18. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  19. Structured Cable for High-Current Coils of Tokamaks

    Science.gov (United States)

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

    2011-10-01

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

  20. Investigations of radial electric field and global circulation layer in limiter tokamaks

    International Nuclear Information System (INIS)

    Zagorski, R.; Gerhauser, H.; Lehnen, M.; Loarer, T.

    2002-01-01

    An updated version of the 2D multifluid code TECXY is used to study the radial electric field structure and the appearance of a global circulation layer (GCL) inside the separatrix of the limiter tokamaks TEXTOR-94 and Tore-Supra-CIEL. The dependence of the driving forces on device geometry, limiter position, magnetic field orientation, impurity content and other parameters is investigated. The centrifugal force in the vicinity of the limiter head always determines the direction of the poloidal velocity in the GCL. There is good agreement with experimentally measured profiles of the poloidal velocity at the TEXTOR low field side. (orig.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. High density operation on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Xiang Gao

    2000-01-01

    The structure of the operation region has been studied in the HT-7 superconducting tokamak, and progress on the extension of the HT-7 ohmic discharge operation region is reported. A density corresponding to 1.2 times the Greenwald limit was achieved by RF boronization. The density limit appears to be connected to the impurity content and the edge parameters, so the best results are obtained with very clean plasmas and peaked electron density profiles. The peaking factors of electron density profiles for different current and line averaged densities were observed. The density behaviour and the fuelling efficiency for gas puffing (20-30%), pellet injection (70-80%) and molecular beam injection (40-50%) were studied. The core crash sawteeth and MHD behaviour, which were induced by an injected pellet, were observed and the events correlated with the change of current profile and reversed magnetic shear. The MARFE phenomena on HT-7 are summarized. The best correlation has been found between the total input ohmic power and the product of the edge line averaged density and Z eff . HT-7 could be easily operated in the high density region MARFE-free using RF boronization. (author)

  3. HIGH PERFORMANCE STATIONARY DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Ferron, J.R.; Politzer, P.A.; Hyatt, A.W.; Sips, A.C.C.; Murakami, M.

    2003-01-01

    Recent experiments in the DIII-D tokamak [J.L. Luxon, Nucl. Fusion 42,614 (2002)] have demonstrated high β with good confinement quality under stationary conditions. Two classes of stationary discharges are observed--low q 95 discharges with sawteeth and higher q 95 without sawteeth. The discharges are deemed stationary when the plasma conditions are maintained for times greater than the current profile relaxation time. In both cases the normalized fusion performance (β N H 89P /q 95 2 ) reaches or exceeds the value of this parameter projected for Q fus = 10 in the International Thermonuclear Experimental Reactor (ITER) design [R. Aymar, et al., Plasma Phys. Control. Fusion 44, 519 (2002)]. The presence of sawteeth reduces the maximum achievable normalized β, while confinement quality (confinement time relative to scalings) is largely independent of q 95 . Even with the reduced β limit, the normalized fusion performance maximizes at the lowest q 95 . Projections to burning plasma conditions are discussed, including the methodology of the projection and the key physics issues which still require investigation

  4. ICRF wave propagation and absorption in tokamak and mirror magnetic fields: a full-wave calculation

    International Nuclear Information System (INIS)

    Jaeger, E.F.; Batchelor, D.B.; Weitzner, H.; Whealton, J.H.

    1985-01-01

    Global solutions for the ion cyclotron resonant frequency (ICRF) wave fields in a straight tokamak with rotational transform and a poloidally symmetric mirror are calculated in the cold plasma limit. The component of the wave electric field parallel to vector Bis assumed zero. Symmetry in each problem allows Fourier decomposition in one ignorable coordinate, and the remaining set of two coupled, two-dimensional partial differential equations is solved by finite differencing. Energy absorption and antenna impedance are calculated using a simple collisional absorption model. When large gradients in vertical barBvertical bar along vectorB are present in either geometry, ICRF heating at the fundamental ion cyclotron resonance is observed. For the mirror, such gradients are always present. But for the tokamak, the rotational transform must be large enough that vectorB . delB greater than or equal to 0(1). For smaller transforms more typical of real tokamaks, only heating at the two-ion hybird resonance is observed. This suggests that direct resonant absorption at the fundamental ion cyclotron resonance may be possible in stellarators where vectorB . delB approx. 0(1) + 11

  5. ICRF wave propagation and absorption in tokamak and mirror magnetic fields: a full-wave calculation

    International Nuclear Information System (INIS)

    Jaeger, E.F.; Batchelor, D.B.; Weitzner, H.; Whealton, J.H.

    1986-01-01

    Global solutions for the ion cyclotron resonant frequency (ICRF) wave fields in a straight tokamak with rotational transform and in a poloidally symmetric mirror are calculated in the cold plasma limit. The component of the wave electric field parallel to B vector is assumed zero. Symmetry in each problem allows Fourier decomposition in one ignorable coordinate, and the remaining set of two coupled, two-dimensional partial differential equations is solved by finite differencing. Energy absorption and antenna impedance are calculated using a simple collisional absorption model. When large gradients in absolute value B along B vector are present in either geometry, ICRF heating at the fundamental ion cyclotron resonance is observed. For the mirror, such gradients are always present. But for the tokamak, the rotational transform must be large enough that B vector . delB greater than or equal to 0(1). For smaller transforms more typical of real tokamaks, only heating at the two-ion hybrid resonance is observed. This suggests that direct resonant absorption at the fundamental ion cyclotron resonance may be possible in stellarators where B vector . delB approx. 0(1) naturally. 13 refs., 23 figs

  6. Workshop on High Power ICH Antenna Designs for High Density Tokamaks

    Science.gov (United States)

    Aamodt, R. E.

    1990-02-01

    A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

  7. Workshop on high power ICH antenna designs for high density tokamaks

    International Nuclear Information System (INIS)

    Aamodt, R.E.

    1990-01-01

    A workshop in high power ICH antenna designs for high density tokamaks was held in Boulder, Colorado on January 31 through February 2, 1990. The purposes of the workshop were to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of rf auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made

  8. 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

  9. 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

  10. Field and ray analyses of antenna excitations in ICRF heating of large Tokamaks

    International Nuclear Information System (INIS)

    Bers, A.; Lister, G.; Jacquinot, J.

    1980-09-01

    We present analytical and computational techniques for determining the electromagnetic fields and associated power flow excited by antenna systems external to large Tokamak plasmas. The finite poloidal and toroidal extension of the poloidal antenna current is modeled by a superposition of current sheets placed at a fixed radius outside the plasma. Antennae both with and without a screen between the current sheet and the plasma are considered. The plama is modeled by its cold dielectric tensor and inhomogeneous density and applied magnetic field. For large Tokamak plasmas in which the plasma dimensions are large compared to the antenna, the field excitation problem can be considered approximately in slab geometry. The field solution of this problem which we present, gives the electromagnetic fields excited in the edge plasma by the antennae and includes the effect of the cutoffs which may exist in this region. To proceed further into the plasma we consider a ray tracing analysis. Starting from an equiphase surface of the excited fields in the edge plasma, the group velocity rays can be followed in full toroidal geometry up to the cyclotron singular resonance region where the power is deposited in the particles. Both the amplitude and phase of the fields can be established in the vicinity of the angular surface so that the power deposition profile can be eventually calculated

  11. High-β steady-state advanced tokamak regimes for ITER and FIRE

    International Nuclear Information System (INIS)

    Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

    2005-01-01

    An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J

    1994-05-01

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

  13. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    vice [1], laser-light scattering [2], parametric interaction of microwave ... As we know that, each level of an atom in a weak magnetic field is split into 2В + 1 .... signal detection channel consists of two channels, one is the signal which is the sum ...

  14. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

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

  15. FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, R. H.; Lee, D. K.; Gaffney, P. W.; Rome, J. A.

    1978-06-01

    The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented.

  16. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    International Nuclear Information System (INIS)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-01-01

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

  17. FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

    International Nuclear Information System (INIS)

    Fowler, R.H.; Lee, D.K.; Gaffney, P.W.; Rome, J.A.

    1978-06-01

    The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented

  18. High-Q plasmas in the TFTR tokamak

    International Nuclear Information System (INIS)

    Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.

    1991-01-01

    In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength S n and D--D fusion power gain Q DD are realized in the neutral-beam-fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, S n increases approximately as P 1.8 b . The highest-Q shots are characterized by high T e (up to 12 keV), T i (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad T e profiles, and lower Z eff . Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, Q DD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [n e (0)/left-angle n e right-angle] during the beam pulse. To date, the best fusion results are S n =5x10 16 n/sec, Q DD =1.85x10 -3 , and neutron yield=4.0x10 16 n/pulse, obtained at I p =1.6--1.9 MA and beam energy E b =95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of S n arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with P b

  19. Active feedback stabilization of axisymmetric modes in highly elongated tokamak plasmas

    International Nuclear Information System (INIS)

    Ward, D.J.; Hofmann, F.

    1993-07-01

    Active feedback stabilization of the vertical instability is studied for highly elongated tokamak plasmas (1≤κ≤3), and evaluated in particular for the TCV configuration. It is shown that the feedback can strongly affect the form of the eigenfunction for these highly elongated equilibria, and this can have detrimental effects on the ability of the feedback system to properly detect and stabilize the plasma. A calculation of the vertical displacement that uses poloidal flux measurements, poloidal magnetic field measurements, and corrections for the vessel eddy currents and active feedback currents was found to be effective even in the cases with the worst deformations of the eigenfunction. We also examine how these deformations affect differently shaped equilibria, and it is seen that the magnitude of the deformation of the eigenfunction is strongly function of the plasma elongation. (author) 15 figs., 13 refs

  20. Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Maingi, R.; Sabbagh, S.A.; Soukhanovskii, V.; Stutman, D.

    2003-01-01

    Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime

  1. The impact of the biasing radial electric field on the SOL in a divertor tokamak

    International Nuclear Information System (INIS)

    Rozhansky, V.; Tendler, M.

    1993-01-01

    Strong radial electric field can be induced within the SOL in a divertor tokamak by applying a voltage to divertor plates with respect to the first wall. This biasing scheme results in the strong radial electric field which is much larger than the natural electric field, usually of the order T e /e. Experiments employing this biasing scheme were carried out on the tokamak TdeV. Many interesting effects such as - modifications of the density profile and radial transport of impurities as a function of the polarity and the magnitude of the biasing voltage, the generation of the flux surface average toroidal rotation proportional to the applied voltage, redistribution of the plasma outflow onto divertor plates and so on - were demonstrated to result from the biasing. Furthermore, in contrast to studies carried out employing a different biasing scheme which primarily results in a poloidal electric field, the strong radial electric field impacts more significantly within SOL than the poloidal electric field. Here, we aim to show that the main effects observed experimentally follow from the analysis, provided continuity and momentum balances are employed invoking anomalous viscosity and inertia. (author) 4 refs

  2. Cryogenic aspects of a demountable toroidal field magnet system for tokamak type fusion reactors

    International Nuclear Information System (INIS)

    Hsieh, S.Y.; Powell, J.; Lehner, J.

    1977-01-01

    A new concept for superconducting Toroidal Field (TF) magnet construction is presented. It is termed the ''Demountable Externally Anchored Low Stress'' (DEALS) magnet system. In contrast to continuous wound conventional superconducting coils, each magnet coil is made from several straight coil segments to form a polygon which can be joined and disjoined to improve reactor maintenance accessibility or to replace failed coil segments if necessary. A design example is presented of a DEALS magnet system for a UWMAK II size reactor. The overall magnet system is described, followed by a detailed analysis of the major heat loads in order to assess the refrigeration requirements for the concept. Despite the increased heat loads caused by high current power leads (200,000 amps) and the coil warm reinforcement support system, the analysis shows that at most, only about one percent (approximately 20 Mw) of the plant electrical output (approximately 2,000 Mw) is needed to operate the magnet cryogenic system. The advantages and the drawbacks of the DEALS magnet system are also discussed. The advantages include: capability to replace failed coils, increased accessibility to the blanket shield assembly, reduced reliability requirements for the magnet, much lower stress in conductor, easier application of improved high field brittle superconductors like Nb 3 Sn, improved magnet safety features, etc. The drawbacks are the increased refrigeration requirements and the necessity of a movable coil support system. A comparison with a conventional magnet system is made. It is concluded that the benefits of the DEALS approach far outweigh its penalties, and that the DEALS concept is the most practical, economical way to construct TF magnet systems for Tokamak reactors

  3. Singular surfaces in the open field line region of a diverted tokamak

    International Nuclear Information System (INIS)

    Reiman, A.

    1996-01-01

    The structure of the open field lines of a slightly nonaxisymmetric, poloidally diverted tokamak is explored by numerical integration of the field line equations for a simple model field. In practice, the nonaxisymmetry could be produced self-consistently by the nonlinear evolution of a free-boundary magnetohydrodynamic (MHD) mode, or it could be produced by field errors, or it could be imposed externally by design. In the presence of a nonaxisymmetric perturbation, the tokamak is shown to develop open field line regions of differing topology separated by singular surfaces. It is argued that the singular surfaces can be expected to play a role analogous to that of rational toroidal flux surfaces, in terms of constraining ideal MHD perturbations and thus constraining the free-energy that can be tapped by ideal MHD instabilities. The possibility of active control of free-boundary instabilities by means of currents driven on the open singular surfaces, which are directly accessible from the divertor plates, is discussed. copyright 1996 American Institute of Physics

  4. Singular surfaces in the open field line region of a diverted tokamak

    International Nuclear Information System (INIS)

    Reiman, A.

    1995-05-01

    The structure of the open field lines of a slightly nonaxisymmetric, poloidally diverted tokamak is explored by numerical integration of the field line equations for a simple model field. In practice, the nonaxisymmetry could be produced self-consistently by the nonlinear evolution of a free-boundary MHD mode, or it could be produced by field errors, or it could be imposed externally by design. In the presence of a nonaxisymmetric perturbation, the tokamak is shown to develop open field line regions of differing topology separated by singular surfaces. It is argued that the singular surfaces can be expected to play a role analogous to that of rational toroidal flux surfaces, in terms of constraining ideal MHD perturbations and thus constraining the free-energy that can be tapped by ideal MHD instabilities. The possibility of active control of free-boundary instabilities by means of currents driven on the open singular surfaces, which are directly accessible from the divertor plates, is discussed. Also discussed is the possibility of early detection of imminent disruptions through localized measurement of the singular surface currents

  5. Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

    Science.gov (United States)

    Ahmadi, S.; Salar Elahi, A.; Ghorannevis, M.

    2018-03-01

    In this paper we have studied an Ergodic Magnetic Limiter (EML) based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist) Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

  6. Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

    Directory of Open Access Journals (Sweden)

    S. Ahmadi

    2018-03-01

    Full Text Available In this paper we have studied an Ergodic Magnetic Limiter (EML based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

  7. Experimental study of external kink instabilities in the Columbia High Beta Tokamak

    International Nuclear Information System (INIS)

    Ivers, T.H.

    1991-01-01

    The generation of power through controlled thermonuclear fusion reactions in a magnetically confined plasma holds promise as a means of supplying mankind's future energy needs. The device most technologically advanced in pursuit of this goal is the tokamak, a machine in which a current-carrying toroidal plasma is thermally isolated from its surroundings by a strong magnetic field. To be viable, the tokamak reactor must produce a sufficiently large amount of power relative to that needed to sustain the fusion reactions. Plasma instabilities may severely limit this possibility. In this work, I describe experimental measurements of the magnetic structure of large-scale, rapidly-growing instabilities that occur in a tokamak when the current or pressure of the plasma exceeds a critical value relative to the magnetic field, and I compare these measurements with theoretical predictions

  8. Improvement of the density limit with an external helical field on JFT-2M tokamak

    International Nuclear Information System (INIS)

    Tamai, H.; Shoji, T.; Nagashima, K.; Miura, Y.; Yamauchi, T.; Ogawa, H.; Kawashima, H.; Matsuda, T.; Mori, M.; Ida, K.; Ohdachi, S.

    1995-01-01

    The density limit is increased by the application of an external helical field in the JFT-2M tokamak. The effect of the magnetic stochasticity due to the external field is investigated to study the mechanism of the improved density limit related to the edge plasma behaviour. The improvement is correlated with the retardation of the increase in the plasma inductance. At the improved density limit, local radiation loss is modified by the helical field, in which that from the vicinity of separatrix X-point is remarkably reduced, while that from outboard edge is slightly increased. The formation of a positive radial electric field at the plasma edge is also observed in the presence of the helical field. ((orig.))

  9. On the energy confinement in the TM-G tokamak with high plasma density

    International Nuclear Information System (INIS)

    Stefanovskij, A.M.

    1986-01-01

    Energy confinement time τ E , when plasma density changing, has been measured at the TM-G-tokamak device with a graphite discharge chamber. The measurements have been carried out in three different discharge modes with a similar stability margin on the limiter (q L )=3) and with different values of the discharge current of a longitudinal field (I p =20, 40 and 60 kA, V T =0.8; 1.6 and 2.4 T). On the basis of experimental data analysis the conclusion is made that saturation of τ E (n e ) dependence at high plasma density occurs due to current channel compression and violation of a ''self-consistent'' profile of current density. Drift wave excitation at densities similar to the limiting Murakami density can also play an important role

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  11. Tokamak poloidal field systems. Progress report, January 1-December 31, 1979

    International Nuclear Information System (INIS)

    Rogers, J.D.

    1980-05-01

    Work is reported on the development of superconducting tokamak poloidal field systems (TPFS). Progress is discussed on the design of a 20 MJ, 50 kA, 7.5 T superconducting pulsed energy storage coil operated in a 1 to 2 s bipolar mode from +7.5 T to -7.5 T in 1982. Conductor development for the coil is presented. A facility that uses a traction motor energy transfer system to test coils in the 20 to 100 MJ energy range is discussed. Current interrupter development and testing for protection and energy transfer circuits are also presented. The 400 kJ METS coil test preparation is under way

  12. Temperature anisotropy in a cyclotron resonance heated tokamak plasma and the generation of poloidal electric field

    International Nuclear Information System (INIS)

    Choe, W.; Ono, M.; Chang, C.S.

    1994-11-01

    The temperature anisotropy generated by cyclotron resonance heating of tokamak plasmas is calculated and the poloidal equilibrium electric field due to the anisotropy is studied. For the calculation of anisotropic temperatures, bounce-averaged Fokker-Planck equation with a bi-Maxwellian distribution function of heated particles is solved, assuming a moderate wave power and a constant quasilinear cyclotron resonance diffusion coefficient. The poloidal electrostatic potential variation is found to be proportional to the particle density and the degree of temperature anisotropy of warm species created by cyclotron resonance heating

  13. Progress in application of high temperature superconductor in tokamak magnets

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.; Svoboda, V.; Stöckel, Jan; Sykes, A.; Sykes, N.; Kingham, D.; Hammond, G.; Apte, P.; Todd, T.N.; Ball, S.; Chappell, S.; Melhem, D.; Ďuran, Ivan; Kovařík, Karel; Grover, O.; Markovič, T.; Odstrčil, M.; Odstrčil, T.; Šindlery, A.; Vondrášek, G.; Kocman, J.; Lilley, M.K.; de Grouchy, P.; Kim, H.-T.

    2013-01-01

    Roč. 88, 9-10 (2013), s. 1593-1596 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] Institutional support: RVO:61389021 Keywords : tokamaks * HTS * magnet s Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001117#

  14. Configuration development of a hydraulic press for preloading the toroidal field coils of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

    The Fusion Engineering Design Center (FEDC) is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak (CIT). To achieve a compact device with the minimum major radius, a vertical preload system is being developed to react the vertical separating force normally carried by the inboard leg of the toroidal field (TF) coils. The preload system is in the form of a hydraulic press. Challenges in the design include the development of hydraulic and structural systems for very large force requirements, which could interface with the CIT machine, while allowing maximum access to the top, bottom, and radial periphery of the machine. Maximum access is necessary for maintenance, diagnostics, instrumentation, and control systems. Materials used in the design must function in the nuclear environment and in the presence of high magnetic fields. This paper presents the configuration development of the hydraulic press used to vertically preload the CIT device

  15. Generalized MHD for numerical stability analysis of high-performance plasmas in tokamaks

    International Nuclear Information System (INIS)

    Mikhailovskii, A.B.

    1998-01-01

    A set of generalized magnetohydrodynamic (MHD) equations is formulated to accommodate the effects associated with high ion and electron temperatures in high-performance plasmas in tokamaks. The effects of neoclassical bootstrap current, neoclassical ion viscosity, the ion finite Larmor radius effect and electron and ion drift effects are taken into account in two-fluid MHD equations together with gyroviscosity, parallel viscosity, electron parallel inertia and collisionless ion heat flux. The ion velocity is identified as the plasma velocity, while the electron velocity is expressed in terms of the plasma velocity and electric current. Ion and electron momentum equations are combined to give the plasma momentum equation. The perpendicular (with respect to the equilibrium magnetic field) ion momentum equation is used as perpendicular Ohm's law and the parallel electron momentum equation - as parallel Ohm's law. Perpendicular Ohm's law allows for the Hall and ion drift effects. Parallel Ohm's law includes the electron drift effect, collisionless skin effect and bootstrap current. In addition, both perpendicular and parallel Ohm's laws contain the resistivity. Due to the quasineutrality condition, the ions and electrons are characterized by the same number density which is described by the ion continuity equation. On the other hand, the ion and electron temperatures are allowed to be different. The ion temperature is described by the ion energy equation allowing for the oblique heat flux, in addition to the perpendicular ion heat flux. The electron temperature is determined by the condition of high parallel electron heat conductivity. The ion and electron parallel viscosities are represented in a form valid for all the collisionality regimes (Pfirsch-Schluter, plateau, and banana). An optimized form of the generalized MHD equations is then represented in terms of the toroidal coordinate system used in the JET equilibrium and stability codes. The derived equations

  16. Influence of external resonant magnetic perturbation field on edge plasma of small tokamak HYBTOK-II

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Y., E-mail: hayashi-yuki13@ees.nagoya-u.ac.jp [Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Suzuki, Y.; Ohno, N. [Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Okamoto, M. [Ishikawa National College of Technology, Kitachujo, Tsubata-cho, Kahoku-gun, Ishikawa 929-0392 (Japan); Kikuchi, Y. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Sakakibara, S.; Watanabe, K.; Takemura, Y. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292 (Japan)

    2015-08-15

    Radial profile of externally applied resonant magnetic perturbation (RMP) field with mode numbers of m = 6 and n = 2 in a small tokamak device HYBTOK-II have been investigated using a magnetic probe array, which is able to measure the radial profile of magnetic field perturbation induced by applying RMP. Results of RMP penetration into the plasma show that the RMP decreased toward the plasma center, while they were amplified around the resonant surface with a safety factor q = 3 due to the formation of magnetic islands. This suggests that RMP fields for controlling edge plasmas may trigger some kind of MHD instabilities. In addition, simulation results, based on a linearized four-field model, which agrees with the experimental ones, indicates that the penetration and amplification process of RMP strongly depend on a Doppler-shifted frequency between the RMP and plasma rotation.

  17. Passing particle toroidal precession induced by electric field in a tokamak

    International Nuclear Information System (INIS)

    Andreev, V. V.; Ilgisonis, V. I.; Sorokina, E. A.

    2013-01-01

    Characteristics of a rotation of passing particles in a tokamak with radial electric field are calculated. The expression for time-averaged toroidal velocity of the passing particle induced by the electric field is derived. The electric-field-induced additive to the toroidal velocity of the passing particle appears to be much smaller than the velocity of the electric drift calculated for the poloidal magnetic field typical for the trapped particle. This quantity can even have the different sign depending on the azimuthal position of the particle starting point. The unified approach for the calculation of the bounce period and of the time-averaged toroidal velocity of both trapped and passing particles in the whole volume of plasma column is presented. The results are obtained analytically and are confirmed by 3D numerical calculations of the trajectories of charged particles

  18. Passing particle toroidal precession induced by electric field in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Andreev, V. V. [Peoples' Friendship University of Russia, Ordzhonikidze St. 3, Moscow 117198 (Russian Federation); Ilgisonis, V. I.; Sorokina, E. A. [Peoples' Friendship University of Russia, Ordzhonikidze St. 3, Moscow 117198 (Russian Federation); NRC “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

    2013-12-15

    Characteristics of a rotation of passing particles in a tokamak with radial electric field are calculated. The expression for time-averaged toroidal velocity of the passing particle induced by the electric field is derived. The electric-field-induced additive to the toroidal velocity of the passing particle appears to be much smaller than the velocity of the electric drift calculated for the poloidal magnetic field typical for the trapped particle. This quantity can even have the different sign depending on the azimuthal position of the particle starting point. The unified approach for the calculation of the bounce period and of the time-averaged toroidal velocity of both trapped and passing particles in the whole volume of plasma column is presented. The results are obtained analytically and are confirmed by 3D numerical calculations of the trajectories of charged particles.

  19. Structure of magnetic field disturbances under development of disruptive instability in the ''Tokamak-6''

    International Nuclear Information System (INIS)

    Merezhkin, V.G.

    1978-01-01

    The structure and dynamics of disturbances of a poloidal field during the development of the breakaway instability in the Tokamak-6 are investigated. The behaviour of the symmetric and dipole field component, and the peculiarities of the structure of screw disturbances in a minor and major breakaways are analyzed. It was established that the structure of screw disturbances in minor breakaways is unchangeable and that the rearrangement in major breakaways is of a discrete nature. The relationship between the symmetric and screw components of disturbances of the poloidal field at the forward front of the disturbance increase was revealed. Data on the increments, scales and structure of screw disturbances, the ratios between the symmetric and screw components of field disturbances, and also on the magnitude of energy losses in typical breakaways are given

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

    Science.gov (United States)

    Ono, Masayuki; Furth, Harold

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Soltwisch, H.; Koslowski, H.R.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, D; Machida, M; Scalabrin, A

    1986-03-01

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

  3. Recent progresses on high performance steady-state plasmas in the superconducting tokamak TRIAM-1M

    International Nuclear Information System (INIS)

    Itoh, Satoshi; Sato, Kohnosuke; Nakamura, Kazuo

    1999-01-01

    The overview of TRIAM-1M experiments is described. The up-to-date issues for steady-state operation are presented through the experience of the achievement of super ultra long tokamak discharges (SULD) sustained by lower hybrid current drive (LHCD) over 2 hours. The importance of the control of an initial phase of plasma, the avoidance of the concentration of huge heat load, the wall conditioning, and abrupt stop of the long discharges are proposed as the indispensable issues for the achievement of the steady-state operation of tokamak. A high ion temperature (HIT) discharge fully sustained by 2.45 GHz LHCD with both high ion temperature and steep temperature gradient is successfully demonstrated for longer than 1 min in the limiter configuration. The HIT discharges can be obtained in the narrow window of density and position. Moreover, the avoidance of the concentration of heat load on a limiter is the key point for the achievement and its long sustainment. As the effective thermal insulation between the wall and the plasma is improved on the single null configuration, HIT discharges with peak ion temperature > 5keV and steeper gradient up to 85 keV/m can be achieved by the exquisite control of density and position. The plasmas with high κ ∼1.5 can be also demonstrated for longer than 1 min. The current profile is also well-controlled for about 2 orders in magnitude longer than the current diffusion time using combined LHCD. The serious damage to the material of the first wall caused by energetic neutral particles produced via charge exchange process is also described. As the neutral particles cannot be affected by magnetic field, this damage by neutral particles must be avoided by the new technique. (author)

  4. Tritium system for compact high field devices

    International Nuclear Information System (INIS)

    Roccella, M.; Bonizzoni, G.; Chiesa, P.; Ghezzi, F.; Nassi, M.; Pavesi, U.; Amedeo, P.; Boschetti, G.; Giffanti, F.; Moriggio, A.

    1988-01-01

    Some theoretical results and the current status of the work on a prototype plant for the Tritium cycle of compact high-field tokamaks (such as, Ignitor, CIT, etc.), using the SAES Getter St 707 getter material, are described in this report. The schematics and present status of the main subplants of the cycle are reported together with some experimental results demostrating the possibility of utilizing the St 707 material to purify the inert atmosphere of the glove-boxes and the secondary containment of the double-containment metal canalization which is to eventually house the various parts of the plant. Finally, as an example, the FTU machine, under construction at ENEA Frascati, has been taken as a reference, and theoretical evaluations are given for the inventory, permeation and release of the Tritium from the first wall and the thermal shieldes of such a tokamak

  5. Use of field ripple for burn phase control in short-pulse tokamak reactors

    International Nuclear Information System (INIS)

    Davidson, J.N.; Stacey, W.M. Jr.

    1979-05-01

    The possible use of toroidal field ripple to control temperature excursions was investigated. In particular, the magnitude of the power savings necessary to control the ion temperature for about 120 seconds during the burn phase of an ETF tokamak was determined through the use of a global dynamics code. Then the amount of field ripple necessary to effect that control was calculated for each case. In the next section the basic properties of the simulation code are given and the various plasma cases investigated are described in some detail. There follows a discussion of the field ripple control mechanisms and their application to the problem at hand. Finally, the results of the study are presented along with the conclusions drawn from them

  6. Toroidal electric field in front of the lower hybrid grill of the castor tokamak

    International Nuclear Information System (INIS)

    Zacek, F.; Petrzilka, V.; Devynck, P.; Goniche, M.

    2003-01-01

    A small tokamak Castor (R/a = 0.4/0.85 m) with low plasma energy density and short pulses (20 ms) offers a unique possibility to carry out probe measurements in front of the grill antenna and as a consequence to provide direct information about the local electric fields in this region. For measurements of the toroidal electrical field, a small double probe with 2 tips separated by 3.5 mm in the toroidal direction has been used. The tips are oriented in the radial direction. The probe is radially movable in front of the central grill waveguide. Cross-correlations and FFT (fast Fourier transform) analysis of the measured V fl signals are given together with an attempt to investigate characteristics of toroidal electric field E tor (up to 500 kHz), derived from V fl measured by 2 toroidally separated tips

  7. Tokamak poloidal-field systems. Progress report, January 1-December 31, 1981

    International Nuclear Information System (INIS)

    Rogers, J.D.

    1982-03-01

    Work on the superconducting tokamak poloidal field system (TPFS) program is being redirected. The development of the 20 MJ, 50 kA, 7.5 T superconducting programmed energy storage coil is being terminated. The superconductor for the 20 MJ coil is being processed only to an intermediate state, and manufacture of the epoxy fiberglass dewar is being stopped. Further, development of the TPFS test facility is in abeyance. Change in program emphasis arises from prospective rf plasma current driven or beam heated tokamaks with programmed coil characteristics for the poloidal field being different from those to have been simulated by the 20 MJ coil and from budgetary constraints. Work is reported on the development of the coil, conductor, nonconducting dewar, and test facility to the recent time when the program change was instigated. Work in support of the Large Coil Test Facility (LCTF) and the Fusion Engineering Design (FED) Center is given. Analysis of the experiments on the 400 kJ METS coil test was completed

  8. Mechanical design of the coils encapsulated of toroidal field of Tokamak TPM1

    International Nuclear Information System (INIS)

    Caldino H, U.; Francois L, J. L.

    2014-10-01

    The TPM1 is a small Tokamak that belongs to the Centro de Investigacion en Ciencias Aplicadas y Tecnologia Avanzada of Instituto Politecnico Nacional (CICATA-IPN); the project is under construction. Currently it has the vacuum chamber, and is intended that the machine can operate with electric pulses of 10 ms to study the behavior of plasmas in order to provide knowledge in the field of nuclear fusion by magnetic confinement. To achieve this goal is necessary to design the toroidal field coils which operate the Tokamak. This paper presents an analysis which was performed to obtain the correct configuration of coils depending on design parameters for operation of the machine. Once determined this configuration, an analysis of electromagnetic forces present in normal machine operation on one coil was conducted, this to know the stresses in the encapsulation of the same. Considering the pulsed operation, a thickness of 5 mm is determined in the encapsulated, considering fatigue failure based on studies of fatigue failures in epoxy resins. (Author)

  9. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Ferro, C.; Zanino, R.

    1992-01-01

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

  11. Nonlinear error-field penetration in low density ohmically heated tokamak plasmas

    International Nuclear Information System (INIS)

    Fitzpatrick, R

    2012-01-01

    A theory is developed to predict the error-field penetration threshold in low density, ohmically heated, tokamak plasmas. The novel feature of the theory is that the response of the plasma in the vicinity of the resonant surface to the applied error-field is calculated from nonlinear drift-MHD (magnetohydrodynamical) magnetic island theory, rather than linear layer theory. Error-field penetration, and subsequent locked mode formation, is triggered once the destabilizing effect of the resonant harmonic of the error-field overcomes the stabilizing effect of the ion polarization current (caused by the propagation of the error-field-induced island chain in the local ion fluid frame). The predicted scaling of the error-field penetration threshold with engineering parameters is (b r /B T ) crit ∼n e B T -1.8 R 0 -0.25 , where b r is the resonant harmonic of the vacuum radial error-field at the resonant surface, B T the toroidal magnetic field-strength, n e the electron number density at the resonant surface and R 0 the major radius of the plasma. This scaling—in particular, the linear dependence of the threshold with density—is consistent with experimental observations. When the scaling is used to extrapolate from JET to ITER, the predicted ITER error-field penetration threshold is (b r /B T ) crit ∼ 5 × 10 −5 , which just lies within the expected capabilities of the ITER error-field correction system. (paper)

  12. A symplectic map for trajectories of magnetic field lines in double-null divertor tokamaks

    Science.gov (United States)

    Crank, Willie; Ali, Halima; Punjabi, Alkesh

    2009-11-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in tokamaks can be any coordinates for which a transformation to (ψ,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψ is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct a map that represents the magnetic topology of double-null divertor tokamaks. For this purpose, the generating function of the simple map [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)] is slightly modified. The resulting map equations for the double-null divertor tokamaks are: x1=x0-ky0(1-y0^2 ), y1=y0+kx1. k is the map parameter. It represents the generic topological effects of toroidal asymmetries. The O-point is at (0.0). The X-points are at (0,±1). The equilibrium magnetic surfaces are calculated. These surfaces are symmetric about the x- and y- axes. The widths of stochastic layer near the X-points in the principal plane, and the fractal dimensions of the magnetic footprints on the inboard and outboard side of upper and lower X-points are calculated from the map. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

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

    DEFF Research Database (Denmark)

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

    2009-01-01

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

  14. Radial electric field at the plasma edge on the FT-2 Tokamak in regimes with large gradients

    International Nuclear Information System (INIS)

    Lashkul, S.; Popov, A.

    2001-01-01

    The transport barrier formation is widely believed to be the fundamental element of transition into improved confinement regimes (H-mode). Experiments on many tokamaks demonstrate that transport barrier formation is connected with the suppression of turbulent transport by shear of E x B drift. Therefore, the calculation of radial electric field is of great importance. Our work is devoted to progress the neoclassical theory by taking into account electron viscosity and non-linear effects (ion inertia), presented results being valuable for interpretation transition into H-mode at the plasma edge in small tokamaks. Calculations of the electric field profile for FT-2 tokamak (a=8cm, R 0 =55cm, Ioffe Institute, Russia) according found expressions are in the good agreement with experimental results obtained. (orig.)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  17. Modifying locally the safety profile to improve the confinement of magnetic field lines in tokamak plasmas

    International Nuclear Information System (INIS)

    Constantinescu, D.; Firpo, M.-C.

    2012-01-01

    Using Hamiltonian models for the magnetic field lines, we propose a methodology to improve their confinement through the creation of transport barriers. A local modification of the safety profile creating a low-shear zone is shown to be sufficient to locally enhance drastically the regularity of the magnetic field lines without requesting a reversed shear. The optimal benefits of low-shear are obtained when the value q 0 of the safety profile in the low-shear zone is sufficiently far from the main resonance values m/n with low m and n, in the case of large enough values of those (m, n) mode amplitudes. A practical implementation in tokamak plasmas should involve electron cyclotron current drive to locally modify the magnetic shear. (paper)

  18. Observation of disruptions in tokamak plasma under the influence of resonant helical magnetic fields

    International Nuclear Information System (INIS)

    Araujo, M.; Vannucci, A.; Caldas, I.

    1996-01-01

    Disruptive instabilities were investigated in the small tokamak TBR-1 during the application of resonant helical magnetic fields created by external helical windings. Indications were found that the main triggering mechanism of the disruptions was the rapid increase of the m=2/n=1 mode which, apparently after reaching a certain amplitude, interacts with other resistive modes: the internal 1/1 mode in the case of minor disruptions. After the coupling, the growth of the associated islands would create a chaotic field line distribution in the region between the corresponding rational magnetic surfaces which caused the gross particle transport and, finally, destroyed the confinement. In addition, investigations on higher Z eff discharges in which a mixture of helium and hydrogen was used resulted in much more unstable plasmas but apparently did not alter basic characteristics of the disruptions

  19. The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

    Science.gov (United States)

    Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

    2009-04-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψt,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψt is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m =1, n =±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are

  20. The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

    International Nuclear Information System (INIS)

    Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

    2009-01-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψ t ,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψ t is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m=1, n=±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are

  1. Observations of toroidal and poloidal rotation in the high beta tokamak Torus II

    International Nuclear Information System (INIS)

    Kostek, C.A.

    1983-01-01

    The macroscopic rotation of plasma in a toroidal containment device is an important feature of the equilibrium. Toroidal and poloidal rotation in the high beta tokamak Torus II is measured experimentally by examining the Doppler shift of the 4685.75 A He II line emitted from the plasma. The toroidal flow at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed, moves in the same direction as the toroidal plasma current. The poloidal flow follows the ion diamagnetic current direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with predictions from neoclassical theory in the collosional, Pfirsch-Schluter regime. The poloidal motion, however results from an E x B drift in a positive radial electric field, approaching a stable ambipolar state. This radial electric field is determined from theory by using the measured poloidal velocity. Mechanisms for the time evolution of rotation are also examined. It appears that the circulation damping is governed by a global decay of the temperature and density gradients which, in turn, may be functions of radiative cooling, loss of equilibrium due to external field decay, or the emergence of a growing instability, occasionally observed in CO 2 interferometry measurements

  2. Scaling of energy confinement and poloidal beta in high density tokamaks

    NARCIS (Netherlands)

    Schram, D.C.; Schüller, F.C.

    1980-01-01

    A semi-empirical analysis of the heat balance of ohmically heated, high density Tokamak plasmas, shows that the observed heat transport can be explained by neoclassical (plateau) ion heat conduction in the central part of the plasma. Experimental values for Te, ß¿e, and tEe and the variation of

  3. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak.

    Science.gov (United States)

    Truong, D D; Austin, M E

    2014-11-01

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of Te(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83-130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1-3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6-0.8 cm) resolution Te measurements. The high resolution subsystem branches off from the regular channels' IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2-4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters' center frequencies (250 MHz). This configuration allows for full coverage of the 83-130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a "zoomed-in" analysis of a ∼2-4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial Te measurements, which demonstrate that the desired resolution is achieved, are presented.

  4. Experimental and calculating study on the stressed state of superconducting coils of toroidal field in the T-15 tokamak

    International Nuclear Information System (INIS)

    Vaulina, I.G.; Gusev, S.V.; Sivkova, G.N.

    1987-01-01

    Results of calculational and experimental atudy of stress-deformed state of superconducting coils of the T-15 tokamak toroidal field are presented. The calculations are made using the method of finite elements and refined theory of cores. Experimental studies were carried out using elastic tensometric model of polymer materials. Test results are compared with the calculational results. Divergence between calculational and experimental values of displacement of characteristic points in the unit does not exceed 20 %. Results of model studies confirm the expediency of the calculational model used for designing SOTP unit for the T-15 tokamak

  5. Stochastization of Magnetic Field Surfaces in Tokamaks by an Inner Coil

    International Nuclear Information System (INIS)

    Chavez-Alarcon, Esteban; Herrera-Velazquez, J. Julio E.; Braun-Gitler, Eliezer

    2006-01-01

    A 3-D code has been developed in order to simulate the magnetic field lines in circular cross-section tokamaks. The toroidal magnetic field can be obtained from the individual fields of circular coils arranged around the torus, or alternatively, as a ripple-less field. The poloidal field is provided by a given toroidal current density profile. Proposing initial conditions for a magnetic filed line, it is integrated along the toroidal angle coordinate, and Poincare maps can be obtained at any desired cross section plane. Following this procedure, the code allows the mapping of magnetic field surfaces for the axisymmetric case. For this work, the density current profile is chosen to be bell-shaped, so that realistic safety factor profiles can be obtained. This code is used in order to study the braking up of external surfaces when the symmetry is broken by an inner coil with tilted circular loops, with the purpose of modelling the behaviour of ergodic divertors, such as those devised for TEXTOR

  6. 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

  7. Steady-state resistive toroidal-field coils for tokamak reactors

    International Nuclear Information System (INIS)

    Kalnavarns, J.; Jassby, D.L.

    1979-12-01

    If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

  8. Tokamak poloidal field systems. Progress report, January 1-December 31, 1980

    International Nuclear Information System (INIS)

    Rogers, J.D.

    1981-03-01

    Work is reported on the development of superconducting tokamak poloidal field system (TPFS) program. Progress is discussed on the design of the 20 MJ, 50 kA, 7.5 T superconducting pulsed energy storage coil to be operated in a bipolar mode from +7.5 T to -7.5 T in an energy transfer period of 1.5 to 5 s in 1982 followed by extensive cyclic testing. The facility to conduct the tests uses a traction motor energy transfer system and a nonconducting dewar. Status of the hardware development for the TPFS program is presented. Current interrupter development and testing for protection and energy transfer circuits are also presented. The 400 kJ METS coil test results are given

  9. Elastic-plastic analysis of the toroidal field coil inner leg of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Horie, T.

    1987-07-01

    Elastic-plastic analyses were made for the inner leg of the Compact Ignition Tokamak toroidal field (TF) coil, which is made of copper-Inconel composite material. From the result of the elastic-plastic analysis, the effective Young's moduli of the inner leg were determined by the analytical equations. These Young's moduli are useful for the three-dimensional, elastic, overall TF coil analysis. Comparison among the results of the baseline design (R = 1.324 m), the bucked pressless design, the 1.527-m major radius design, and the 1.6-m major radius design was also made, based on the elastic-plastic TF coil inner leg analyses

  10. Design of next step tokamak: Consistent analysis of plasma performance flux composition and poloidal field system

    International Nuclear Information System (INIS)

    Ane, J.M.; Grandgirard, V.; Albajar, F.; Johner, J.

    2001-01-01

    A consistent and simple approach to derive plasma scenarios for next step tokamak design is presented. It is based on successive plasma equilibria snapshots from plasma breakdown to end of ramp-down. Temperature and density profiles for each equilibrium are derived from a 2D plasma model. The time interval between two successive equilibria is then computed from the toroidal field magnetic energy balance, the resistive term of which depends on n, T profiles. This approach provides a consistent analysis of plasma performance, flux consumption and PF system, including average voltages waveforms across the PF coils. The plasma model and the Poynting theorem for the toroidal magnetic energy are presented. Application to ITER-FEAT and to M2, a Q=5 machine designed at CEA, are shown. (author)

  11. Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

    International Nuclear Information System (INIS)

    Goldston, R.J.; Towner, H.H.

    1980-02-01

    Analytic calculations of three important effects of toroidal field ripple on suprathermal ions in tokamak plasmas are presented. In the first process, collisional ripple-trapping, beam ions become trapped in local magnetic wells near their banana tips due to pitch-angle scattering as they traverse the ripple on barely unripple-trapped orbits. In the second process, collisionless ripple-trapping, near-perpendicular untrapped ions are captured (again near a banana tip) due to their finite orbits, which carry them out into regions of higher ripple. In the third process, banana-drift diffusion, fast-ion banana orbits fail to close precisely, due to a ripple-induced variable lingering period near the banana tips. These three mechanisms lead to substantial radial transport of banana-trapped, neutral-beam-injected ions when the quantity α* identical with epsilon/sin theta/Nqdelta is of order unity or smaller

  12. Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

    International Nuclear Information System (INIS)

    Goldston, R.J.; Towner, H.H.

    1981-01-01

    Analytic calculations of three important effects of toroidal field ripple on suprathermal ions in tokamak plasmas are presented. In the first process, collisional ripple-trapping, ions become trapped in local magnetic wells near their banana tips owing to pitch-angle scattering as they traverse the ripple on barely unripple-trapped orbits. In the second process, collisionless ripple-trapping, ions are captured (again near a banana tip) owing to their finite orbits, which carry them out into regions of higher ripple. In the third process, banana-drift diffusion, fast-ion banana orbits fail to close precisely, due to a ripple-induced 'variable lingering period' near the banana tips. These three mechanisms lead to substantial radial transport of banana-trapped, neutral-beam-injected ions when the quantity α* is identical with epsilonsinthetaNqdelta is of order unity or smaller. (author)

  13. Determination of magnetic field direction in tokamaks from laser-induced Lyman-α fluorescence

    International Nuclear Information System (INIS)

    Voslamber, D.

    1988-04-01

    Resonant laser scattering in the Lyman-α line of hydrogen is investigated as a possible tool for measuring the magnetic field direction in tokamak plasmas. The method is based on the depolarisation-dependence of the scattering process. Limitations arising from depolarising collisions are studied in detail by employing a previously developed theory for the collisional redistribution of light. An error analysis is performed to derive the expected experimental precision under various plasma conditions and for laser energies ranging between 1 micronJ and 10 mJ. This analysis also includes the measurement of neutral hydrogen densities. It is shown that with presently available laser powers application of the method would be restricted to the border regions of the plasma. Application to the central regions would require further developments in laser technology, especially with regard to higher powers at the wavelength of Lyman-α and (or) to fast repetition rates

  14. Observations of plasma rotation in the high-beta tokamak Torus II

    International Nuclear Information System (INIS)

    Kostek, C.; Marshall, T.C.

    1982-01-01

    Toroidal and poloidal plasma rotation are measured in a high Beta tokamak device by studying the Doppler shift of the 4686 A He II line. The toroidal flow motion is in the same direction as the plasma current at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed. The poloidal flow follows the ion diamagnetic direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with the predictions of neoclassical transport theory in the collisional regime. For the poloidal motion, however, it appears that an (E/sub r/ x B)/B 2 drift in a positive radial electric field, approaching a stable ambipolar state (STRINGER, 1970) is responsible. Mechanisms for the time evolution of the rotation are also examined. The radial electric field responsible for the (E/sub r/ x B)/B 2 drift is determined from the theory using the measured poloidal velocity

  15. Measurement of tokamak error fields using plasma response and its applicability to ITER

    International Nuclear Information System (INIS)

    Strait, E.J.; Buttery, R.J.; Chu, M.S.; Garofalo, A.M.; La Haye, R.J.; Schaffer, M.J.; Casper, T.A.; Gribov, Y.; Hanson, J.M.; Reimerdes, H.; Volpe, F.A.

    2014-01-01

    The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER's diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible. (paper)

  16. Radial electric field evolution in various operational modes in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Askinazi, L G; Kornev, V A; Krikunov, S V; Lebedev, S V; Smirnov, A I; Tukachinsky, A S; Vildjunas, M I; Zhubr, N A; Krupnik, L I; Tendler, M

    2008-01-01

    Radial electric field evolution has been studied on the TUAMN-3M tokamak in different modes of operation: ohmic and NBI heating, L- and H-modes, with and without strong MHD activity. Peripheral radial electric field was measured using Langmuire probes, which were inserted up to 2cm inside LCFS, while core plasma potential evolution was measured using HIBP diagnostic. It was found, that in presence of strong MHD activity radial electric field in a vicinity of the island changed sign from negative to positive and could reach up to 4kV/m. Central plasma potential exhibited a positive perturbation of ∼700V during the MHD burst. This positive radial electric field might lead to H-mode termination, both in ohmic and NBI heating cases. Possible mechanism of the positive E r generation, namely the electron losses along ergodized magnetic field lines in the presence of MHD-island, is discussed. The same mechanism might be responsible for the positive potential spikes during a saw-tooth crash, also observed using HIBP. Another phenomenon observed using HIBP was quasi-coherent potential oscillations with the frequency close to one of the GAM. Possible location of these oscillations in the core region r/a ∼ 0.33 is discussed

  17. Compact tokamak reactors

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Mechanical design of the coils encapsulated of toroidal field of Tokamak TPM1; Diseno mecanico del encapsulado de las bobinas de campo toroidal del Tokamak TPM1

    Energy Technology Data Exchange (ETDEWEB)

    Caldino H, U.; Francois L, J. L., E-mail: ucaldino@outlook.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    The TPM1 is a small Tokamak that belongs to the Centro de Investigacion en Ciencias Aplicadas y Tecnologia Avanzada of Instituto Politecnico Nacional (CICATA-IPN); the project is under construction. Currently it has the vacuum chamber, and is intended that the machine can operate with electric pulses of 10 ms to study the behavior of plasmas in order to provide knowledge in the field of nuclear fusion by magnetic confinement. To achieve this goal is necessary to design the toroidal field coils which operate the Tokamak. This paper presents an analysis which was performed to obtain the correct configuration of coils depending on design parameters for operation of the machine. Once determined this configuration, an analysis of electromagnetic forces present in normal machine operation on one coil was conducted, this to know the stresses in the encapsulation of the same. Considering the pulsed operation, a thickness of 5 mm is determined in the encapsulated, considering fatigue failure based on studies of fatigue failures in epoxy resins. (Author)

  19. High beta, sawtooth-free tokamak operation using energetic trapped particles

    International Nuclear Information System (INIS)

    White, R.B.; Bussac, M.N.; Romanelli, F.

    1988-08-01

    It is shown that a population of high energy trapped particles, such as that produced by ion cyclotron heating in tokamaks, can result in a plasma completely stable to both sawtooth oscillations and the fishbone mode. The stable window of operation increases in size with plasma temperature and with trapped particle energy, and provides a means of obtaining a stable plasma with high current and high beta. 13 refs., 2 figs

  20. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

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

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  1. Accelerating the numerical simulation of magnetic field lines in tokamaks using the GPU

    International Nuclear Information System (INIS)

    Kalling, R.C.; Evans, T.E.; Orlov, D.M.; Schissel, D.P.; Maingi, R.; Menard, J.E.; Sabbagh, S.A.

    2011-01-01

    Highlights: → Tokamak magnetic field lines are simulated on a GPU. → Numerical integration of a set of nonlinear differential equations is required. → Using the GPU yields a significant reduction in processing time compared to the CPU. → Computational runs that took days now take hours. → These gains have been accomplished without significant hardware expense. - Abstract: TRIP3D is a field line simulation code that numerically integrates a set of nonlinear magnetic field line differential equations. The code is used to study properties of magnetic islands and stochastic or chaotic field line topologies that are important for designing non-axisymmetric magnetic perturbation coils for controlling plasma instabilities in future machines. The code is very computationally intensive and for large runs can take on the order of days to complete on a traditional single CPU. This work describes how the code was converted from Fortran to C and then restructured to take advantage of GPU computing using NVIDIA's CUDA. The reduction in computing time has been dramatic where runs that previously took days now take hours allowing a scale of problem to be examined that would previously not have been attempted. These gains have been accomplished without significant hardware expense. Performance, correctness, code flexibility, and implementation time have been analyzed to gauge the success and applicability of these methods when compared to the traditional multi-CPU approach.

  2. Radial electric field and rotation of the ensemble of plasma particles in tokamak

    International Nuclear Information System (INIS)

    Sorokina, E. A.; Ilgisonis, V. I.

    2012-01-01

    The velocity of macroscopic rotation of an ensemble of charged particles in a tokamak in the presence of an electric field has been calculated in a collisionless approximation. It is shown that the velocity of toroidal rotation does not reduce to a local velocity of electric drift and has opposite directions on the inner and outer sides of the torus. This result is supplemented by an analysis of the trajectories of motion of individual particles in the ensemble, which shows that the passing and trapped particles of the ensemble acquire in the electric field, on the average, different toroidal velocities. For the trapped particles, this velocity is equal to that of electric drift in the poloidal magnetic field, while the velocity of passing particles is significantly different. It is shown that, although the electric-field-induced shift of the boundaries between trapped and passing particles in the phase space depends on the particle mass and charge and is, in the general case, asymmetric, this does not lead to current generation.

  3. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn 3 conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended

  4. Experimental observations of MHD instabilities in the high-beta tokamak Torus-II

    International Nuclear Information System (INIS)

    Machida, M.

    1982-01-01

    The CO 2 laser scattering and interferometry diagnostics have been used to study the MHD instabilities in the high-beta tokamak Torus-II. Detailed measurements of the density and density fluctuation profiles have been performed. In order to measure density fluctuations with wavelengths longer than 2 cm, an interferometric like, phase matching technique has been developed. The toroidal and poloidal mode numbers have been measured using a double-beam, two-position technique. Working at high-beta values, average β greater than or equal to 10%, we have found parameters where the growing instabilities are created or suppressed. The plasma lifetime for both cases is seen to be about the same and the loss of the plasma appears to be caused by the decay in the external fields. The growing instability parameters are within the MHD regime, and it only grows at the outer edge of the plasma. This is in agreement with the theoretical Ballooning mode instability. The frequency and mode number measurements also agree with the Kinetic theory description of Ballooning modes. The comparison with possible other modes, such as Tearing and Drift instabilities, is performed and the Ballooning growth rate is shown to be the best fit to the experimental values

  5. Parameter study of hightokamak reactors with circular and strongly elongated cross section

    International Nuclear Information System (INIS)

    Herold, H.

    1977-05-01

    A simplified reactor model is used to study the influence of critical β-values on economy parameters and dimensions of possible long time pulsed tokamak reactors. Various betas deduced from stability and equilibrium MHD theory are introduced and put into the scaling in context with technological constraints, as maximum B-field, core constraint, maximum wall loading a.o. The plasma physical concepts treated comprise circular and strongly elongated cross section and approximated FCT equilibria. The computational results are presented as plots of possible economy parameter ranges (magnet energy, wall loading, volumina, investment costs per unit power) dependent on β for suitably chosen hierarchies of the constraints. - A burn time reduction by the build ups of α-pressure may be possible for the pressure profile sensitive high-β equilibria (FCT). Burn times in the 1O sec range, resulting from simple estimates, would about cancel the economic advantages of reactors with high-β equilibria compared to a β = 5% standardreactor (UWMAK I). (orig.) [de

  6. A high resolution Mirnov array for the Mega Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Hole, M. J.; Appel, L. C.; Martin, R.

    2009-01-01

    Over the past two decades, the increase in neutral-beam heating and α particle production in magnetically confined fusion plasmas has led to an increase in energetic particle driven mode activity, much of which has an electromagnetic signature which can be detected by the use of external Mirnov coils. Typically, the frequency and spatial wave number band of such oscillations increase with increasing injection energy, offering new challenges for diagnostic design. In particular, as the frequency approaches the megahertz range, care must be taken to model the stray capacitance of the coil, which limits the resonant frequency of the probe; model transmission line effects in the system, which if unchecked can produce system resonances; and minimize coil conductive shielding, so as to minimize skin currents which limit the frequency response of the coil. As well as optimizing the frequency response, the coils should also be positioned to confidently identify oscillations over a wide wave number band. This work, which draws on new techniques in stray capacitance modeling and coil positioning, is a case study of the outboard Mirnov array for high-frequency acquisition in the Mega Ampere Spherical Tokamak, and is intended as a roadmap for the design of high frequency, weak field strength magnetic diagnostics.

  7. Poloidal magnetic field profile measurements on the microwave tokamak experiment using far-infrared polarimetry

    International Nuclear Information System (INIS)

    Rice, B.W.

    1992-09-01

    The measurement of plasma poloidal magnetic field (B) profiles in tokamaks with good temporal and spatial resolution has proven to be a difficult but important measurement. A large range of toroidal confinement phenomena is expected to depend sensitively on the radial variation of B including the tearing instability, sawtooth oscillations, disruptions, and transport. Experimental confirmation of theoretical models describing these phenomena has been hampered by the lack of detailed B measurements. A fifteen chord far-infrared (FIR) polarimeter has been developed to measure B in the Microwave Tokamak, Experiment (MTX). Polarimetry utilizes the well known Faraday rotation effect, which causes a rotation of the polarization of an FIR beam propagating in the poloidal plane. The rotation angle is proportional to the component of B parallel to the beam. A new technique for determining the Faraday rotation angle is introduced, based on phase measurements of a rotating polarization ellipse. This instrument has been used successfully to measure B profiles for a wide range of experiments on MTX. For ohmic discharges, measurements of the safety factor on axis give q 0 ∼ 0.75 during sawteeth and q 0 > 1 without sawteeth. Large perturbations to the polarimeter signals correlated with the sawtooth crash are observed during some discharges. Measurements in discharges with electron cyclotron heating (ECH) show a transition from a hollow to peaked J profile that is triggered by the ECH pulse. Current-ramp experiments were done to perturb the J profile from the nominal Spitzer conductivity profile. Profiles for initial current ramps and ramps starting from a stable equilibrium have been measured and are compared with a cylindrical diffusion model. Finally, the tearing mode stability equation is solved using measured J profiles. Stability predictions are in good agreement with the existence of oscillations observed on the magnetic loops

  8. The high beta tokamak-extended pulse magnetohydrodynamic mode control research program

    International Nuclear Information System (INIS)

    Maurer, D A; Bialek, J; Byrne, P J; De Bono, B; Levesque, J P; Li, B Q; Mauel, M E; Navratil, G A; Pedersen, T S; Rath, N; Shiraki, D

    2011-01-01

    The high beta tokamak-extended pulse (HBT-EP) magnetohydrodynamic (MHD) mode control research program is studying ITER relevant internal modular feedback control coil configurations and their impact on kink mode rigidity, advanced digital control algorithms and the effects of plasma rotation and three-dimensional magnetic fields on MHD mode stability. A new segmented adjustable conducting wall has been installed on the HBT-EP and is made up of 20 independent, movable, wall shell segments instrumented with three distinct sets of 40 saddle coils, totaling 120 in-vessel modular feedback control coils. Each internal coil set has been designed with varying toroidal angular coil coverage of 5, 10 and 15 0 , spanning the toroidal angle range of an ITER port plug based internal coil to test resistive wall mode (RWM) interaction and multimode MHD plasma response to such highly localized control fields. In addition, we have implemented 336 new poloidal and radial magnetic sensors to quantify the applied three-dimensional fields of our control coils along with the observed plasma response. This paper describes the design and implementation of the new control shell incorporating these control and sensor coils on the HBT-EP, and the research program plan on the upgraded HBT-EP to understand how best to optimize the use of modular feedback coils to control instability growth near the ideal wall stabilization limit, answer critical questions about the role of plasma rotation in active control of the RWM and the ferritic resistive wall mode, and to improve the performance of MHD control systems used in fusion experiments and future burning plasma systems.

  9. 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

  10. Low cost alternative of high speed visible light camera for tokamak experiments

    Czech Academy of Sciences Publication Activity Database

    Odstrčil, T.; Odstrčil, Michal; Grover, O.; Svoboda, V.; Ďuran, Ivan; Mlynář, Jan

    2012-01-01

    Roč. 83, č. 10 (2012), 10E505-10E505 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostic * high speed camera * GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4731003

  11. Conceptual design of plasma position control of SST-1 tokamak using vertical field coil

    International Nuclear Information System (INIS)

    Gulati, Hitesh Kumar; Patel, Kiritkumar B.; Dhongde, Jasraj

    2015-01-01

    SST-1 (Steady State Superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. Based on the solution of Grad-Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec. The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span. In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results. (author)

  12. Measurement of the poloidal magnetic field in the PBX-M tokamak using the motional Stark effect

    International Nuclear Information System (INIS)

    Levinton, F.M.; Fonck, R.J.; Gammel, G.M.; Kaita, R.; Kugel, H.W.; Powell, E.T.; Roberts, D.W.

    1989-05-01

    Polarimetry measurements of the Doppler-shifted H/sub α/ emission from a hydrogen neutral beam on the PBX-M tokamak have been employed in a novel technique for obtaining q(0) and poloidal magnetic field profiles. The electric field from the beam particle motion across the magnetic field (E = V/sub beam/ /times/ B) causes a wavelength splitting of several angstroms, and polarization of the emitted radiation (Stark effect). Viewed transverse to the fields, the emission is linearly polarized with the angle of polarization related to the direction of the magnetic field. 14 refs., 5 figs

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

    International Nuclear Information System (INIS)

    Punjabi, A.

    1994-01-01

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

  14. Transport theory for energetic alpha particles and tolerable magnitude of error fields in tokamaks with broken symmetry

    International Nuclear Information System (INIS)

    Shaing, K.C.; Hsu, C.T.

    2014-01-01

    A transport theory for energetic fusion born alpha particles in tokamaks with broken symmetry has been developed. The theory is a generalization of the theory for neoclassical toroidal plasma viscosity for thermal particles in tokamaks. It is shown that the radial energy transport rate can be comparable to the slowing down rate for energetic alpha particles when the ratio of the typical magnitude of the perturbed magnetic field strength to that of the equilibrium magnetic field strength is of the order of 10 −4 or larger. This imposes a constraint on the magnitude of the error fields in thermonuclear fusion reactors. The implications on stellarators as potential fusion reactors are also discussed. (paper)

  15. Investigation of magnetic reconnection during a sawtooth crash in a high temperature tokamak

    International Nuclear Information System (INIS)

    Yamada, M.; Pomphrey, N.; Budney, R.; Macickam, J.; Nagayama, Y.

    1994-09-01

    This paper reports on a recent laboratory investigation on magnetic reconnection in high temperature tokamak plasmas. The motional stark effect(MSE) diagnostic is employed to measure the pitch angle of magnetic field lines, and hence the q profile. An analytical expression that relates pitch angle to q profile has been developed for a toroidal plasma with circular cross section. During the crash phase of sawtooth oscillations in the plasma discharges, the ECE (electron cyclotron emission) diagnostic measures a fast flattening of the 2-D electron temperature profile in a poloidal plane, an observation consistent with the Kadomtsev reconnection theory. On the other hand motional the MSE measurements indicate that central q values do not relax to unity after the crash, but increase only by 5-10%, typically from 0.7 to 0.75. The latter result is in contradiction with the models of Kadomtsev and/or Wesson. The present study addresses this puzzle by a simultaneous analysis of electron temperature and q profile evolutions. Based on a heuristic model for the magnetic reconnection during the sawtooth crash, the small change of q, i.e. partial reconnection, is attributed to the precipitous drop of pressure gradients which drive the instability and the reconnection process as well as flux conserving plasma dynamics

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

    Science.gov (United States)

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

    2017-10-01

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

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

    International Nuclear Information System (INIS)

    Frerichs, Heinke Gerd

    2010-04-01

    The development of nuclear fusion as an alternative energy source requires the research on magnetically confined, high temperature plasmas. In particular, the quantification of plasma flows in the domain near exposed material surfaces of the plasma container by computer simulations is of key importance, both for guiding interpretation of present fusion experiments and for aiding the ongoing design activities for large future devices such as ITER, W7-X or the DEMO reactor. There is a large number of computational issues related to the physics of hot, fully ionized and magnetized plasmas near surfaces of the vacuum chamber. This thesis is dedicated to one particular such challenge, namely the numerical quantification of self-consistent kinetic neutral gas and plasma fluid flows in very complex 3D (partially chaotic) magnetic fields, in the absence of any common symmetries for plasma and neutral gas dynamics. Such magnetic field configurations are e.g. generated by externally applied magnetic perturbations at the plasma edge, and are of great interest for the control of particle and energy exhausts. In the present thesis the 3D edge plasma and neutral particle transport code EMC3-EIRENE is applied to two distinct configurations of open chaotic magnetic system: at the TEXTOR and DIII-D tokamaks. Improvements of the edge transport model and extensions of the transport code are presented, which have allowed such simulations for the first time for 3D scenarios at DIII-D with ITER similar plasmas. A strong 3D effect of the chaotic magnetic field on the DIII-D edge plasma is found and analyzed in detail. It is found that a pronounced striation pattern of target particle and heat fluxes at DIII-D can only be obtained up to a certain upper limiting level of anomalous cross-field transport. Hence, in comparison to experimental data, these findings allow to narrow down the range of this model parameter. One particular interest at TEXTOR is the achievement of a regime with

  18. Alpha-particle effects on high-n instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rewoldt, G.

    1988-06-01

    Hot α-particles and thermalized helium ash particles in tokamaks can have significant effects on high toroidal mode number instabilities such as the trapped-electron drift mode and the kinetically calculated magnetohydrodynamic ballooning mode. In particular, the effects can be stabilizing, destabilizing, or negligible, depending on the parameters involved. In high-temperature tokamaks capable of producing significant numbers of hot α-particles, the predominant interaction of the mode with the α-particles is through resonances of various sorts. In turn, the modes can cause significant anomalous transport of the α-particles and the helium ash. Here, results of comprehensive linear eigenfrequency-eigenfunction calculations are presented for relevant realistic cases to show these effects. 24 refs., 12 figs., 6 tabs

  19. Measurement of high-energy electrons by means of a Cherenkov detector in ISTTOK tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jakubowski, L., E-mail: lech.Jjakubowski@ipj.gov.p [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Zebrowski, J. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Plyusnin, V.V. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal); Malinowski, K.; Sadowski, M.J.; Rabinski, M. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal)

    2010-10-15

    The paper concerns detectors of the Cherenkov radiation which can be used to measure high-energy electrons escaping from short-living plasma. Such detectors have high temporal (about 1 ns) and spatial (about 1 mm) resolution. The paper describes a Cherenkov-type detector which was designed, manufactured and installed in the ISTTOK tokamak in order to measure fast runaway electrons. The radiator of that detector was made of an aluminium nitride (AlN) tablet with a light-tight filter on its front surface. Cherenkov signals from the radiator were transmitted through an optical cable to a fast photomultiplier. It made possible to perform direct measurements of the runaway electrons of energy above 80 keV. The measured energy values and spatial characteristics of the recorded electrons appeared to be consistent with results of numerical modelling of the runaway electron generation process in the ISTTOK tokamak.

  20. Structural design of the superconducting Poloidal Field coils for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    O'Connor, T.G.; Zbasnik, J.P.

    1993-01-01

    The Tokamak Physics Experiment concept design uses superconducting coils made from cable-in-conduit conductor to accomplish both magnetic confinement and plasma initiation. The Poloidal Field (PF) magnet system is divided into two subsystems, the central solenoid and the outer ring coils, the latter is focus of this paper. The eddy current heating from the pulsed operation is excessive for a case type construction; therefore, a ''no case'' design has been chosen. This ''no case'' design uses the conductor conduit as the primary structure and the electrical insulation (fiberglass/epoxy wrap) as a structural adhesive. The model integrates electromagnetic analysis and structural analysis into the finite element code ANSYS to solve the problem. PF coil design is assessed by considering a variety of coil current wave forms, corresponding to various operating modes and conditions. The structural analysis shows that the outer ring coils are within the requirements of the fatigue life and fatigue crack growth requirements. The forces produced by the Toroidal Field coils on the PF coils have little effect on the maximum stresses in the PF coils. In addition in an effort to reduce the cost of the coils new elongated PF coils design was proposed which changes the aspect ratio of the outer ring coils to reduce the number of turns in the coils. The compressive stress in the outer ring coils is increased while the tensile stress is decreased

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. Plasma rotation and radial electric field with a density ramp in an ohmically heated tokamak

    International Nuclear Information System (INIS)

    Duval, B.P.; Joye, B.; Marchal, B.

    1991-10-01

    Measurements of toroidal and poloidal rotation of the TCA plasma with Alfven Wave Heating and different levels of gas feed are reported. The temporal evolution of the rotation was inferred from intrinsic spectral lines of CV, CIII and, using injected helium gas, from HeII. The light collection optics and line intensity permitted the evolution of the plasma rotation to be measured with a time resolution of 2ms. The rotation velocities were used to deduce the radial electric field. With Alfven heating there was no observable change of this electric field that could have been responsible for the density rise which is characteristic of the RF experiments on TCA. The behaviour of the plasma rotation with different plasma density ramp rates was investigated. The toroidal rotation was observed to decrease with increasing plasma density. The poloidal rotation was observed to follow the value of the plasma density. With hard gas puffing, changes in the deduced radial electric field were found to coincide with changes in the peaking of the plasma density profile. Finally, with frozen pellet injection, the expected increase in the radial electric field due to the increased plasma density was not observed, which may explain the poorer confinement of the injected particles. Even in an ohmically heated tokamak, the measurement of the plasma rotation and the radial electric field are shown to be strongly related to the confinement. A thorough statistical analysis of the systematic errors is presented and a new and significant source of uncertainty in the experimental technique is identified. (author) 18 figs., 18 refs

  3. A distributed high speed data acquisition system for KT5C tokamak

    International Nuclear Information System (INIS)

    Sun Xiang; Wang Zhijiang; Lu Ronghua; Wang Jun; Yu Yi; Zhu Zhenghua; Wen Yizhi; Wan Shude; Liu Wandong; Yu Changxuan

    2005-01-01

    The development of a distributed data acquisition system with low cost to implement high speed data collection through the campus networks for a small tokamak, KT5C, is presented. Data of 512 k bytes at 5 MHz from 5 channels for each can be collected during about 10s after three researchers at different positions demand this system for acquisitions. This system realizes long distance multiuser operations; virtually efficiency of the data acquisition is enhanced. (authors)

  4. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Truong, D. D., E-mail: dtruong@wisc.edu [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Austin, M. E. [Institute for Fusion Studies, University of Texas, Austin, Texas, 78712 (United States)

    2014-11-15

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of T{sub e}(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83–130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1–3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6–0.8 cm) resolution T{sub e} measurements. The high resolution subsystem branches off from the regular channels’ IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2–4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters’ center frequencies (250 MHz). This configuration allows for full coverage of the 83–130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a “zoomed-in” analysis of a ∼2–4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial T{sub e} measurements, which demonstrate that the desired resolution is achieved, are presented.

  5. Compatibility of advanced tokamak plasma with high density and high radiation loss operation in JT-60U

    International Nuclear Information System (INIS)

    Takenaga, H.; Asakura, N.; Kubo, H.; Higashijima, S.; Konoshima, S.; Nakano, T.; Oyama, N.; Ide, S.; Fujita, T.; Takizuka, T.; Kamada, Y.; Miura, Y.; Porter, G.D.; Rognlien, T.D.; Rensink, M.E.

    2005-01-01

    Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high β p H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operation regime is extended to high density above the Greenwald density (n GW ) with high confinement (HH y2 >1) and high radiation loss fraction (f rad >0.9) by tailoring the internal transport barriers (ITBs). High confinement of HH y2 =1.2 is sustained even with 80% radiation from the main plasma enhanced by accumulated metal impurity. The divertor radiation is enhanced by Ne seeding and the ratio of the divertor radiation to the total radiation is increased from 20% without seeding to 40% with Ne seeding. In the high β p H-mode plasmas, high confinement (HH y2 =0.96) is maintained at high density (n-bar e /n GW =0.92) with high radiation loss fraction (f rad ∼1) by utilizing high-field-side pellets and Ar injections. The high n-bar e /n GW is obtained due to a formation of clear density ITB. Strong core-edge parameter linkage is observed, as well as without Ar injection. In this linkage, the pedestal β p , defined as β p ped =p ped /(B p 2 /2μ 0 ) where p ped is the plasma pressure at the pedestal top, is enhanced with the total β p . The radiation profile in the main plasma is peaked due to Ar accumulation inside the ITB and the measured central radiation is ascribed to Ar. The impurity transport analyses indicate that Ar accumulation by a factor of 2 more than the electron, as observed in the high β p H-mode plasma, is acceptable even with peaked density profile in a fusion reactor for impurity seeding. (author)

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

    International Nuclear Information System (INIS)

    Tani, Keiji; Kobayashi, Tomofumi; Tamura, Sanae

    1975-02-01

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

  7. 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.)

  8. Influence of error fields on the plasma confining field and the plasma confinement in tokamak

    International Nuclear Information System (INIS)

    Matsuda, Shinzaburo

    1977-05-01

    Influence of error fields on the plasma confining field and the plasma confinement is treated in the standpoint of design. In the initial breakdown phase before formation of the closed magnetic surfaces, the vertical field properly applied is the most important. Once the magnetic surfaces are formed, the non-axisymmetric error field is important. Effect of the shell gap associated with iron core and with pulsed vertical coils is thus studied. The formation of magnetic islands due to the external non-axisymmetric error field is studied with a simple model. A method of suppressing the islands by choosing the minor periodicity is proposed. (auth.)

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

    International Nuclear Information System (INIS)

    Amrollahi, R.

    1997-01-01

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

  10. 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)

  11. Multi-megajoule heating of large tokamaks with high energy heavy ion beams

    International Nuclear Information System (INIS)

    Dei-Cas, R.

    1981-07-01

    The fast neutral injection heating and RF heating for tokamak like plasmas are now well established. We consider in this paper the use of high energy (approximately 1 GeV) heavy ions (Xe 132 ) to reach ignition in JET or INTOR like tokamaks. The main advantages of such a method will be outlined. The capture and the confinement of heavy ions have been analysed in a particular case and with the described RF linac it seems possible to inject in the order of 50 MJ in 1 sec with a modest increase of the effective charge Zsub(eff)<1.05 in a JET-like plasma for a particle life time of 1 sec and then the additional radiated power should be maintained at a relatively low level in comparison to the injected power

  12. Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

    International Nuclear Information System (INIS)

    Beiersdorfer, P.; von Goeler, S.; Bitter, M.; Hill, K.W.

    1989-07-01

    An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 ≤ Z ≤ 28) and neonlike (34 ≤ Z ≤ 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs

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

    Science.gov (United States)

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

    1980-01-01

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

  14. Power plant design study of a high aspect ratio Tokamak using a SiC composite structure

    International Nuclear Information System (INIS)

    Murakami, Y.; Takase, H.; Shinya, K.

    1998-01-01

    The DREAM (drastically easy maintenance) tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio (A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high-A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m -2 , the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m -2 . A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition. (orig.)

  15. Development of high-speed and wide-angle visible observation diagnostics on Experimental Advanced Superconducting Tokamak using catadioptric optics

    International Nuclear Information System (INIS)

    Yang, J. H.; Hu, L. Q.; Zang, Q.; Han, X. F.; Shao, C. Q.; Sun, T. F.; Chen, H.; Wang, T. F.; Li, F. J.; Hu, A. L.; Yang, X. F.

    2013-01-01

    A new wide-angle endoscope for visible light observation on the Experimental Advanced Superconducting Tokamak (EAST) has been recently developed. The head section of the optical system is based on a mirror reflection design that is similar to the International Thermonuclear Experimental Reactor-like wide-angle observation diagnostic on the Joint European Torus. However, the optical system design has been simplified and improved. As a result, the global transmittance of the system is as high as 79.6% in the wavelength range from 380 to 780 nm, and the spatial resolution is <5 mm for the full depth of field (4000 mm). The optical system also has a large relative aperture (1:2.4) and can be applied in high-speed camera diagnostics. As an important diagnostic tool, the optical system has been installed on the HT-7 (Hefei Tokamak-7) for its final experimental campaign, and the experiments confirmed that it can be applied to the investigation of transient processes in plasma, such as ELMy eruptions in H-mode, on EAST

  16. Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

    Science.gov (United States)

    Punjabi, Alkesh; Ali, Halima

    2008-12-01

    A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates (ψ,θ) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. ψ is the toroidal magnetic flux and θ is the poloidal angle. Natural canonical coordinates (ψ,θ,φ) can be transformed to physical position (R,Z,φ) using a canonical transformation. (R,Z,φ) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonical coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.

  17. Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

    International Nuclear Information System (INIS)

    Punjabi, Alkesh; Ali, Halima

    2008-01-01

    A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates (ψ,θ) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. ψ is the toroidal magnetic flux and θ is the poloidal angle. Natural canonical coordinates (ψ,θ,φ) can be transformed to physical position (R,Z,φ) using a canonical transformation. (R,Z,φ) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonical coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.

  18. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  19. Comments on thermal runaway experiments in sub-ignition tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.

    1982-09-01

    Justification of deuterium-tritium operations is discussed from the physics viewpoint and optimal thermal runaway experiments in high-field, high-density compact tokamaks are suggested within the minimization of the induced radioactivation. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. 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

  2. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    International Nuclear Information System (INIS)

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

  3. Configuration development of a hydraulic press for preloading the toroidal field coils of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

    The Fusion Engineering Design Center (FEDC) is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak (CIT). To achieve a compact device with the minimum major radius, a vertical preload system is being developed to react the vertical separating force normally carried by the inboard leg of the toroidal field (TF) coils. The preload system is in the form of a hydraulic press. Challenges in the design include the development of hydraulic and structural systems for very large force requirements, which could interface with the CIT machine, while allowing maximum access to the top, bottom, and radial periphery of the machine. Maximum access is necessary for maintenance, diagnostics, instrumentation, and control systems. Materials used in the design must function in the nuclear environment and in the presence of high magnetic fields. The structural system developed is an arrangement in which the CIT device is installed in the jaws of the press. Large built-up beams above and below the CIT span the machine and deliver the vertical force to the center cylinder formed by the inboard legs of the TF coils. During the conceptual design study, the vertical force requirement has ranged between 25,000 and 52,000 t. The access requirement on top and bottom limits the width of the spanning beams. Nonmagnetic steel materials are also required because of operation in the high magnetic fields. In the hydraulic system design for the press, several options are being explored. These range from small-diameter jacks operating at very high pressure [228 MPa (33 ksi)] to large-diameter jacks operating at pressures up to 69 MPa (10 ksi). Configurations with various locations for the hydraulic cylinders have also been explored. The nuclear environment and maintenance requirements are factors that affect cylinder location. This paper presents the configuration development of the hydraulic press used to vertically preload the CIT device

  4. Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak

    Directory of Open Access Journals (Sweden)

    Casali L.

    2014-01-01

    Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.

  5. Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

    NARCIS (Netherlands)

    Crombe, K.; Andrew, Y.; Biewer, T. M.; Blanco, E.; de Vries, P. C.; Giroud, C.; Hawkes, N. C.; Meigs, A.; Tala, T.; von Hellermann, M.; Zastrow, K. D.

    2009-01-01

    A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been

  6. World's largest DC flywheel generator for the toroidal field power supply of JAERI's JFT-2M Tokamak nuclear fusion reactor

    International Nuclear Information System (INIS)

    Tani, Takashi; Nakanishi, Yuji; Horita, Tsuyoshi; Kawase, Chiharu; Oyabu, Isao; Kishimoto, Takeshi.

    1996-01-01

    Mitsubishi Electric has delivered the world's largest DC generator for the toroidal field coil power supply of the JFT-2M Tokamak at the Japan Atomic Energy Research Institute. The unit rotates at 225 or 460 rpm, providing a maximum rated output of 2,700 V, 19,000 A and 51.3 MW. The toroidal field is a DC field, so use of a DC generator permits a simpler design consuming less floor space than an AC drive system. The generator was manufactured following extensive studies on commutation, mechanical strength and insulation. (author)

  7. Model validation for radial electric field excitation during L-H transition in JFT-2M tokamak

    Science.gov (United States)

    Kobayashi, T.; Itoh, K.; Ido, T.; Kamiya, K.; Itoh, S.-I.; Miura, Y.; Nagashima, Y.; Fujisawa, A.; Inagaki, S.; Ida, K.; Hoshino, K.

    2017-07-01

    In this paper, we elaborate the electric field excitation mechanism during the L-H transition in the JFT-2M tokamak. Using time derivative of the Poisson’s equation, models of the radial electric field excitation is examined. The sum of the loss-cone loss current and the neoclassical bulk viscosity current is found to behave as the experimentally evaluated radial current that excites the radial electric field. The turbulent Reynolds stress only plays a minor role. The wave convection current that produces a negative current at the edge can be important to explain the ambipolar condition in the L-mode.

  8. Neoclassical transport analysis for a class of hightokamak equilibria

    International Nuclear Information System (INIS)

    Rieser, H.; Werthmann, H.; Kuhn, S.

    1995-01-01

    Balescu's neoclassical transport theory is extended to the case of non-circular flux-surface geometries. Modified classical and neoclassical transport equations, governing particle and heat fluxes in the short- and long-mean-free-path regimes, are derived. These equations are shown to coincide to leading order with the corresponding equations given by Hirshman and Sigmar. They are then applied to an ideal MHD equilibrium, suitable as a simplified but analytically tractable model of a hightokamak. Numerical results for the radial profiles of the global (i.e. flux-surface integrated) particle and heat fluxes in the classical, Pfirsch-Schlueter and banana regimes are presented for geometry and plasma parameters realized in some tokamaks, like the divertor and injection tokamak experiment (DITE). This spatial representation provides direct insight into the overall collisional transport behaviour of a given equilibrium, whereas the anomalous transport problem is not addressed here. Our results demonstrate that for a given pressure profile the global neoclassical fluxes may depend very sensitively on the temperature profiles and that, in particular, the global classical and neoclassical ion heat fluxes exhibit a characteristic non-monotonic behaviour. (author)

  9. High performance experiments on high pressure supersonic molecular beam injection in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Dong Jiafu; Zhou Yan; Feng Beibing; Cao Jianyong; Li Wei; Feng Zhen; Zhang Jiquan; Hong Wenyu; Cui Zhengying; Wang Enyao; Liu Yong

    2004-01-01

    Supersonic molecular beam injection (SMBI) was first proposed and demonstrated on the HL-1 tokamak and was successfully developed and used on HL-1M. Recently, new results of SMBI experiments were obtained by increasing the gas pressure from 0.5 to over 1.0 MPa. A stair-shaped density increment was obtained with high-pressure multi-pulse SMBI that was similar to the density evolution behaviour during multi-pellet injection. This demonstrated the effectiveness of SMBI as a promising fuelling tool for steady-state operation. The penetration depth and injection speed of the high-pressure SMBI were roughly measured from the contour plot of the Hα emission intensity. It was shown that injected particles could penetrate into the core region of the plasma. The penetration speed of high-pressure SMBI particles in the plasma was estimated to be about 1200 m s -1 . In addition, clusters within the beam may play an important role in the deeper injection. (author)

  10. Electron cyclotron wave absorption by the fast tail generated by the dc electric field in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1985-03-01

    Wave damping near the electron gyrofrequency in a tokamak plasma with the energetic tail generated by the dc electric field is investigated. The electron tail is computed by a Fokker-Planck initial value code as a function of the relevant parameter Esub(parallel)/Esub(c)=Esub(parallel)Tsub(e)/(2πsub(e)c 3 Λ). It is shown that in most cases of physical interest strong damping of the e-mode occurs for oblique propagation. The results are of relevance for studies of ECRH in present-day tokamaks and in future reactors where a mildly relativistic electron tail is naturally present for large tsub(e). Special emphasis is therefore given to wave absorption for frequencies f significantly below the central electron gyrofrequency, and to the associated rf-driven current

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  13. Phenomenology of high density disruptions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; McGuire, K.M.; Bell, M.G.

    1993-01-01

    Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) 'cold bubble' precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the 'vacuum bubble' model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. - JETP 38 (1974) 283). (author). Letter-to-the-editor. 25 refs, 3 figs

  14. Background field coils for the High Field Test Facility

    International Nuclear Information System (INIS)

    Zbasnik, J.P.; Cornish, D.N.; Scanlan, R.M.; Jewell, A.M.; Leber, R.L.; Rosdahl, A.R.; Chaplin, M.R.

    1980-01-01

    The High Field Test Facility (HFTF), presently under construction at LLNL, is a set of superconducting coils that will be used to test 1-m-o.d. coils of prototype conductors for fusion magnets in fields up to 12 T. The facility consists of two concentric sets of coils; the outer set is a stack of Nb-Ti solenoids, and the inner set is a pair of solenoids made of cryogenically-stabilized, multifilamentary Nb 3 Sn superconductor, developed for use in mirror-fusion magnets. The HFTF system is designed to be parted along the midplane to allow high-field conductors, under development for Tokamak fusion machines, to be inserted and tested. The background field coils were wound pancake-fashion, with cold-welded joints at both the inner and outer diameters. Turn-to-turn insulation was fabricated at LLNL from epoxy-fiberglass strip. The coils were assembled and tested in our 2-m-diam cryostat to verify their operation

  15. High precision wavelength measurements of X-ray lines emitted from TS-Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Platz, P. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Cornille, M.; Dubau, J. [Observatoire de Paris, 92 - Meudon (France)

    1996-01-01

    X-ray line spectra from highly charged impurity ions have been taken with a high-resolution Bragg-crystal spectrometer on the Tore Supra (TS) tokamak. By cross-checking the wavelengths of reference lines from the heliumlike ions Ti20 + (2.6 Angstroms) and Ar16 + (3.95 Angstroms) we first demonstrate that it is possible to measure wavelengths with a precision, {lambda}/{delta}{lambda}, of better than 50000. We than determine the wavelengths of n=3 to n=2 transitions of neonlike Ag37+ in the 4 Angstroms spectral range. (authors). 16 refs., 7 figs., 3 tabs.

  16. Modelling of prompt losses of high energy charged particles in Tokamaks

    International Nuclear Information System (INIS)

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

    1990-01-01

    A simple analytical expression for the total prompt loss fraction of high energy charged particles in an axisymmetric Tokamak is derived. The results are compared with predictions obtained from numerical simulations and show good agreement. An application is made to sawtooth induced changes in the losses of fusion generated high energy charged particles. Particular emphasis is given to the importance of sawtooth induced profile changes of the background ion densities and temperature as well as to redistribution of particles which have accumulated during the sawtooth rise but are being lost by redistribution at the sawtooth crash. (au)

  17. High #betta# and toroidal effects on the internal kink mode in tokamaks

    International Nuclear Information System (INIS)

    Schmalz, R.

    1982-09-01

    The inclusion of high-#betta# and first-order toroidal terms in the reduced set of (resistive) MHD equations affords the possibility of improving the study of tokamak plasma behaviour by three-dimensional numerical simulation. A new code, GALA, based on the reduced equations is developed. It is used to analyse the linear and nonlinear behaviour of the internal kink mode in equilibria which are generated by a simple relaxation procedure. We find that the inclusion of toroidal effects in high-#betta# equilibria provides considerable stabilization. (orig.)

  18. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    International Nuclear Information System (INIS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-01-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

  19. Modifications of plasma edge electric field and confinement properties by limiter biasing on the KT-5C tokamak

    International Nuclear Information System (INIS)

    Hui Gao; Kan Zhai; Yizhi Wen; Shude Wan; Guiding Wang; Changxun Yu

    1995-01-01

    Experiments using a biased multiblock limiter in the KT-5C tokamak show that positive biasing is more effective than negative biasing in modifying the edge electric field, suppressing fluctuations and improving plasma confinement. The biasing effect varies with the limiter area, the toroidal magnetic field and the biasing voltage. By positive biasing, the edge profiles of the plasma potential, the electron temperature and the density become steeper, resulting in a reduced edge particle flux, an increased global particle confinement time and lower fluctuation levels of the edge plasma. (author)

  20. Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

    Energy Technology Data Exchange (ETDEWEB)

    Crombe, K [Postdoctoral Fellow of the Research Foundation - Flanders (FWO), Department of Applied Physics, Ghent University, Rozier 44, B-9000 Gent (Belgium); Andrew, Y; De Vries, P C; Giroud, C; Hawkes, N C; Meigs, A; Zastrow, K-D [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Biewer, T M [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169, TN (United States); Blanco, E [Laboratorio Nacional de Fusion, Asociacion EURATOM-CIEMAT, Madrid (Spain); Tala, T [VTT Technical Research Centre of Finland, Association EURATOM-Tekes, PO Box 1000, FIN-02044 VTT (Finland); Von Hellermann, M [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands)], E-mail: Kristel.Crombe@jet.uk

    2009-05-15

    A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been calculated from charge exchange measurements of impurity ion temperature, density and rotation velocity profiles, using the force balance equation. The ion temperature and the toroidal and poloidal rotation velocities are compared in plasmas with both reversed and optimized magnetic shear profiles. Poloidal rotation velocity (v{sub {theta}}) in the ITB region is measured to be of the order of a few tens of km s{sup -1}, significantly larger than the neoclassical predictions. Increasing levels of the TF ripple are found to decrease the ion temperature gradient in the ITB region, a measure for the quality of the ITB, and the maximum value of v{sub {theta}} is reduced. The poloidal rotation term dominates in the calculations of the total radial electric field (E{sub r}), with the largest gradient in E{sub r} measured in the radial region coinciding with the ITB.

  1. Method and apparatus for steady-state magnetic measurement of poloidal magnetic field near a tokamak plasma

    Science.gov (United States)

    Woolley, Robert D.

    1998-01-01

    A method and apparatus for the steady-state measurement of poloidal magnetic field near a tokamak plasma, where the tokamak is configured with respect to a cylindrical coordinate system having z, phi (toroidal), and r axes. The method is based on combining the two magnetic field principles of induction and torque. The apparatus includes a rotor assembly having a pair of inductive magnetic field pickup coils which are concentrically mounted, orthogonally oriented in the r and z directions, and coupled to remotely located electronics which include electronic integrators for determining magnetic field changes. The rotor assembly includes an axle oriented in the toroidal direction, with the axle mounted on pivot support brackets which in turn are mounted on a baseplate. First and second springs are located between the baseplate and the rotor assembly restricting rotation of the rotor assembly about its axle, the second spring providing a constant tensile preload in the first spring. A strain gauge is mounted on the first spring, and electronic means to continually monitor strain gauge resistance variations is provided. Electronic means for providing a known current pulse waveform to be periodically injected into each coil to create a time-varying torque on the rotor assembly in the toroidal direction causes mechanical strain variations proportional to the torque in the mounting means and springs so that strain gauge measurement of the variation provides periodic magnetic field measurements independent of the magnetic field measured by the electronic integrators.

  2. The High Field Path to Practical Fusion Energy

    Science.gov (United States)

    Mumgaard, Robert; Whyte, D.; Greenwald, M.; Hartwig, Z.; Brunner, D.; Sorbom, B.; Marmar, E.; Minervini, J.; Bonoli, P.; Irby, J.; Labombard, B.; Terry, J.; Vieira, R.; Wukitch, S.

    2017-10-01

    We propose a faster, lower cost development path for fusion energy enabled by high temperature superconductors, devices at high magnetic field, innovative technologies and modern approaches to technology development. Timeliness, scale, and economic-viability are the drivers for fusion energy to combat climate change and aid economic development. The opportunities provided by high-temperature superconductors, innovative engineering and physics, and new organizational structures identified over the last few years open new possibilities for realizing practical fusion energy that could meet mid-century de-carbonization needs. We discuss re-factoring the fusion energy development path with an emphasis on concrete risk retirement strategies utilizing a modular approach based on the high-field tokamak that leverages the broader tokamak physics understanding of confinement, stability, and operational limits. Elements of this plan include development of high-temperature superconductor magnets, simplified immersion blankets, advanced long-leg divertors, a compact divertor test tokamak, efficient current drive, modular construction, and demountable magnet joints. An R&D plan culminating in the construction of an integrated pilot plant and test facility modeled on the ARC concept is presented.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Goodall, D.H.J.

    1982-01-01

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

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

    Science.gov (United States)

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

    2017-04-28

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

  6. Design and testing of a magnetic shield for the Thomson scattering photomultiplier tubes in the stray fields of the ERASMUS tokamak

    International Nuclear Information System (INIS)

    Desoppere, E.; Van Oost, G.

    1983-01-01

    A multiple coaxial shield system has been designed for the photomultiplier tubes of the ERASMUS tokamak Thomson scattering diagnostic. A stray field of 75 x 10 -4 T was reduced to 0.01 x 10 -4 T for a field parallel to the tube axis, and to 0.03 x 10 -4 T for a perpendicular field

  7. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Marinak, M.

    1988-01-01

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

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

    Czech Academy of Sciences Publication Activity Database

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Progress in Developing a High-Availability Advanced Tokamak Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.; Goldston, R.; Kessel, C.; Neilson, G.; Menard, J.; Prager, S.; Scott, S.; Titus, P.; Zarnstorff, M., E-mail: tbrown@pppl.gov [Princeton University, Princeton Plasma Physics Laboratory, Princeton (United States); Costley, A. [Henley on Thames (United Kingdom); El-Guebaly, L. [University of Wisconsin, Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Waganer, L. [St. Louis (United States)

    2012-09-15

    Full text: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. In this paper the AT pilot plant design will be presented. The selected maintenance approach, the device arrangement and sizing of the in-vessel components and details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will be discussed. Efforts made to enhance the interaction of in-vessel maintenance activities, the hot cell and the transfer process to develop simplifying solutions will also be addressed. (author)

  11. High Field Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1984-12-15

    A Workshop was held in Frascati at the end of September under the title 'Generation of High Fields for Particle Acceleration to Very High Energies'. It was organized by the CERN Accelerator School, the European Committee for Future Accelerators (ECFA) and the Italian INFN and was a further stage in the exploratory moves towards new techniques of acceleration. Such techniques might become necessary to respond to the needs of high energy physics some decades from now when the application of conventional techniques will probably have reached their limits.

  12. Asymmetric electron cyclotron emission from superthermal electrons in the TFR Tokamak

    International Nuclear Information System (INIS)

    1981-03-01

    Measurements of electron cyclotron radiation near the fundamental frequency on the high and low magnetic field side of the TFR Tokamak are reported. In the presence of a superthermal electron component the measured intensities are asymmetric. A theoretical explanation based on the combined effects of the electron relativistic mass variation and the 1/R variation of the tokamak magnetic field is discussed

  13. Long pulse high performance discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Politzer, P.A.

    2001-01-01

    Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 , indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β min >1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼34τ E . These discharges appear to have stationary current profiles with q min ∼1.05, in agreement with the current profile relaxation time ∼1.8 s. (author)

  14. STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    LUCE, TC; WADE, MR; FERRON, JR; HYATT, AW; KELLMAN, AG; KINSEY, JE; LAHAYE, RJ; LASNIER, CJ; MURAKAMI, M; POLITZER, PA; SCOVILLE, JT

    2002-01-01

    A271 STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DII-D TOKAMAK. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak under stationary conditions at relatively low plasma current (q 95 > 4). A figure of merit for fusion gain (β N H 89 /q 95 2 ) has been maintained at values corresponding to | = 10 operation in a burning plasma for > 6 s or 36τ E and 2τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall β limit. Feedback control of the energy content and particle inventory allow reproducible, stationary operation. The particle inventory is controlled by gas fueling and active pumping; the wall plays only a small role in the particle balance. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. In addition, the pulse length capability is greatly increased, which is essential for a technology testing phase of a burning plasma experiment where fluence (duty cycle) is important

  15. Modification of the internal electric field by biasing of the divertor plates in the Tokamak de Varennes (TdeV)

    International Nuclear Information System (INIS)

    Lafrance, D.; Huang, R.; Stansfield, B.L.; Haddad, E.; Lachambre, J.

    1997-01-01

    The radial electric field inside the separatrix has been deduced from spectroscopic measurements of impurities on TdeV (Tokamak de Varennes), using the reduced radial momentum balance and two neoclassical models [R. D. Hazeltine, Phys. Fluids 17, 961 (1974) and Y. B. Kim, P. H. Diamond, and R. J. Groebner, Phys. Fluids B 3, 2050 (1991)]. The results from all three models are in fair agreement. Furthermore, the electric field has been deduced using the same models both with and without biasing the divertor plates relative to the machine wall, showing an inward propagation of the effect of the biasing created in the scrape-off layer (SOL). Undeniably, the electric field has been modified well inside the separatrix (0.6 approx-lt r/a approx-lt 0.9), revealing the possibility of modifying the internal electric field by external means. copyright 1997 American Institute of Physics

  16. High-frequency gyrotrons and their application to tokamak plasma heating

    International Nuclear Information System (INIS)

    Kreischer, K.E.

    1981-01-01

    A comprehensive analysis of high frequency (100 to 200 GHz) and high power (> 100 kW) gyrotrons has been conducted. It is shown that high frequencies will be required in order for electron cyclotron radiation to propagate to the center of a compact tokamak power reactor. High power levels will be needed in order to ignite the plasma with a reasonable number of gyrotron units. In the first part of this research, a set of analytic expressions, valid for all TE cavity modes and all harmonics, is derived for the starting current and frequency detuning using the Vlasov-Maxwell equations in the weakly relativistic limit. The use of an optical cavity is also investigated

  17. Effects of low and high mode number tearing modes in divertor tokamaks

    International Nuclear Information System (INIS)

    Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

    2007-01-01

    The topological effects of magnetic perturbations on a divertor tokamak, such as DIII-D, are studied using field-line maps that were developed by Punjabi et al. [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)]. The studies consider both long-wavelength perturbations, such as those of m=1, n=1 tearing modes, and localized perturbations, which are represented as a magnetic dipole. The parameters of the dipole map are set using DIII-D data from shot 115467 in which the C-coils were activated [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The long-wavelength perturbations alter the structure of the interception of magnetic field lines with the divertor plates, but the interception is in sharp lines. The dipole perturbations cause a spreading of the interception of the field lines with the divertor plates, which alleviates problems associated with heat deposition. Magnetic field lines are the trajectories of a one-and-a-half degree of freedom Hamiltonian, which strongly constrains the topological features of the lines. Although the field line maps that we use do not accurately represent the trajectories through ordinary space of individual field lines, they do represent their topological structure

  18. High temperature outgassing tests on materials used in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Holtrop, K.L.; Hansink, M.J.

    2006-01-01

    This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10 -6 mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented

  19. Disruption mitigation with high-pressure helium gas injection on EAST tokamak

    Science.gov (United States)

    Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

    2018-03-01

    High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

  20. 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.

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

    International Nuclear Information System (INIS)

    Airoldi, A.; Ramponi, G.

    1997-01-01

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

  2. Energy composition of high-energy neutral beams on the COMPASS tokamak

    Directory of Open Access Journals (Sweden)

    Mitosinkova Klara

    2016-12-01

    Full Text Available The COMPASS tokamak is equipped with two identical neutral beam injectors (NBI for additional plasma heating. They provide a beam of deuterium atoms with a power of up to ~(2 × 300 kW. We show that the neutral beam is not monoenergetic but contains several energy components. An accurate knowledge of the neutral beam power in each individual energy component is essential for a detailed description of the beam- -plasma interaction and better understanding of the NBI heating processes in the COMPASS tokamak. This paper describes the determination of individual energy components in the neutral beam from intensities of the Doppler-shifted Dα lines, which are measured by a high-resolution spectrometer viewing the neutral beam-line at the exit of NBI. Furthermore, the divergence of beamlets escaping single aperture of the last accelerating grid is deduced from the width of the Doppler-shifted lines. Recently, one of the NBI systems was modified by the removal of the Faraday copper shield from the ion source. The comparison of the beam composition and the beamlet divergence before and after this modification is also presented.

  3. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

  4. ICRF boronization. A new technique towards high efficiency wall coating for superconducting tokamak reactors

    International Nuclear Information System (INIS)

    Li Jiangang; Zhao Yan Ping; Gu Xue Mao

    1999-01-01

    A new technique for wall conditioning that will be especially useful for future larger superconducting tokamaks, such as ITER, has been successfully developed and encouraging results have been obtained. Solid carborane powder, which is non-toxic and non-explosive, was used. Pulsed RF plasma was produced by a non-Faraday shielding RF antenna with RF power of 10 kW. The ion temperature was about 2 keV with a toroidal magnetic field of 1.8 T and a pressure of 3x10 -1 Pa. Energetic ions broke up the carborane molecules, and the resulting boron ions struck and were deposited on the first wall. In comparison with glow discharge cleaning boronization, the B/C coating film shows higher adhesion, more uniformity and longer lifetime during plasma discharges. The plasma performance was improved after ICRF boronization. (author). Letter-to-the-editor

  5. Study of the Tokamak-15 Superconducting Toroidal Field Coil (STFC) heating under the quench

    International Nuclear Information System (INIS)

    Anashkin, I.O.; Kabanovsky, S.V.; Chudnovsky, A.N.; Khvostenko, P.P.; Vertiporokh, A.N.; Ivanov, D.P.; Posadsky, I.A.

    1994-01-01

    Experiments in Tokamak-15 were performed to study the STFC heating under the quench. The quench was specially caused by current introduction into STFC at the unchanged input helium temperature. The experimental results and simulation data on temperature heating and amount of heat realized in the pancakes under the quench are given. In the experiments was shown that quench occurs in the internal turns of pancakes and estimations of maximal temperature heating corresponds to calculated ones

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

    International Nuclear Information System (INIS)

    Shi Bingren

    2002-01-01

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

  7. Simulation study on dynamics of runaways in tokamaks

    International Nuclear Information System (INIS)

    Liu Jian; Qin Hong; Fisch, Nathaniel J.

    2014-01-01

    Electrons with high velocities can be accelerated to very high energies by a strong electric field to form runaway electrons. In tokamak, runaway electrons are produced in many different processes, including the acceleration from the high-energy tail of thermal distribution, through the runaway avalanche, during the rf wave heating and other non-Ohmic current drive, and even in the magnetic reconnection. This proceeding focus on different dynamical problems of runaway electrons in tokamaks. (author)

  8. Radial electric field evolution in the vicinity of a rotating magnetic island in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Askinazi, L G; Golant, V E; Kornev, V A; Lebedev, S V; Tukachinsky, A S; Vildjunas, M I; Zhubr, N A

    2006-01-01

    Radial electric field is known to be an important factor affecting transport and confinement in toroidal fusion plasmas. Langmuire probe measurements of peripheral radial electric field evolution in the presence of a rotating MHD island were performed on the TUMAN-3M tokamak in order to clear up the possible connection between the radial electric field and the island rotation, both in L and H-modes. The measurements showed that E r became positive, if the island was large enough, in spite of the constant direction of the island's rotation. Comparing similar ohmic H-mode discharges with or without a rotating MHD island, it was found that in the presence of the large island E r was always more positive. Possible explanations of this observation are discussed

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

    International Nuclear Information System (INIS)

    Evans, T.E.

    1984-09-01

    The first direct observation of the internal structure of driven global Alfven eigenmodes in a tokamak plasma is presented. A carbon dioxide laser scattering/interferometer has been designed, built, and installed on the PRETEXT tokamak. By using this diagnostic system in the interferometer configuration, we have for the first time, thoroughly investigated the resonance conditions required for, and the spatial wave field structure of, driven plasma eigenmodes at frequencies below the ion cyclotron frequency in a confined, high temperature, tokamak plasma

  10. Structure of the radial electric field and toroidal/poloidal flow in high temperature toroidal plasma

    International Nuclear Information System (INIS)

    Ida, Katsumi

    2001-01-01

    The structure of the radial electric field and toroidal/poloidal flow is discussed for the high temperature plasma in toroidal systems, tokamak and Heliotron type magnetic configurations. The spontaneous toroidal and poloidal flows are observed in the plasma with improved confinement. The radial electric field is mainly determined by the poloidal flow, because the contribution of toroidal flow to the radial electric field is small. The jump of radial electric field and poloidal flow are commonly observed near the plasma edge in the so-called high confinement mode (H-mode) plasmas in tokamaks and electron root plasma in stellarators including Heliotrons. In general the toroidal flow is driven by the momentum input from neutral beam injected toroidally. There is toroidal flow not driven by neutral beam in the plasma and it will be more significant in the plasma with large electric field. The direction of these spontaneous toroidal flows depends on the symmetry of magnetic field. The spontaneous toroidal flow driven by the ion temperature gradient is in the direction to increase the negative radial electric field in tokamak. The direction of spontaneous toroidal flow in Heliotron plasmas is opposite to that in tokamak plasma because of the helicity of symmetry of the magnetic field configuration. (author)

  11. Design and construction of high-frequency magnetic probe system on the HL-2A tokamak

    Science.gov (United States)

    Liang, S. Y.; Ji, X. Q.; Sun, T. F.; Xu, Yuan; Lu, J.; Yuan, B. S.; Ren, L. L.; Yang, Q. W.

    2017-12-01

    A high-frequency magnetic probe system is designed, calibrated and constructed on the HL-2A tokamak. To investigate the factors which affect the probe frequency response, the inductance and capacitance in the probe system are analyzed using an equivalent circuit. Suitable sizes and turn number of the coil, and the length of transmission cable are optimized based on the theory and detailed test in the calibration. To deal with the frequency response limitation and bake-out, the ceramic grooved technique is used and the probe is wound with a bare copper wire. A cascade filter is manufactured with a suitable bandwidth as well as a good phase consistency between channels. The system has been used in the experiment to measure high frequency (≤300 kHz) magnetohydrodynamic fluctuations, which can meet the requirement of physical analysis on HL-2A.

  12. Inward particle transport at high collisionality in the Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.

    2013-01-01

    We have made the first drift wave study of particle transport in the Experimental Advanced Superconducting Tokamak (Wan et al., Nucl. Fusion 49, 104011 (2009)). The results reveal that collisions make the particle flux more inward in the high collisionality regime. This can be traced back to effects that are quadratic in the collision frequency. The particle pinch is due to electron trapping which is not very efficient in the high collisionality regime so the approach to equilibrium is slow. We have included also the electron temperature gradient (ETG) mode to give the right electron temperature gradient, since the Trapped Electron Mode (TE mode) is weak in this regime. However, at the ETG mode number ions are Boltzmann distributed so the ETG mode does not give particle transport

  13. 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)

  14. Regime of very high confinement in the boronized DIII-D tokamak

    International Nuclear Information System (INIS)

    Jackson, G.L.; Winter, J.; Taylor, T.S.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Holtrop, K.; Lazarus, E.A.; Lao, L.L.; Lippmann, S.I.; Osborne, T.H.; Petrie, T.W.; Phillips, J.; James, R.; Schissel, D.P.; Strait, E.J.; Turnbull, A.D.; West, W.P.; DIII-D Team

    1991-01-01

    Following boronization, tokamak discharges in DIII-D have been obtained with confinement times up to a factor of 3.5 above the ITER89-P L-mode scaling and 1.8 times greater than the DIII-D/JET H-mode scaling relation. Very high confinement phases are characterized by relatively high central density with n e (0)∼1x10 20 m -3 , and central ion temperatures up to 13.6 keV at moderate plasma currents (1.6 MA) and heating powers (12.5--15.3 MW). These discharges exhibit a low fraction of radiated power, P≤25%, Z eff (0) close to unity, and lower impurity influxes than comparable DIII-D discharges before boronization

  15. Measurement of high-beta tokamak pressure profiles with multipoint Thomson scattering

    International Nuclear Information System (INIS)

    Levinton, F.M.

    1983-01-01

    A multipoint Thomson-scattering system has been developed to obtain pressure profiles along the major radius of Torus II, a high-beta tokamak. The profiles obtained during the 20 to 25 μs lifetime of the discharge indicates that the plasma has a peak temperature of 80 eV and density of 1.0 x 10 15 cm - 3 . The profiles remain fairly constant during this time until the equilibrium is lost, after which the temperature and density decays to 10 eV and 10 14 cm - 3 very quickly (approx. 1 μs). Experimental results show Torus II has a high-beta ( approx. 10%) equilibrium, with a strong shift of the peak of the pressure profile towards the outside. Numerical results from a 2-D free boundary MHD equilibrium code have obtained equilibria which closely approximate the experimentally measured profiles

  16. High field electron linacs

    International Nuclear Information System (INIS)

    Le Duff, J.

    1985-12-01

    High field electron linacs are considered as potential candidates to provide very high energies beyond LEP. Since almost twenty years not much improvement has been made on linac technologies as they have been mostly kept at low and medium energies to be used as injectors for storage rings. Today, both their efficiency and their performances are being reconsidered, and for instance the pulse compression sheme developed at SLAC and introduced to upgrade the energy of that linac is a first step towards a new generation of linear accelerators. However this is not enough in terms of power consumption and more development is needed to improve both the efficiency of accelerating structures and the performances of RF power sources

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

    Science.gov (United States)

    Ren, Q.

    2015-11-01

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

  18. High-pressure duo-multichannel soft x-ray spectrometer for tokamak plasma diagnostics

    International Nuclear Information System (INIS)

    Schwob, J.L.; Wouters, A.W.; Suckewer, S.

    1987-03-01

    A high-resolution, time-resolving soft X-ray multichannel spectrometer (SOXMOS) that permits the simultaneous measurement of emission in two different spectral ranges has been developed and tested extensively for tokamak plasma diagnostics. The basic instrument is a high-resolution, interferometrically adjusted, extreme grazing incidence Schwob-Fraenkel duochromator. The instrument is equipped with two multichannel detectors that are adjusted interferometrically and scan along the Rowland circle. Each consists of an MgF 2 coated, funneled microchannel plate, associated with a phosphor screen image intensifier that is coupled to a 1024-element photodiode array by a flexible fibrer optic conduit. The total wavelength coverage of the instrument is 5 to 340 0 A with a measured resolution (FWHM) of about 0.2 A when equipped with a 600 g/mm grating, and 5 to 85 A with a resolution of about 0.06 A using a 2400 g/mm grating. The simultaneous spectral coverage of each detector varies from 15 A at the short wavelength limit to 70 A at the long wavelength limit with the lower dispersion grating. The minimum read-out time for a full spectral portion is 17 ms, but several individual lines can be measured with 1 ms time resolution by selected pixel readout. Higher time resolution can be achieved by replacing one multichannel detector with a single channel electron multiplier detector. Examples of data from the PLT and TFTR tokamaks are presented to illustrate the instrument's versatility, high spectral resolution, and high signal-to-noise ratio even in the 10 A region. 44 refs., 20 figs

  19. Finite element and node point generation computer programs used for the design of toroidal field coils in tokamak fusion devices

    International Nuclear Information System (INIS)

    Smith, R.A.

    1975-06-01

    The structural analysis of toroidal field coils in Tokamak fusion machines can be performed with the finite element method. This technique has been employed for design evaluations of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The application of the finite element method can be simplified with computer programs that are used to generate the input data for the finite element code. There are three areas of data input where significant automation can be provided by supplementary computer codes. These concern the definition of geometry by a node point mesh, the definition of the finite elements from the geometric node points, and the definition of the node point force/displacement boundary conditions. The node point forces in a model of a toroidal field coil are computed from the vector cross product of the coil current and the magnetic field. The computer programs named PDXNODE and ELEMENT are described. The program PDXNODE generates the geometric node points of a finite element model for a toroidal field coil. The program ELEMENT defines the finite elements of the model from the node points and from material property considerations. The program descriptions include input requirements, the output, the program logic, the methods of generating complex geometries with multiple runs, computational time and computer compatibility. The output format of PDXNODE and ELEMENT make them compatible with PDXFORC and two general purpose finite element computer codes: (ANSYS) the Engineering Analysis System written by the Swanson Analysis Systems, Inc., and (WECAN) the Westinghouse Electric Computer Analysis general purpose finite element program. The Fortran listings of PDXNODE and ELEMENT are provided

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-05-01

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

  1. Progress Toward Long Pulse, High Performance Plasmas in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    P.A. Politzer; T.C. Luce; M.E. Austin; J.R. Ferron, A.M. Garofalo; C.M. Greenfield; A.W. Hyatt; R.J. La Haye; L.L. Lao; E.A. Lazarus; M.A. Makowski; M. Murakami; C.C. Petty; R.I. Pinsker; B.W. Rice; E.J. Strait, M.R. Wade; J.G. Watkins

    2000-01-01

    A major portion of the research program of the DIII-D tokamak collaboration is devoted to the development and demonstration of high performance advanced tokamak plasmas, with profiles as close as possible to those anticipated for steady-state operation. The work during the 1999 campaign has resulted in significant progress toward this goal. High normalized performance ((beta)(sub N)(approx) 4 and(beta)(sub N) H(sub 89)(approx) 9) discharges have been sustained for up to 2 s. These plasmas are in H-mode with rapid ELMs. The most common limiting phenomena are resistive wall modes (RWMs) rather than neoclassical tearing modes (NTMs). NTMs do occur, apparently triggered by the RWMs. The observed pressure is well above the calculated beta limit without a wall, and(beta)(sub N) and gt; 4(ell)(sub i) throughout the high performance phase. The bootstrap current is estimated to be and gt;50% of the total, and measurements of the internal loop voltage show that only about 25% of the current is inductively driven. The central q profile is flat, as is the calculated bootstrap current profile, due to the absence of any localized pressure gradients. The residual inductive current is localized around r/a(approx) 0.5. To demonstrate quasi-stationary operation, it will be necessary to replace the residual inductive current with ECCD at the same minor radius. To effectively apply ECH and ECCD to these discharges, density control will be needed. Preliminary experiments using the DIII-D cryopump have reduced the density by(approx)20%. A new EC power system and a new private flux cryopump will be available for the 2000 campaign

  2. Reconstruction of Magnetic Field Surfaces of the NOVILLO Tokamak by means of the 3D-MAPTOR Code

    International Nuclear Information System (INIS)

    Chavez-Alarcon, Esteban; Herrera-Velazquez, J. Julio E.

    2008-01-01

    A 3-D code has been developed in order to simulate the magnetic field lines in circular cross-section tokamaks. The toroidal magnetic field can be obtained from the individual fields of circular coils arranged around the torus, or alternatively, as a ripple-less field, as well as the vertical field coils, and divertor-like coils. The poloidal field is provided by a given toroidal current density profile. Proposing initial conditions for a magnetic filed line, it is integrated along the toroidal angle coordinate, and the Poincare maps can be obtained at any desired cross section plane along the torus. Following this procedure, the code allows to explore the necessary current values for the existence of magnetic field surfaces, allowing for deviations from axial symmetry, such as ripple effects. Therefore it is a good design instrument, in which different parameters and arrangements of coils can be tested. On the other hand, the current signals from experimental devices can be used in order to reconstruct the behaviour of the magnetic field surfaces, including the q(r) profiles. The reconstruction properties of the code are shown in this work

  3. High-beta characteristics of first and second-stable spherical tokamaks in reconnection heating experiments of TS-3

    International Nuclear Information System (INIS)

    Ono, Y.

    2002-01-01

    Novel formations of ultra-high-beta Spherical Tokamak (ST) have been developed in the TS-3 device using high power heating of merging/ reconnection. In Type-A merging, two STs were merged together to build up the plasma beta. In Type-B merging, an oblate FRC was initially formed by merging of two spheromaks with opposing toroidal field B t and was transformed into an ultra-high-beta ST by applying external B t . Ballooning stability analyses confirmed formations of the first-stable STs by Type- A merging and the second-stable STs by Type-B merging and also the unstable STs by both mergings, revealing the ballooning stability window consistent with measured high-n instabilities. We made (1) those model analyses of the produced STs for the first time using the BALLOO stability code, revealing that hollowness/ broadness of current/pressure profiles widen significantly the window to the second-stable regime. This paper also addresses (2) normalized betas of the second-stable STs as large as 6-17 for comparison with the Troyon scaling and (3) a promising scaling of the reconnection heating energy. (author)

  4. Theory of high-n toroidicity-induced shear Alfven eigenmode in tokamaks

    International Nuclear Information System (INIS)

    Fu, G.Y.; Cheng, C.Z.; Princeton Univ., NJ

    1989-07-01

    High-n WKB-ballooning mode equation is employed to study toroidicity-induced shear Alfven eigenmodes (TAE) in the δ - α space, where δ = (r/q)(dq/dr) is the magnetic shear, and α = -(2Rq 2 /B 2 )(dp/dr) is the normalized pressure gradient for tokamak plasmas. In the ballooning mode first stability region, TAE modes are found to exist only for α less than some critical value α c . We also find that these TAE modes reappear in the ballooning mode second stability region for bands of α values. The global envelope structures of these TAE modes are studied by WKB method and are found to be bounded radially if the local mode frequency has a maximum in radius. 15 refs., 14 figs

  5. Proposal for the construction of a High-Beta Tokamak at LASL

    International Nuclear Information System (INIS)

    Van der Laan, P.C.T.; Freidberg, J.P.; Thomas, K.S.

    1976-06-01

    The large heating rate inherent to implosion heating allows the rapid generation of high-beta tokamak plasmas. A study of these plasmas in the proposed HBT machine can give information on how MHD equilibrium and stability limit β and q. Both a wide current profile and a moderate elongation of the minor cross section should help to raise the permissible peak β values in HBT to at least 20 percent. The longer term loss processes occurring in MHD-stable plasmas are to be investigated. The main parameters of HBT are: R = 0.30 m, minor cross section a racetrack of width and height 0.24 m and 0.48 m, B/sub phi/ = 2 T, I/sub phi/ approximately 750 kA

  6. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  7. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    International Nuclear Information System (INIS)

    Weinzettl, Vladimir; Shukla, Gaurav; Ghosh, Joydeep; Melich, Radek; Panek, Radomir; Tomes, Matej; Imrisek, Martin; Naydenkova, Diana; Varju, Josef; Pereira, Tiago; Gomes, Rui; Abramovic, Ivana; Jaspers, Roger; Pisarik, Michael; Odstrcil, Tomas; Van Oost, Guido

    2015-01-01

    Highlights: • We built a new diagnostic of poloidal plasma rotation on the COMPASS tokamak. • Improvements in throughput via toroidal integration and fiber optimizations shown. • Poloidal rotation and ion temperature measured in L- and H-mode and during RMP. • Design and parameters of a new CXRS diagnostic for COMPASS are introduced. - Abstract: High-resolution spectroscopy is a powerful tool for the measurement of plasma rotation as well as ion temperature using the Doppler shift of the emitted spectral lines and their Doppler broadening, respectively. Both passive and active diagnostic variants for the COMPASS tokamak are introduced. The passive diagnostic focused on the C III lines at about 465 nm is utilized for the observation of the poloidal plasma rotation. The current set-up of the measuring system is described, including the intended high-throughput optics upgrade. Different options to increase the fiber collection area are mentioned, including a flower-like fiber bundle, and the use of micro-lenses or tapered fibers. Recent measurements of poloidal plasma rotation of the order of 0–6 km/s are shown. The design of the new active diagnostic using a deuterium heating beam and based on charge exchange recombination spectroscopy (C VI line at 529 nm) is introduced. The tool will provide both space (0.5–5 cm) and time (10 ms) resolved toroidal plasma rotation and ion temperature profiles. The results of the Simulation of Spectra code used to examine the feasibility of charge exchange measurements on COMPASS are shown and connected with a selection of the spectrometer coupled with the CCD camera.

  8. Long-pulse high-performance discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Politzer, P.A.

    2001-01-01

    Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β≤5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼7 for up to 6.3 s or ∼34 τ E . These discharges appear to be in resistive equilibrium with q min ∼1.05, in agreement with the current profile relaxation time of 1.8 s. (author)

  9. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Weinzettl, Vladimir, E-mail: vwei@ipp.cas.cz [Institute of Plasma Physics ASCR, Prague (Czech Republic); Shukla, Gaurav [Institute of Plasma Physics ASCR, Prague (Czech Republic); Department of Applied Physics, Ghent University, Ghent (Belgium); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Ghosh, Joydeep [Institute for Plasma Research, Bhat, Gandhinagar (India); Melich, Radek; Panek, Radomir [Institute of Plasma Physics ASCR, Prague (Czech Republic); Tomes, Matej; Imrisek, Martin; Naydenkova, Diana [Institute of Plasma Physics ASCR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Varju, Josef [Institute of Plasma Physics ASCR, Prague (Czech Republic); Pereira, Tiago [Instituto de Plasmas e Fusão Nuclear, Lisboa (Portugal); Instituto Superior Técnico, Universidade de Lisboa, Lisboa (Portugal); Gomes, Rui [Instituto de Plasmas e Fusão Nuclear, Lisboa (Portugal); Abramovic, Ivana; Jaspers, Roger [Eindhoven University of Technology, Eindhoven (Netherlands); Pisarik, Michael [SQS Vlaknova optika a.s., Nova Paka (Czech Republic); Department of Electromagnetic Field, Faculty of Electrical Engineering, Czech Technical University in Prague (Czech Republic); Odstrcil, Tomas [Max-Planck-Institut fur Plasmaphysik, Garching (Germany); Van Oost, Guido [Department of Applied Physics, Ghent University, Ghent (Belgium)

    2015-10-15

    Highlights: • We built a new diagnostic of poloidal plasma rotation on the COMPASS tokamak. • Improvements in throughput via toroidal integration and fiber optimizations shown. • Poloidal rotation and ion temperature measured in L- and H-mode and during RMP. • Design and parameters of a new CXRS diagnostic for COMPASS are introduced. - Abstract: High-resolution spectroscopy is a powerful tool for the measurement of plasma rotation as well as ion temperature using the Doppler shift of the emitted spectral lines and their Doppler broadening, respectively. Both passive and active diagnostic variants for the COMPASS tokamak are introduced. The passive diagnostic focused on the C III lines at about 465 nm is utilized for the observation of the poloidal plasma rotation. The current set-up of the measuring system is described, including the intended high-throughput optics upgrade. Different options to increase the fiber collection area are mentioned, including a flower-like fiber bundle, and the use of micro-lenses or tapered fibers. Recent measurements of poloidal plasma rotation of the order of 0–6 km/s are shown. The design of the new active diagnostic using a deuterium heating beam and based on charge exchange recombination spectroscopy (C VI line at 529 nm) is introduced. The tool will provide both space (0.5–5 cm) and time (10 ms) resolved toroidal plasma rotation and ion temperature profiles. The results of the Simulation of Spectra code used to examine the feasibility of charge exchange measurements on COMPASS are shown and connected with a selection of the spectrometer coupled with the CCD camera.

  10. LONG-PULSE, HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    T.C. LUCE; M.R. WADE; P.A. POLITZER; S.L. ALLEN; M E. AUSTIN; D.R. BAKER; B.D. BRAY; D.P. BRENNAN; K.H. BURRELL; T.A. CASPER; M.S. CHU; J.D. De BOO; E.J. DOYLE; J.R. FERRON; A.M. GAROFALO; P.GOHIL; I.A. GORELOV; C.M. GREENFIELD; R.J. GROEBNER; W.W. HEIBRINK; C.-L. HSIEH; A.W. HYATT; R.JAYAKUMAR; J.E.KINSEY; R.J. LA HAYE; L.L. LAO; C.J. LASNIER; E.A. LAZARUS; A.W. LEONARD; Y.R. LIN-LIU; J. LOHR; M.A. MAKOWSKI; M. MURAKAMI; C.C. PETTY; R.I. PINSKER; R. PRATER; C.L. RETTIG; T.L. RHODES; B.W. RICE; E.J. STRAIT; T.S. TAYLOR; D.M. THOMAS; A.D. TURNBULL; J.G. WATKINS; W.P.WEST; K.-L. WONG

    2000-01-01

    Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5 τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β ∼(le) 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼ 34 τ E . These discharges appear to be in resistive equilibrium with q min ∼ 1.05, in agreement with the current profile relaxation time of 1.8 s

  11. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. Multi-machine studies of the role of turbulence and electric fields in the establishment of improved confinement in tokamak plasma

    Czech Academy of Sciences Publication Activity Database

    Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Kraemer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Nanobashvili, S.; Peleman, P.; Razumova, K.A.; Stöckel, Jan; Vershkov, V.; Adámek, Jiří; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Brotánková, Jana; Dnestrovskij, A.Yu.; Ďuran, Ivan; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Hron, Martin; Ionita, C.; Jachmich, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Martines, E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Schrittwieser, R.; Soldatov, S.; Spolaore, M.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Žáček, František; Zimmermann, O.

    2007-01-01

    Roč. 49, 5A (2007), A29-A44 ISSN 0741-3335. [International Congress on Plasma Physics/13th./. Kiev, 22.05.2006-26.05.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * turbulence * electric fields Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 3.070, year: 2007

  13. Ripple induced trapped particle loss in tokamaks

    International Nuclear Information System (INIS)

    White, R.B.

    1996-05-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks

  14. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    International Nuclear Information System (INIS)

    Zheng, Linjin; Horton, W.; Miura, H.; Shi, T.H.; Wang, H.Q.

    2016-01-01

    Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew–Goldburger–Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

  15. Characteristics of solid-target charge-exchange analyzers for energetic ion diagnostics on tokamaks

    International Nuclear Information System (INIS)

    Beiersdorfer, P.; Roquemore, A.L.; Kaita, R.

    1987-05-01

    Compact electrostatic charge-exchange analyzers have been constructed for installation in areas of high magnetic fields and restricted access near tokamak fusion devices. The analyzers employed carbon stripping foils, and have been calibrated for proton energies between 1 and 70 keV. They have been successfully used to study charge-exchange losses in auxiliary-heated tokamak plasmas

  16. Cryogenic magnet case and distributed structural materials for high-field superconducting magnets

    International Nuclear Information System (INIS)

    Summers, L.T.; Miller, J.R.; Kerns, J.A.; Myall, J.O.

    1987-01-01

    The superconducting magnets of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) will generate high magnetic fields over large bores. The resulting electromagnetic forces require the use of large volumes of distributed steel and thick magnet case for structural support. Here we review the design allowables, calculated loads and forces, and structural materials selection for TIBER II. 7 refs., 2 figs., 3 tabs

  17. Measurements of the parallel wavenumber of lower hybrid waves in the scrape-off layer of a high-density tokamak

    International Nuclear Information System (INIS)

    Baek, S. G.; Wallace, G. M.; Parker, R. R.; Shiraiwa, S.; Bonoli, P. T.; Brunner, D.; Faust, I.; LaBombard, B. L.; Wukitch, S.; Shinya, T.; Takase, Y.

    2016-01-01

    In lower hybrid current drive (LHCD) experiments on tokamaks, the parallel wavenumber of lower hybrid waves is an important physics parameter that governs the wave propagation and absorption physics. However, this parameter has not been experimentally well-characterized in the present-day high density tokamaks, despite the advances in the wave physics modeling. In this paper, we present the first measurement of the dominant parallel wavenumber of lower hybrid waves in the scrape-off layer (SOL) of the Alcator C-Mod tokamak with an array of magnetic loop probes. The electric field strength measured with the probe in typical C-Mod plasmas is about one-fifth of that of the electric field at the mouth of the grill antenna. The amplitude and phase responses of the measured signals on the applied power spectrum are consistent with the expected wave energy propagation. At higher density, the observed k || increases for the fixed launched k || , and the wave amplitude decreases rapidly. This decrease is correlated with the loss of LHCD efficiency at high density, suggesting the presence of loss mechanisms. Evidence of the spectral broadening mechanisms is observed in the frequency spectra. However, no clear modifications in the dominant k || are observed in the spectrally broadened wave components, as compared to the measured k || at the applied frequency. It could be due to (1) the probe being in the SOL and (2) the limited k || resolution of the diagnostic. Future experiments are planned to investigate the roles of the observed spectral broadening mechanisms on the LH density limit problem in the strong single pass damping regime.

  18. Observation of plasma rotation driven by static nonaxisymmetric magnetic fields in a tokamak.

    Science.gov (United States)

    Garofalo, A M; Burrell, K H; DeBoo, J C; deGrassie, J S; Jackson, G L; Lanctot, M; Reimerdes, H; Schaffer, M J; Solomon, W M; Strait, E J

    2008-11-07

    We present the first evidence for the existence of a neoclassical toroidal rotation driven in a direction counter to the plasma current by nonaxisymmetric, nonresonant magnetic fields. At high beta and with large injected neutral beam momentum, the nonresonant field torque slows down the plasma toward the neoclassical "offset" rotation rate. With small injected neutral beam momentum, the toroidal rotation is accelerated toward the offset rotation, with resulting improvement in the global energy confinement time. The observed magnitude, direction, and radial profile of the offset rotation are consistent with neoclassical theory predictions.

  19. Plasma heating in the TM-3 Tokamak at electron-cyclotron resonance with magnetic fields up to 25 ke

    International Nuclear Information System (INIS)

    Alikaev, V.V.; Bobrovskii, G.A.; Poznyak, V.I.; Razumova, K.A.; Sannikov, V.V.; Sokolov, Yu.A.; Shmarin, A.A.

    Experiments were conducted in heating plasma at electron-cyclotron resonance (ECR) with longitudinal magnetic fields up to 25 ke. It was shown by the aid of laser diagnosis that the temperature of the basic component of the electrons increases in accordance with the classical mechanism of heating at ECR in the process of electron-cyclotron heating (ECH). The distribution of the temperature of electrons with respect to radius was measured. The relationship of energetic lifetime in the Tokamak and electron temperature was obtained and the magnitude of energetic lifetime of accelerated electrons in the function of their energy was estimated. The value β/sub tau/ approximately equal to 2.2 was obtained by the aid of ECH in a regime with small discharge currents

  20. Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

    International Nuclear Information System (INIS)

    Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

    2009-01-01

    The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Plasma rotation effect on interaction of low frequency fields with plasmas at the rational surfaces in tokamaks

    International Nuclear Information System (INIS)

    Rondan, E.R.; Elfimov, A.G.; Galvao, R.M.O.; Pires, C.J.A.

    2006-01-01

    The effect of plasma rotation on low frequency (LF) field penetration, absorption and ponderomotive forces in TEXTOR and in Tokamak Chauffage Alfven Bresilien (TCABR) is investigated in the frequency band of 1-10 kHz. The LF fields are driven by the dynamic ergodic divertor in TEXTOR and the ergodic magnetic limiter in TCABR. Alfven wave mode conversion is responsible for the LF field absorption at the rational magnetic surface where q = -M/N is the integer. Analytical and numerical calculations show the maxima of the LF field absorption at the local Alfven wave resonance ω - k · U = k parallel c A , where ω and k are the frequency and the wave vector, respectively, and c A is the Alfven velocity at the rational magnetic surface q = 2, 3 in TEXTOR and TCABR. The rotation velocity U along the magnetic surfaces, taken into account in the dielectric tensor, can strongly modify the LF field and dissipated power profiles. The absorption in the local AW resonances begins to be non-symmetric in relation to the resonance surface. Calculations show that coil impedance has a maximum related to excitation of some stable (possibly Suydam) modes for waves travelling in the direction of plasma rotation

  4. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

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

  5. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  7. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

  8. Passive stabilization of MHD instabilities at high βn in the HBT-EP Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gates, David A. [Columbia Univ., New York, NY (United States)

    1993-01-01

    The HBT-EP Tokamak has been designed, built, and is now fully operational in the Columbia University Plasma Physics Laboratory. One of the primary purposes of this facility is to study the effects of a conducting wall on the MHD modes that lead up to plasma disruptions. Of particular interest are the types of instabilities that are driven by the kinetic pressure of the plasma, because these instabilities are believed to be responsible for the present limit to plasma β with β ∝/B2, where the is the volume averaged pressure and B is the magnetic field. To this end, a movable conducting wall has been installed inside the HBT-EP vacuum chamber. The primary result of this thesis are the initial results from experiments that study the effect of this wall on plasma instabilities. The experiment shows that the conducting wall significantly reduces the growth rate of instabilities that precede a plasma disruption that occurs when the value of β is near the Troyon limit. The location of the wall required for significant stabilization is b/a ~1.2 where a is the minor radius of the plasma and b is the minor radial location of the wall. Moving the wall closer than b/a = 1.2 slightly degrades the stabilizing effect, which is consistent with recent theories.

  9. Passive stabilization of MHD instabilities at high βn in the HBT-EP Tokamak

    International Nuclear Information System (INIS)

    Gates, D.A.

    1993-01-01

    The HBT-EP Tokamak has been designed, built, and is now fully operational in the Columbia University Plasma Physics Laboratory. One of the primary purposes of this facility is to study the effects of a conducting wall on the MHD modes that lead up to plasma disruptions. Of particular interest are the types of instabilities that are driven by the kinetic pressure of the plasma, because these instabilities are believed to be responsible for the present limit to plasma β with β ∝ /B 2 , where the is the volume averaged pressure and B is the magnetic field. To this end, a movable conducting wall has been installed inside the HBT-EP vacuum chamber. The primary result of this thesis are the initial results from experiments that study the effect of this wall on plasma instabilities. The experiment shows that the conducting wall significantly reduces the growth rate of instabilities that precede a plasma disruption that occurs when the value of β is near the Troyon limit. The location of the wall required for significant stabilization is b/a ∼1.2 where a is the minor radius of the plasma and b is the minor radial location of the wall. Moving the wall closer than b/a = 1.2 slightly degrades the stabilizing effect, which is consistent with recent theories

  10. Demonstration of high performance negative central magnetic shear discharges on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Rice, B.W.; Burrell, K.H.; Lao, L.L.

    1996-01-01

    Reliable operation of discharges with negative central magnetic shear has led to significant increases in plasma performance and reactivity in both low confinement, L-mode, and high confinement, H-mode, regimes in the DIII-D tokamak. Using neutral beam injection early in the initial current ramp, a large range of negative shear discharges have been produced with durations lasting up to 3.2 s. The total non- inductive current (beam plus bootstrap) ranges from 50% to 80% in these discharges. In the region of shear reversal, significant peaking of the toroidal rotation [f φ ∼ 30-60 kHz] and ion temperature [T i (0) ∼ 15-22 keV] profiles are observed. In high power discharges with an L-mode edge, peaked density profiles are also observed. Confinement enhancement factors up to H ≡ τ E /τ ITER-89P ∼ 2.5 with an L-mode edge, and H ∼ 3.3 in an Edge Localized Mode (ELM)-free H-mode, are obtained. Transport analysis shows both ion thermal diffusivity and particle diffusivity to be near or below standard neoclassical values in the core. Large pressure peaking in L- mode leads to high disruptivity with Β N ≡ Β T /(I/aB) ≤ 2.3, while broader pressure profiles in H- mode gives low disruptivity with Β N ≤ 4.2

  11. New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

    Science.gov (United States)

    Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

    2015-02-06

    A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

  12. High frequency ion Bernstein wave heating experiment on JIPP T-IIU tokamak

    International Nuclear Information System (INIS)

    Seki, T.; Kumazawa, R.; Watari, T.

    1992-08-01

    An experiment in a new regime of ion Bernstein wave (IBW) heating has been carried out using 130 MHz high power transmitters in the JIPP T-IIU tokamak. The heating regime utilized the IBW branch between the 3rd and 4th harmonics of the hydrogen ion cyclotron frequencies. This harmonic number is the highest among those used in the IBW experiments ever conducted. The net radio-frequency (RF) power injected into the plasma is around 400 kW, limited by the transmitter output power. Core heating of ions and electrons was confirmed in the experiment and density profile peaking was found to feature the IBW heating (IBWH). The peaking of the density profile was also found when IBW was applied to the neutral beam injection heated discharges. An analysis by use of a transport code with these experimental data indicates that the particle confinement should be improved in the plasma core region on the application of IBWH. It is also found that the ion energy distribution function observed during IBWH has less high energy tail than those in conventional ion cyclotron range of frequency heating regimes. The observed IBWH-produced ion energy distribution function is in a reasonable agreement with the calculation based on the quasi-linear RF diffusion / Fokker-Planck model. (author)

  13. High spatial and temporal resolution charge exchange recombination spectroscopy on the HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Y. L.; Yu, D. L., E-mail: yudl@swip.ac.cn; Liu, L.; Cao, J. Y.; Sun, A. P.; Ma, Q.; Chen, W. J.; Liu, Yi; Yan, L. W.; Yang, Q. W.; Duan, X. R.; Liu, Yong [Southwestern Institute of Physics, Chengdu 610041 (China); Ida, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Hellermann, M. von [ITER Diagnostic Team, IO, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); FOM-Institute for Plasma physics “Rijnhuizen,” Association EURATOM, Trilateral Euregio Cluster, 3430 BE Nieuwegein (Netherlands)

    2014-10-01

    A 32/64-channel charge exchange recombination spectroscopy (CXRS) diagnostic system is developed on the HL-2A tokamak (R = 1.65 m, a = 0.4 m), monitoring plasma ion temperature and toroidal rotation velocity simultaneously. A high throughput spectrometer (F/2.8) and a pitch-controlled fiber bundle enable the temporal resolution of the system up to 400 Hz. The observation geometry and an optimized optic system enable the highest radial resolution up to ~1 cm at the plasma edge. The CXRS system monitors the carbon line emission (C VI, n = 8–7, 529.06 nm) whose Doppler broadening and Doppler shift provide ion temperature and plasma rotation velocity during the neutral beam injection. The composite CX spectral data are analyzed by the atomic data and analysis structure charge exchange spectroscopy fitting (ADAS CXSFIT) code. First experimental results are shown for the case of HL-2A plasmas with sawtooth oscillations, electron cyclotron resonance heating, and edge transport barrier during the high-confinement mode (H-mode)

  14. High magnetic field MRI system

    International Nuclear Information System (INIS)

    Maeda, Hideaki; Urata, Masami; Satoh, Kozo

    1990-01-01

    A high field superconducting magnet, 4-5 T in central magnetic field, is required for magnetic resonance spectroscopic imaging (MRSI) on 31 P, essential nuclei for energy metabolism of human body. This paper reviews superconducting magnets for high field MRSI systems. Examples of the cross-sectional image and the spectrum of living animals are shown in the paper. (author)

  15. Predictions of toroidal rotation and torque sources arising in non-axisymmetric perturbed magnetic fields in tokamaks

    Science.gov (United States)

    Honda, M.; Satake, S.; Suzuki, Y.; Shinohara, K.; Yoshida, M.; Narita, E.; Nakata, M.; Aiba, N.; Shiraishi, J.; Hayashi, N.; Matsunaga, G.; Matsuyama, A.; Ide, S.

    2017-11-01

    Capabilities of the integrated framework consisting of TOPICS, OFMC, VMEC and FORTEC-3D, have been extended to calculate toroidal rotation in fully non-axisymmetric perturbed magnetic fields for demonstrating operation scenarios in actual tokamak geometry and conditions. The toroidally localized perturbed fields due to the test blanket modules and the tangential neutral beam ports in ITER augment the neoclassical toroidal viscosity (NTV) substantially, while do not significantly influence losses of beam ions and alpha particles in an ITER L-mode discharge. The NTV takes up a large portion of total torque in ITER and fairly decelerates toroidal rotation, but the change in toroidal rotation may have limited effectiveness against turbulent heat transport. The error field correction coils installed in JT-60SA can externally apply the perturbed fields, which may alter the NTV and the resultant toroidal rotation profiles. However, the non-resonant n=18 components of the magnetic fields arising from the toroidal field ripple mainly contribute to the NTV, regardless of the presence of the applied field by the coil current of 10 kA , where n is the toroidal mode number. The theoretical model of the intrinsic torque due to the fluctuation-induced residual stress is calibrated by the JT-60U data. For five JT-60U discharges, the sign of the calibration factor conformed to the gyrokinetic linear stability analysis and a range of the amplitude thereof was revealed. This semi-empirical approach opens up access to an attempt on predicting toroidal rotation in H-mode plasmas.

  16. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Purcell, J.R.; Chen, W.; Thomas, R.

    1976-01-01

    Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

  17. Helical variation of density profiles and fluctuations in the tokamak pedestal with applied 3D fields and implications for confinement

    Science.gov (United States)

    Wilcox, R. S.; Rhodes, T. L.; Shafer, M. W.; Sugiyama, L. E.; Ferraro, N. M.; Lyons, B. C.; McKee, G. R.; Paz-Soldan, C.; Wingen, A.; Zeng, L.

    2018-05-01

    Small 3D perturbations to the magnetic field in DIII-D ( δB /B ˜2 ×10-4 ) result in large modulations of density fluctuation amplitudes in the pedestal, which are shown using Doppler backscattering measurements to vary by a factor of 2. Helical perturbations of equilibrium density within flux surfaces have previously been observed in the pedestal of DIII-D plasmas when 3D fields are applied and were correlated with density fluctuation asymmetries in the pedestal. These intra-surface density and pressure variations are shown through two fluid MHD modeling studies using the M3D-C1 code to be due to the misalignment of the density and temperature equilibrium iso-surfaces in the pedestal region. This modeling demonstrates that the phase shift between the two iso-surfaces corresponds to the diamagnetic direction of the two species, with the mass density surfaces shifted in the ion diamagnetic direction relative to the temperature and magnetic flux iso-surfaces. The resulting pedestal density, potential, and turbulence asymmetries within flux surfaces near the separatrix may be at least partially responsible for several poorly understood phenomena that occur with the application of 3D fields in tokamaks, including density pump out and the increase in power required to transition from L- to H-mode.

  18. Evaluation of the toroidal torque driven by external non-resonant non-axisymmetric magnetic field perturbations in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Institute of Plasma Physics National Science Center “Kharkov Institute of Physics and Technology” ul. Akademicheskaya 1, 61108 Kharkov (Ukraine); Kernbichler, Winfried; Martitsch, Andreas F.; Heyn, Martin F. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Maassberg, Henning [Max-Planck Institut für Plasmaphysik, D-17491 Greifswald (Germany)

    2014-09-15

    The toroidal torque driven by external non-resonant magnetic perturbations (neoclassical toroidal viscosity) is an important momentum source affecting the toroidal plasma rotation in tokamaks. The well-known force-flux relation directly links this torque to the non-ambipolar neoclassical particle fluxes arising due to the violation of the toroidal symmetry of the magnetic field. Here, a quasilinear approach for the numerical computation of these fluxes is described, which reduces the dimension of a standard neoclassical transport problem by one without model simplifications of the linearized drift kinetic equation. The only limiting condition is that the non-axisymmetric perturbation field is small enough such that the effect of the perturbation field on particle motion within the flux surface is negligible. Therefore, in addition to most of the transport regimes described by the banana (bounce averaged) kinetic equation also such regimes as, e.g., ripple-plateau and resonant diffusion regimes are naturally included in this approach. Based on this approach, a quasilinear version of the code NEO-2 [W. Kernbichler et al., Plasma Fusion Res. 3, S1061 (2008).] has been developed and benchmarked against a few analytical and numerical models. Results from NEO-2 stay in good agreement with results from these models in their pertinent range of validity.

  19. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

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

  20. Comparative studies of stellarator and tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  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. High heat flux testing of CFC composites for the tokamak physics experiment

    Science.gov (United States)

    Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

    1996-10-01

    High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

  3. A new high sensitivity far-infrared laser interferometer for the HL-2A tokamak

    Science.gov (United States)

    Li, Y. G.; Zhou, Y.; Li, Y.; Deng, Z. C.; Wang, H. X.; Yi, J.

    2017-08-01

    A new four-chord Michelson-type formic acid (HCOOH, λ = 432.5 μm) laser interferometer has been successfully commissioned on the HL-2A tokamak to measure the electron density and density fluctuations. Due to the employment of the two-laser heterodyne technique, the time resolution of the interferometer reached 1.0 microseconds (μs). Four chords of line electron densities with a line-averaged density resolution 2 × 1016/m3 were obtained in a recent HL-2A experimental campaign, and detailed electron density fluctuations, caused by events such as edge localized mode, sawtooth precursor-oscillations, and energetic particle driven instabilities, were distinctly measured. In particular, the high-frequency electron density fluctuations (up to 500 kHz) caused by the reversed shear Alfvénic eigenmode were observed by the internal two interferometry channels, and their fluctuation location could be approximately identified from the spectra characteristics of multi-chord line electron densities.

  4. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

  5. High beta tokamak operation in DIII-D limited at low density/collisionality by resistive tearing modes

    International Nuclear Information System (INIS)

    La Haye, R.J.; Lao, L.L.; Strait, E.J.; Taylor, T.S.

    1997-01-01

    The maximum operational high beta in single-null divertor (SND) long pulse tokamak discharges in the DIII-D tokamak with a cross-sectional shape similar to the proposed International Thermonuclear Experimental Reactor (ITER) device is found to be limited by the onset of resistive instabilities that have the characteristics of neoclassically destabilized tearing modes. There is a soft limit due to the onset of an m/n=3/2 rotating tearing mode that saturates at low amplitude and a hard limit at slightly higher beta due to the onset of an m/n=2/1 rotating tearing mode that grows, slows down and locks. By operating at higher density and thus collisionality, the practical beta limit due to resistive tearing modes approaches the ideal magnetohydrodynamic (MHD) limit. (author). 15 refs, 4 figs

  6. Effects of fuelling by using high-pressure supersonic molecular beam in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Feng Beibin; Feng Zhen; Dong Jiafu; Li Wenzhong; Xu Deming; Hong Wenyu

    2002-01-01

    Supersonic molecular beam (SMB), as a new fuelling method, has been successfully developed and used in HL-1M tokamak and HT-7 superconducting tokamak. The hydrogen clusters have been found in the beam produced by high working-gas pressure in recent experiments. With a penetration depth of hydrogen particles greater than 17 cm, the rate of increase of electron density for SMB injection, dn e -bar/dt, approaches that of the small ice pellet injection. The plasma density increases step by step after multi-pulse SMB injection, just as multi-pellet fuelling. Comparison of fuelling effects was made between SMB and ice pellet injection on the same shot of ohmic discharge in HL-1M

  7. High field superconducting magnets

    Science.gov (United States)

    Hait, Thomas P. (Inventor); Shirron, Peter J. (Inventor)

    2011-01-01

    A superconducting magnet includes an insulating layer disposed about the surface of a mandrel; a superconducting wire wound in adjacent turns about the mandrel to form the superconducting magnet, wherein the superconducting wire is in thermal communication with the mandrel, and the superconducting magnet has a field-to-current ratio equal to or greater than 1.1 Tesla per Ampere; a thermally conductive potting material configured to fill interstices between the adjacent turns, wherein the thermally conductive potting material and the superconducting wire provide a path for dissipation of heat; and a voltage limiting device disposed across each end of the superconducting wire, wherein the voltage limiting device is configured to prevent a voltage excursion across the superconducting wire during quench of the superconducting magnet.

  8. Comparative study of fundamental and second-harmonic ICRF wave propagation and damping at high density in the Alcator tokamak

    International Nuclear Information System (INIS)

    Gaudreau, M.P.J.

    1981-09-01

    Due to the versatility of the high power apparatus, the fast magnetosonic branch is used with ω 0 = 1,2,3,4 ω/sub ci/, unlike most other ICRF experiments. Unusually high magnetic field (B 0 = 40 to 80 kG), plasma density (n/sub e/ = 10 13 - 5 x 10 14 /cm 3 ), generator frequency (f 0 = 90 to 200 MHz) and transmitter power, with shielded and unshielded antennas, are the key parameters of the experiment. This wide parameter range allows a direct comparison between fundamental and second harmonic regimes, and shielded and unshielded antennas, our prime goals. The real and imaginary parts of the parallel and perpendicular wave numbers are measured with extensive magnetic probe diagnostics for a spectrum of plasma parameters and compared with theory. Qualitative and quantitative evaluations of the wave structure and scaling laws are derived analytically in simple geometries and computed numerically for realistic plasma parameters and profiles. General figures of merit, such as radiation resistance and quality factor, are also derived and compared with the experiment. Secondary effects of the high power wave launching, such as changes in plasma current, density, Z/sub eff/, energetic neutral flux, soft x-rays, neutron flux, and impurities are also discussed. Most important, a general synthesis of the many engineering, physics, and experimental problems and conclusions of the Alcator A ICRF program are inspected in detail. Finally, the derived and experimentally determined scaling laws and engineering constraints are used to estimate the ICRF requrements, advantages, and potential pitfalls of the next generations of experiments on the Alcator tokamaks

  9. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

    International Nuclear Information System (INIS)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.

    2015-01-01

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started

  10. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Capece, A.; Koel, B.; Roszell, J. [Princeton University, Princeton, New Jersey 08544 (United States); Biewer, T. M.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2015-05-15

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

  11. 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

  12. Discharge cleaning for a tokamak

    International Nuclear Information System (INIS)

    Ishii, Shigeyuki

    1983-01-01

    Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)

  13. Effects of impurities and magnetic divertors on high-temperature tokamaks

    International Nuclear Information System (INIS)

    Meade, D.M.; Furth, H.P.; Rutherford, P.H.; Seidl, F.G.P.; Duechs, D.F.

    1974-10-01

    A one-dimensional tokamak plasma transport code has been adapted to include impurity influx, stripping, radiation, and diffusion, as well as the usual processes of hydrogen plasma and heat transport, recycling at the boundary, and multigeneration charge-exchange. Neutral-beam heating, adiabatic compression, and divertor boundary conditions are included as optional features. Illustrative computations are given for present-day and next-generation tokamaks. The problems of impurity control are discussed, and two technical approaches are examined in greater detail: the transient cold-plasma shield, and the poloidal divertor. (auth)

  14. Recent developments in Bayesian inference of tokamak plasma equilibria and high-dimensional stochastic quadratures

    International Nuclear Information System (INIS)

    Von Nessi, G T; Hole, M J

    2014-01-01

    We present recent results and technical breakthroughs for the Bayesian inference of tokamak equilibria using force-balance as a prior constraint. Issues surrounding model parameter representation and posterior analysis are discussed and addressed. These points motivate the recent advancements embodied in the Bayesian Equilibrium Analysis and Simulation Tool (BEAST) software being presently utilized to study equilibria on the Mega-Ampere Spherical Tokamak (MAST) experiment in the UK (von Nessi et al 2012 J. Phys. A 46 185501). State-of-the-art results of using BEAST to study MAST equilibria are reviewed, with recent code advancements being systematically presented though out the manuscript. (paper)

  15. Extremely fast vertical displacement event induced by a plasma βp collapse in high βp tokamak disruptions

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

    1996-05-01

    In a vertically elongated (κ ∼ 1.5), high β p (β p ∼ 1.7) tokamak with a resistive shell, extremely fast vertical displacement events (VDE's) induced by a model of strong β p collapse were found through computer simulations using the Tokamak Simulation Code. Although the plasma current quench, which had been shown to be the prime cause of VDE's in a relatively low β p tokamak (β p ∼ 0.2), was not observed during the VDE evolution, the observed growth rate of VDE's was almost five times (γ ∼ 655 sec -1 ) faster than the growth rate of the usual positional instability (γ ∼ 149 sec -1 ). The essential mechanism of the β p collapse-induced VDE was clarified to be the significant destabilization of positional instability due to a large and sudden degradation of the decay n-index in addition to a reduction of the stability index n s . It is pointed out that the shell-geometry characterizes the VDE dynamics, and that the VDE rate depends strongly both on the magnitude of the β p collapse and the n-index of the equilibria just before the β p collapse occurs. A new guide line for designing the fusion reactor is proposed with considering the impact of disruptions. (author)

  16. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  18. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

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

  19. Study on the characters of high voltage charging power supply system for diagnostics neutral beam on HT-7 Tokamak

    International Nuclear Information System (INIS)

    Zhang Jian; Huang Yiyun; Liu Baohua; Guo Wenjun; Shen Xiaoling; Wei Wei

    2011-01-01

    A high voltage power supply system has been developed for the diagnostic neutral beam on the HT-7 experimental Tokamak, and the over-voltage phenomenon of storage capacitor was founded in the experiment. In order to analyse and resolve this problem, the structure and principle of high voltage power supply is described and the primary high voltage charging power supply system is introduced in detail. The phenomenon of over-voltage on the capacitors is also studied with circuit model, and the conclusion is obtained that the leakage inductance is the mA in reason which causes the over-voltage on the capacitors. (authors)

  20. Excitation of short wavelength Alfven oscillations by high energy ions in tokamak

    International Nuclear Information System (INIS)

    Beasley, C.O. Jr.; Lominadze, J.G.; Mikhailovskii, A.B.

    1975-08-01

    The excitation of Alfven waves by fast untrapped ions in axisymmetric tokamaks is described by the dispersion relation epsilon 11 - c 2 k/sub parallel bars/ 2 /ω 2 = 0. Using this relation a new class of instability connected with the excitation of Alfven oscillations is described. (U.S.)

  1. Boundary plasma heat flux width measurements for poloidal magnetic fields above 1 Tesla in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Alcator C-Mod Team

    2017-10-01

    The boundary heat flux width, along with the total power flowing into the boundary, sets the power exhaust challenge for tokamaks. A multi-machine boundary heat flux width database found that the heat flux width in H-modes scaled inversely with poloidal magnetic field (Bp) and was independent of machine size. The maximum Bp in the database was 0.8 T, whereas the ITER 15 MA, Q =10 scenario will be 1.2 T. New measurements of the boundary heat flux width in Alcator C-Mod extend the international database to plasmas with Bp up to 1.3 T. C-Mod was the only experiment able to operate at ITER-level Bp. These new measurements are from over 300 plasma shots in L-, I-, and EDA H-modes spanning essentially the whole operating space in C-Mod. We find that the inverse-Bp dependence of the heat flux width in H-modes continues to ITER-level Bp, further reinforcing the empirical projection of 500 μm heat flux width for ITER. We find 50% scatter around the inverse-Bp scaling and are searching for the `hidden variables' causing this scatter. Supported by USDoE award DE-FC02-99ER54512.

  2. Resonance localization and poloidal electric field due to cyclo- tron wave heating in tokamak plasmas

    International Nuclear Information System (INIS)

    Hsu, J.Y.; Chan, V.S.; Harvey, R.W.; Prater, R.; Wong, S.K.

    1984-01-01

    The perpendicular heating in cyclotron waves tends to pile up the resonant particles toward the low magnetic field side with their banana tips localized to the resonant surface. A poloidal electric field with an E x B drift comparable to the ion vertical drift in a toroidal magnetic field may result. With the assumption of anomalous electron and neoclassical ion transport, density variations due to wave heating are discussed

  3. Three-dimensional tokamak equilibria in the presence of resonant field errors

    International Nuclear Information System (INIS)

    Reiman, A.; Monticello, D.

    1992-01-01

    Numerical solutions are described for three-dimensional MHD equilibria in the presence of resonant magnetic field perturbations. The effects of a realistic spectrum of resonant field errors are calculated for a range of current profiles. It is found that field errors of the magnitude existing in present day devices, and contemplated for future devices, can produce a set of magnetic islands occupying a significant fraction of the plasma cross-section

  4. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1996-06-01

    A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

  5. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  6. Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II

    International Nuclear Information System (INIS)

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

    1992-03-01

    In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

  7. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10 15 cm -3 , high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  8. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  9. Magnetic ''islandography'' in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Waddell, B.V.; Hicks, H.R.

    1978-09-01

    Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

  10. First measurement of poloidal-field-induced Faraday rotation in a tokamak plasma

    International Nuclear Information System (INIS)

    Kunz, W.; Association Euratom-CEA sur la Fusion, Centre d'Etudes Nucleaires de Fontenay-aux-Roses, 92

    1978-01-01

    Faraday rotation measurements using a ferrite modulation technique were performed on one channel of the 337 μm-interferometer on TFR. The experiment is intended as a preparatory step towards poloidal-field determination on the basis of the Faraday effect in a multi-channel configuration. The technical feasibility of precise Faraday rotation measurements under machine conditions is demonstrated. The measured rotation is unambiguously due to the poloidal magnetic field and agrees fairly with what can be estimated. (author)

  11. 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

  12. Field load and displacement boundary condition computer program used for the finite element analysis and design of toroidal field coils in a tokamak

    International Nuclear Information System (INIS)

    Smith, R.A.

    1975-06-01

    The design evaluation of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX) and the Tokamak Fusion Test Reactor (TFTR) has been performed by structural analysis with the finite element method. The technique employed has been simplified with supplementary computer programs that are used to generate the input data for the finite element computer program. Significant automation has been provided by computer codes in three areas of data input. These are the definition of coil geometry by a mesh of node points, the definition of finite elements via the node points and the definition of the node point force/displacement boundary conditions. The computer programs by name that have been used to perform the above functions are PDXNODE, ELEMENT and PDXFORC. The geometric finite element modeling options for toroidal field coils provided by PDXNODE include one-fourth or one-half symmetric sections of circular coils, oval shaped coils or dee-shaped coils with or without a beveled wedging surface. The program ELEMENT which defines the finite elements for input to the finite element computer code can provide considerable time and labor savings when defining the model of coils of non-uniform cross-section or when defining the model of coils whose material properties are different in the R and THETA directions due to the laminations of alternate epoxy and copper windings. The modeling features provided by the program ELEMENT have been used to analyze the PLT and the TFTR toroidal field coils with integral support structures. The computer program named PDXFORC is described. It computes the node point forces in a model of a toroidal field coil from the vector crossproduct of the coil current and the magnetic field. The model can be of one-half or one-fourth symmetry to be consistent with the node model defined by PDXNODE, and the magnetic field is computed from toroidal or poloidal coils

  13. High-Field Accelerator Magnets

    International Nuclear Information System (INIS)

    Rijk, G de

    2014-01-01

    In this lecture an overview is given of the present technology for high field accelerator magnets. We indicate how to get high fields and what are the most important parameters. The available conductors and their limitations are presented followed by the most relevant types of coils and support structures. We conclude by showing a number of recent examples of development magnets which are either pure R&D objects or models for the LHC luminosity upgrade

  14. On the role of poloidal and toroidal fluctuating electric fields in tokamak transport

    International Nuclear Information System (INIS)

    Isichenko, M.B.; Wootton, A.J.

    1995-01-01

    The two different expressions for the radial particle flux Λ found in the literature, as given by equations (2) and (4), are identical if the parallel electric field is small. The first expression is derivable in a fluid approach, whereas the second follows from the analysis of individual particle orbits. These expressions, without change, are also valid for an arbitrary axisymmetric magnetic geometry. In a situation where the parallel electric field is significant, the more accurate expression for the particle flux is in terms of the standard E x B velocity

  15. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1987-10-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared to conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA . cm -2 (at 77 K and ∼10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA . cm -2 most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of ∼7% are forecast for the overall capital cost of the power plant in the best case. An additional ∼3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

  16. Reaction-rate coefficients, high-energy ions slowing-down, and power balance in a tokamak fusion reactor plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo

    1978-07-01

    Described are the reactivity coefficient of D-T fusion reaction, slowing-down processes of deuterons injected with high energy and 3.52 MeV alpha particles generated in D-T reaction, and the power balance in a Tokamak reactor plasma. Most of the results were obtained in the first preliminary design of JAERI Experimental Fusion Reactor (JXFR) driven with stationary neutral beam injection. A manual of numerical computation program ''BALTOK'' developed for the calculations is given in the appendix. (auth.)

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

    Czech Academy of Sciences Publication Activity Database

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

    2012-01-01

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

  18. Plasma equilibrium profiles with applied resonant fields on TBR-1 tokamak

    International Nuclear Information System (INIS)

    Castro, R.M. de; Heller, M.V.A.P.; Caldas, I.L.; Silva, R.P. da; Brasilio, Z.A.; Oda, G.A.

    1995-01-01

    In this work we present the measurements of the plasma potential, in the edge and in the scrape-off layer regions of plasma, with and without the presence of the magnetic field perturbations produced by resonant helical windings. (author). 6 refs., 6 figs

  19. Edge pedestal characteristics in JET and JT-60U tokamaks under variable toroidal field ripple

    NARCIS (Netherlands)

    Urano, H.; Saibene, G.; Oyama, N.; Parail, V.; P. de Vries,; Sartori, R.; Kamada, Y.; Kamiya, K.; Loarte, A.; Lonnroth, J.; Sakamoto, Y.; Salmi, A.; Shinohara, K.; Takenaga, H.; Yoshida, M.

    2011-01-01

    The effects of toroidal field (TF) ripple on the edge pedestal characteristics were examined in the TF ripple scan experiments at the plasma current I(p) of 1.1 MA in JET and JT-60U. The TF ripple amplitude delta(R) was defined as a value averaged over the existing ripple wells at the separatrix on

  20. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

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

  1. Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

    International Nuclear Information System (INIS)

    Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T.; Nakayama, T.; Hasegawa, M.; Urata, K.

    1998-01-01

    In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude (δ) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The δ depends on the toroidal magnetic field. The value of B r21 /B t in the case of displacement of few cm is about 1 x 10 -5 which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

  2. Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

    Energy Technology Data Exchange (ETDEWEB)

    Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T. [Japan Atomic Energy Research Inst. (Japan); Nakayama, T. [Hitachi Ltd. (Japan); Hasegawa, M. [Mitsubishi Electric Corp. (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan)

    1998-07-01

    In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude ({delta}) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The {delta} depends on the toroidal magnetic field. The value of B{sub r21} /B{sub t} in the case of displacement of few cm is about 1 x 10{sup -5} which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

  3. The role of the radial electric field in confinement and transport in H-mode and VH-mode discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Gohil, P.; Burrell, K.H.; Groebner, R.J.; Osborne, T.H.; Doyle, E.J.; Rettig, C.L.

    1993-08-01

    Measurements of the radial electric field, E r , with high spatial and high time resolution in H-mode and VH-mode discharges in the DIII-D tokamak have revealed the significant influence of the shear in E r on confinement and transport in these discharges. These measurements are made using the DIII-D Charge Exchange Recombination (CER) System. At the L-H transition in DIII-D plasmas, a negative well-like E r profile develops just within the magnetic separatrix. A region of shear in E r results, which extends 1 to 2 cm into the plasma from the separatrix. At the transition, this region of sheared E r exhibits the greatest increase in impurity ion poloidal rotation velocity and the greatest reduction in plasma fluctuations. A transport barrier is formed in this same region of E x B velocity shear as is signified by large increases in the observed gradients of the ion temperature, the carbon density, the electron temperature and electron density. The development of the region of sheared E r , the increase in impurity ion poloidal rotation, the reduction in plasma turbulence, and the transport barrier all occur simultaneously at the L-H transition. Measurements of the radial electric field, plasma turbulence, thermal transport, and energy confinement have been performed for a wide range of plasma conditions and configurations. The results support the supposition that the progression of improving confinement at the L-H transition, into the H-mode and then into the VH-mode can be explained by the hypothesis of the suppression of plasma turbulence by the increasing penetration of the region of sheared E x B velocity into the plasma interior

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Development of hard X-ray spectrometer with high time resolution on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ma, T.K.; Chen, Z.Y., E-mail: zychen@hust.edu.cn; Huang, D.W.; Tong, R.H.; Yan, W.; Wang, S.Y.; Dai, A.J.; Wang, X.L.

    2017-06-01

    A hard X-ray (HXR) spectrometer has been developed to study the runaway electrons during the sawtooth activities and during the runaway current plateau phase on the J-TEXT tokamak. The spectrometer system contains four NaI scintillator detectors and a multi-channel analyzer (MCA) with 0.5 ms time resolution. The dedicated peak detection circuit embedded in the MCA provides a pulse height analysis at count rate up to 1.2 million counts per second (Mcps), which is the key to reach the high time resolution. The accuracy and reliability of the system have been verified by comparing with the hardware integrator of HXR flux. The temporal evolution of HXR flux in different energy ranges can be obtained with high time resolution by this dedicated HXR spectrometer. The response of runaway electron transport with different energy during the sawtooth activities can be studied. The energy evolution of runaway electrons during the plateau phase of runaway current can be obtained. - Highlights: • A HXR spectrometer with high time resolution has been developed on J-TEXT tokamak. • The response of REs transport during the sawtooth activities can be investigated. • The energy evolution of REs following the disruptions can be monitored.

  6. Realization of toroidal field power supply control system for J-TEXT tokamak

    International Nuclear Information System (INIS)

    Qiu Shengshun; Zhuang Ge; Zhang Ming; Feng Jianming

    2009-01-01

    Based on the integrated development environment provided by QNX real-time operation system, the control system of toroidal field power supply is designed and developed. The system is proved to be reliable, stable and in real-time. It can control the power supply successfully to produce a constant current up to 92.5kA lasting for 1s and 1.74T at the magnetic axis. In conclusion, the control system can meet the requirements of the J-TEXT routine operation at present. (authors)

  7. An algorithm for the calculation of three-dimensional ICRF fields in tokamak geometry

    International Nuclear Information System (INIS)

    Smithe, D.N.; Kammash, T.

    1987-01-01

    A computational scheme is developed which permits tractable calculation of three-dimensional full-wave solutions to the Vlasov-Maxwell equations under typical ion cyclotron range of frequencies (ICRF) experimental conditions. The method is unique in that power deposition to the plasma is determined via the anti-Hermitian part of a truncated warm plasma dielectric operator, rather than as the result of an assumed phenomenological collision frequency. The resulting computer code allows arbitrary variation of density, temperature, magnetic field and minority concentration in the poloidal plane by performing a convolution of poloidal modes to produce a coupled system of differential equations in the radial variable. By assuming no inhomogeneity along the toroidal axis, an inverse transform over k parallel is performed, yielding the global three-dimensional fast wave field solutions. The application of the code to TFTR-like plasmas shows a mild resonance structure in antenna loading related to the changing number of wavelengths between the antenna and the resonance layer. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

    International Nuclear Information System (INIS)

    Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

    2006-01-01

    The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

  10. Experimental measurement of magnetic field null in the vacuum chamber of KTM tokamak based on matrix of 2D Hall sensors

    Energy Technology Data Exchange (ETDEWEB)

    Shapovalov, G.; Chektybayev, B., E-mail: chektybaev@nnc.kz; Sadykov, A.; Skakov, M.; Kupishev, E.

    2016-11-15

    Experimental technique of measurement of magnetic field null region inside of the KTM tokamak vacuum chamber has been developed. Square matrix of 36 2D Hall sensors, which used in the technique, allows carrying out direct measurements of poloidal magnetic field dynamics in the vacuum chamber. To better measuring accuracy, Hall sensor’s matrix was calibrated with commercial Helmholtz coils and in situ measurement of defined magnetic field from poloidal and toroidal coils. Standard KTM Data-Acquisition System has been used to collect data from Hall sensors. Experimental results of measurement of magnetic field null in the vacuum chamber of KTM are shown in the paper. Additionally results of the magnetic field null reconstruction from signals of inductive total flux loops are shown in the paper.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  13. Advanced control scenario of high-performance steady-state operation for JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    Tamai, H.; Kurita, G.; Matsukawa, M.; Urata, K.; Sakurai, S.; Tsuchiya, K.; Morioka, A.; Miura, Y.M.; Kizu, K.; Kamada, Y.; Sakasai, A.; Ishida, S.

    2004-01-01

    Plasma control on high-β N steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-β N exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected. (authors)

  14. Multi-scale full-orbit analysis on phase-space behavior of runaway electrons in tokamak fields with synchrotron radiation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yulei; Liu, Jian, E-mail: jliuphy@ustc.edu.cn [School of Nuclear Science and Technology and Department of Modern Physics, University of Science and Technology of China, Hefei, Anhui 230026 (China); Key Laboratory of Geospace Environment, CAS, Hefei, Anhui 230026 (China); Qin, Hong [School of Nuclear Science and Technology and Department of Modern Physics, University of Science and Technology of China, Hefei, Anhui 230026 (China); Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    2016-06-15

    In this paper, the secular full-orbit simulations of runaway electrons with synchrotron radiation in tokamak fields are carried out using a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different dynamical timescales spanning 11 orders. In the small timescale, i.e., the characteristic timescale imposed by Lorentz force, the severely deformed helical trajectory of energetic runaway electron is witnessed. A qualitative analysis of the neoclassical scattering, a kind of collisionless pitch-angle scattering phenomena, is provided when considering the coupling between the rotation of momentum vector and the background magnetic field. In large timescale up to 1 s, it is found that the initial condition of runaway electrons in phase space globally influences the pitch-angle scattering, the momentum evolution, and the loss-gain ratio of runaway energy evidently. However, the initial value has little impact on the synchrotron energy limit. It is also discovered that the parameters of tokamak device, such as the toroidal magnetic field, the loop voltage, the safety factor profile, and the major radius, can modify the synchrotron energy limit and the strength of neoclassical scattering. The maximum runaway energy is also proved to be lower than the synchrotron limit when the magnetic field ripple is considered.

  15. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Shukla, G.; Ghosh, J.; Melich, Radek; Pánek, Radomír; Tomeš, Matěj; Imríšek, Martin; Naydenkova, Diana; Varju, Jozef; Pereira, T.; Gomes, R.; Abramovic, I.; Jaspers, R.; Písařík, M.; Odstrčil, T.; Van Oost, G.

    96-97, October (2015), s. 1006-1011 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR(CZ) GA14-35260S; GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Plasma spectroscopy * Plasma rotation * Ion temperature * CXRS Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002355

  16. Energy composition of high-energy neutral beams on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Mitošinková, Klára; Stöckel, Jan; Varju, Jozef; Weinzettl, Vladimír

    2016-01-01

    Roč. 61, č. 4 (2016), s. 419-423 ISSN 0029-5922. [Summer School of Plasma Diagnostics PhDiaFusion 2015: “Soft X-ray Diagnostics for Fusion Plasma”. Bezmiechowa, 16.06.2015-20.06.2015] R&D Projects: GA MŠk(CZ) LM2011021; GA MŠk(CZ) 8D15001 Institutional support: RVO:61389021 Keywords : tokamak * neutral beam injection (NBI) * Doppler effect * beam composition * beam composition Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.760, year: 2016 http://www.ichtj.waw.pl/nukleonikaa/?p=1256

  17. 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

  18. The primary results for the mixed carbon material used for high flux steady-state tokamak operation in China

    International Nuclear Information System (INIS)

    Guo, Q.G.; Li, J.G.; Zhai, G.T.; Liu, L.; Song, J.R.; Zhang, L.F.; He, Y.X.; Chen, J.L.

    2001-01-01

    Several types of carbon mixed materials have been developed in China to be used for high flux steady-state tokamak operation. Performance evaluation of these materials is necessary to determine their applicability as PFCs for high flux steady state. This paper describes the primary results of carbon mixed materials and the effects of dopants on properties are primarily discussed. Test results reveal that bulk boronized graphite has excellent physical and mechanical properties while their thermal conductivity is no more than 73 W/m K due to the formation of a uniform boron-carbon solid solution. In case of multi-element doped graphite, titanium dopant or a decreased boron content is favorable to enhance thermal conductivity. A kind of doped graphite has been developed with thermal conductivity as high as 278 W/m K by optimizing the compositions. Correlations among compositions, microstructure and properties of such doped graphite are discussed

  19. Simulation models for tokamak plasmas

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. 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

  1. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1988-08-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA /center dot/ cm/sup /minus/2/ (at 77 K and /approximately/10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA /center dot/ cm/sup /minus/2/ most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of /approximately/7% are forecast for the overall capital cost of the power plant in the best case. An additional /approximately/3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. 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

  4. The disparate impact of the ion temperature gradient and the density gradient on edge transport and the low-high transition in tokamaks

    International Nuclear Information System (INIS)

    Kleva, Robert G.; Guzdar, Parvez N.

    2009-01-01

    Steepening of the ion temperature gradient in nonlinear fluid simulations of the edge region of a tokamak plasma causes a rapid degradation in confinement. As the density gradient steepens, there is a continuous improvement in confinement analogous to the low (L) to high (H) transition observed in tokamaks. In contrast, as the ion temperature gradient steepens, there is a rapid increase in the particle and energy fluxes and no L-H transition. For a given pressure gradient, confinement always improves when more of the pressure gradient arises from the density gradient, and less of the pressure gradient arises from the ion temperature gradient.

  5. Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Casali, Livia

    2015-11-24

    Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the f{sub ELM} is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be

  6. Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II; Calculo de modificacion al campo magnetico toroidal del Tokamak nivillo. Parte II

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-15

    In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

  7. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  8. Study on assembly techniques and procedures for ITER tokamak device

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

    2006-06-01

    The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

  9. Development and performance of high speed processing system of magnetohydrodynamic equilibria for discharge analyses on the J T-60 tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Yukihiro; Nakamura, Yukiharu; Shirai, Hiroshi; Hamamatsu, Kiyotaka; Harada, Yoshio; Kikuchi, Mitsuru; Nakata, Yoshihiro

    1999-01-01

    In order to provide a set of magnetohydrodynamic (MHD) equilibrium database which is indispensable for both the studies on improvement of energy confinement and stabilization of MHD activities in tokamaks, a high speed data-processing system synchronizing with J T-60 discharge sequence was newly developed by utilizing the latest model of hugh speed workstation and by optimizing the parallel processing technique to perform fast calculation of MHD equilibria. This high speed system was found to have a sufficient ability to complete the whole equilibrium calculations during each inter-shot period. Cooperating with the mass data storage subsystem preserving the latest equilibrium database automatically, the animated discharge monitoring subsystem provides valuable information for the J T-60 operator to determine control parameters of the succeeding discharge. This report describes the system performance realized in the J T-60 experiment. (author)

  10. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  11. Classical tokamak transport theory

    International Nuclear Information System (INIS)

    Nocentini, Aldo

    1982-01-01

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

  12. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-22

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided.

  13. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    International Nuclear Information System (INIS)

    1995-01-01

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided

  14. Low–intermediate–high confinement transition in HL-2A tokamak plasmas

    International Nuclear Information System (INIS)

    Cheng, J.; Dong, J.Q.; Yan, L.W.; Hong, W.Y.; Zhao, K.J.; Huang, Z.H.; Ji, X.Q.; Zhong, W.L.; Yu, D.L.; Nie, L.; Song, X.M.; Yang, Q.W.; Ding, X.T.; Duan, X.R.; Liu, Yong; Itoh, K.; Itoh, S.-I.; Zou, X.L.

    2014-01-01

    The dynamics of low–intermediate–high confinement transitions was studied using a four-step Langmuir probe in the HL-2A edge plasma. Two types (dubbed type-Y and type-J) of limit cycle oscillations (LCOs) with opposite temporal ordering between the radial electric field and turbulence were first observed. In type-Y, the turbulence grows first, followed by the localized electric field. In contrast, the electric field leads turbulence in type-J. In addition, the Reynolds stress gradient is found not enough to drive the LCO flow and the three-wave nonlinear coupling is weak there. The continuously increasing amplitude of magnetic fluctuations and the significant correlation between the magnetic fluctuation and the electron pressure gradient indicate an important role of diamagnetic drifts in the L–H transition. Mode numbers of magnetic fluctuations in the LCO frequency are identified to be m/n = 1/0. (paper)

  15. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  16. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  17. Effects of classical and neo-classical cross-field transport of tungsten impurity in realistic tokamak geometry

    Energy Technology Data Exchange (ETDEWEB)

    Yamoto, S.; Inoue, H.; Sawada, Y.; Hatayama, A. [Faculty of Science and Technology, Keio University, Yokohama (Japan); Homma, Y.; Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Bonnin, X. [ITER Organization, St. Paul Lez Durance (France); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Schneider, R. [Ernst-Moritz-Arndt University Greifswald (Germany)

    2016-08-15

    The initial simulation study of the neoclassical perpendicular self-diffusion transport in the SOL/Divertor regions for a realistic tokamak geometry with the IMPGYRO code has been performed in this paper. One of the most unique features of the IMPGYRO code is calculating exact Larmor orbit of the test particle instead of assuming guiding center approximation. Therefore, effects of the magnetic drifts in realistic tokamaks are naturally taken into account in the IMPGYRO code. This feature makes it possible to calculate neoclassical transport processes, which possibly become large in the SOL/divertor plasma. Indeed, neoclassical self-diffusion process, the resultant effect of the combination of magnetic drift and Coulomb collisions with background ions, has already been included in the IMPGYRO model. In the present paper, prior to implementing the detailed model of neoclassical transport process into IMPGYRO, we have investigated the effect of neoclassical selfdiffusion in a realistic tokamak geometry with lower single null X-point. We also use a model with guiding center approximation in order to compare with the IMPGYRO full orbit model. The preliminary calculation results of each model have shown differences in the perpendicular average velocity of impurity ions at the top region of the SOL. The mechanism which leads to the difference has been discussed. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  18. Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

    Directory of Open Access Journals (Sweden)

    Lohr John

    2017-01-01

    Full Text Available Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

  19. Detector array for measurement of high-frequency fluctuations in visible and near-UV emission from tokamaks

    International Nuclear Information System (INIS)

    Hurwitz, P.D.; Hall, B.F.; Rowan, W.L.

    1992-01-01

    We developed an imaging detector to measure high-frequency fluctuations in visible and near-UV emission from tokamaks. The detector is intended for the study of plasma turbulence, mhd phenomena, and edge-localized modes. Particularly in the first two applications, it will complement existing techniques by providing higher spatial resolution as well as measurement capability in otherwise inaccessible regions of the plasma. The device consists of an optical system, a linear array of 32 photodiodes, and an amplifier for each photodiode. The amplifiers have a transimpedance gain of 10 5 --10 6 and the frequency response is flat to 100 kHz. Experience with this device has shown that optical imaging systems can be easily designed and tailored to a specific measurement because of the small size and close spacing of the individual light-sensitive elements. The device has been successfully tested on TEXT-U in measurements of H α fluctuations

  20. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

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

  1. Quantify Plasma Response to Non-Axisymmetric (3D) Magnetic Fields in Tokamaks, Final Report for FES (Fusion Energy Sciences) FY2014 Joint Research Target

    International Nuclear Information System (INIS)

    Strait, E. J.; Park, J. K.; Marmar, E. S.; Ahn, J. W.; Berkery, J. W.; Burrell, K. H.; Canik, J. M.; Delgado-Aparicio, L.; Ferraro, N. M.; Garofalo, A. M.; Gates, D. A.; Greenwald, M.; Kim, K.; King, J. D.; Lanctot, M. J.; Lazerson, S. A.; Liu, Y. Q.; Lore, J. D.; Menard, J. E.; Nazikian, R.; Shafer, M. W.; Paz-Soldan, C.; Reiman, A. H.; Rice, J. E.; Sabbagh, S. A.; Sugiyama, L.; Turnbull, A. D.; Volpe, F.; Wang, Z. R.; Wolfe, S. M.

    2014-01-01

    The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10 -4 of the main axisymmetric field, such ''3D'' fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data

  2. Quantify Plasma Response to Non-Axisymmetric (3D) Magnetic Fields in Tokamaks, Final Report for FES (Fusion Energy Sciences) FY2014 Joint Research Target

    Energy Technology Data Exchange (ETDEWEB)

    Strait, E. J. [General Atomics, San Diego, CA (United States); Park, J. -K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Marmar, E. S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ahn, J. -W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Berkery, J. W. [Columbia Univ., New York, NY (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Canik, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delgado-Aparicio, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. M. [General Atomics, San Diego, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Greenwald, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kim, K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); King, J. D. [General Atomics, San Diego, CA (United States); Lanctot, M. J. [General Atomics, San Diego, CA (United States); Lazerson, S. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, Y. Q. [Culham Science Centre, Abingdon (United Kingdom). Euratom/CCFE Association; Logan, N. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lore, J. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Menard, J. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Nazikian, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Shafer, M. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Paz-Soldan, C. [General Atomics, San Diego, CA (United States); Reiman, A. H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Rice, J. E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Sabbagh, S. A. [Columbia Univ., New York, NY (United States); Sugiyama, L. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Turnbull, A. D. [General Atomics, San Diego, CA (United States); Volpe, F. [Columbia Univ., New York, NY (United States); Wang, Z. R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Wolfe, S. M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2014-09-30

    The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10-4 of the main axisymmetric field, such “3D” fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data, in

  3. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

    Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

  4. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  5. High performance gamma-ray spectrometer for runaway electron studies on the FT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Shevelev, A.E., E-mail: Shevelev@cycla.ioffe.ru [Ioffe Institute, Politekhnicheskaya 26, St. Petersburg 194021 (Russian Federation); Khilkevitch, E.M.; Lashkul, S.I.; Rozhdestvensky, V.V.; Altukhov, A.B.; Chugunov, I.N.; Doinikov, D.N.; Esipov, L.A.; Gin, D.B.; Iliasova, M.V.; Naidenov, V.O.; Nersesyan, N.S.; Polunovsky, I.A.; Sidorov, A.V. [Ioffe Institute, Politekhnicheskaya 26, St. Petersburg 194021 (Russian Federation); Kiptily, V.G. [CCFE, Culham Science Centre, Abingdon, Oxon X14 3DB (United Kingdom)

    2016-09-11

    A gamma-ray spectrometer based on LaBr{sub 3}(Ce) scintillator has been used for measurements of hard X-ray emission generated by runaway electrons in the FT-2 tokamak plasmas. Using of the fast LaBr{sub 3}(Ce) has allowed extending count rate range of the spectrometer by a factor of 10. A developed digital processing algorithm of the detector signal recorded with a digitizer sampling rate of 250 MHz has provided a pulse height analysis at count rates up to 10{sup 7} s{sup −1}. A spectrum deconvolution code DeGaSum has been applied for inferring the energy distribution of runaway electrons escaping from the plasma and interacting with materials of the FT-2 limiter in the vacuum chamber. The developed digital signal processing technique for LaBr{sub 3}(Ce) spectrometer has allowed studying the evolution of runaways energy distribution in the FT-2 plasma discharges with time resolution of 1–5 ms.

  6. Electron temperature profiles in high power neutral-beam-heated TFTR [Tokamak Fusion Test Reactor] plasmas

    International Nuclear Information System (INIS)

    Taylor, G.; Grek, B.; Stauffer, F.J.; Goldston, R.J.; Fredrickson, E.D.; Wieland, R.M.; Zarnstorff, M.C.

    1987-09-01

    In 1986, the maximum neutral beam injection (NBI) power in the Tokamak Fusion Test Reactor (TFTR) was increased to 20 MW, with three beams co-parallel and one counter-parallel to I/sub p/. TFTR was operated over a wide range of plasma parameters; 2.5 19 19 m -3 . Data bases have been constructed with over 600 measured electron temperature profiles from multipoint TV Thomson scattering which span much of this parameter space. We have also examined electron temperature profile shapes from electron cyclotron emission at the fundamental ordinary mode and second harmonic extraordinary mode for a subset of these discharges. In the light of recent work on ''profile consistency'' we have analyzed these temperature profiles in the range 0.3 < (r/a) < 0.9 to determine if a profile shape exists which is insensitive to q/sub cyl/ and beam-heating profile. Data from both sides of the temperature profile [T/sub e/(R)] were mapped to magnetic flux surfaces [T/sub e/(r/a)]. Although T/sub e/(r/a), in the region where 0.3 < r/a < 0.9 was found to be slightly broader at lower q/sub cyl/, it was found to be remarkably insensitive to β/sub p/, to the fraction of NBI power injected co-parallel to I/sub p/, and to the heating profile going from peaked on axis, to hollow. 10 refs., 8 figs

  7. Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E ×B Flow During High Performance DIII-D Tokamak Discharges

    Science.gov (United States)

    Barada, K.; Rhodes, T. L.; Burrell, K. H.; Zeng, L.; Bardóczi, L.; Chen, Xi; Muscatello, C. M.; Peebles, W. A.

    2018-03-01

    A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high performance DIII-D tokamak plasma discharges. These LCOs are localized and composed of density turbulence, gradient drives, and E ×B velocity shear damping (E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E ×B velocity shear. Reported here for the first time is a unique spatiotemporal variation of the local E ×B velocity, which is found to be essential for the existence of this system. The LCO system is quasistationary, existing from 3 to 12 plasma energy confinement times (˜30 - 900 LCO cycles) limited by hardware constraints. This plasma system appears to contribute strongly to the edge transport in these high performance and transient-free plasmas, as evident from oscillations in transport relevant edge parameters at LCO time scale.

  8. Comparison of particle confinement in the high confinement mode plasmas with the edge localized mode of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the DIII-D tokamak

    International Nuclear Information System (INIS)

    Takenaga, H.; Mahdavi, M.A.; Baker, D.R.

    2001-01-01

    Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges. However, it decreased to a much smaller value in the high density discharges

  9. ADX: a high field, high power density, Advanced Divertor test eXperiment

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

    2014-10-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

  10. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  11. Adaptive high learning rate probabilistic disruption predictors from scratch for the next generation of tokamaks

    International Nuclear Information System (INIS)

    Vega, J.; Moreno, R.; Pereira, A.; Acero, A.; Murari, A.; Dormido-Canto, S.

    2014-01-01

    The development of accurate real-time disruption predictors is a pre-requisite to any mitigation action. Present theoretical models of disruptions do not reliably cope with the disruption issues. This article deals with data-driven predictors and a review of existing machine learning techniques, from both physics and engineering points of view, is provided. All these methods need large training datasets to develop successful predictors. However, ITER or DEMO cannot wait for hundreds of disruptions to have a reliable predictor. So far, the attempts to extrapolate predictors between different tokamaks have not shown satisfactory results. In addition, it is not clear how valid this approach can be between present devices and ITER/DEMO, due to the differences in their respective scales and possibly underlying physics. Therefore, this article analyses the requirements to create adaptive predictors from scratch to learn from the data of an individual machine from the beginning of operation. A particular algorithm based on probabilistic classifiers has been developed and it has been applied to the database of the three first ITER-like wall campaigns of JET (1036 non-disruptive and 201 disruptive discharges). The predictions start from the first disruption and only 12 re-trainings have been necessary as a consequence of missing 12 disruptions only. Almost 10 000 different predictors have been developed (they differ in their features) and after the chronological analysis of the 1237 discharges, the predictors recognize 94% of all disruptions with an average warning time (AWT) of 654 ms. This percentage corresponds to the sum of tardy detections (11%), valid alarms (76%) and premature alarms (7%). The false alarm rate is 4%. If only valid alarms are considered, the AWT is 244 ms and the standard deviation is 205 ms. The average probability interval about the reliability and accuracy of all the individual predictions is 0.811 ± 0.189. (paper)

  12. Adaptive high learning rate probabilistic disruption predictors from scratch for the next generation of tokamaks

    Science.gov (United States)

    Vega, J.; Murari, A.; Dormido-Canto, S.; Moreno, R.; Pereira, A.; Acero, A.; Contributors, JET-EFDA

    2014-12-01

    The development of accurate real-time disruption predictors is a pre-requisite to any mitigation action. Present theoretical models of disruptions do not reliably cope with the disruption issues. This article deals with data-driven predictors and a review of existing machine learning techniques, from both physics and engineering points of view, is provided. All these methods need large training datasets to develop successful predictors. However, ITER or DEMO cannot wait for hundreds of disruptions to have a reliable predictor. So far, the attempts to extrapolate predictors between different tokamaks have not shown satisfactory results. In addition, it is not clear how valid this approach can be between present devices and ITER/DEMO, due to the differences in their respective scales and possibly underlying physics. Therefore, this article analyses the requirements to create adaptive predictors from scratch to learn from the data of an individual machine from the beginning of operation. A particular algorithm based on probabilistic classifiers has been developed and it has been applied to the database of the three first ITER-like wall campaigns of JET (1036 non-disruptive and 201 disruptive discharges). The predictions start from the first disruption and only 12 re-trainings have been necessary as a consequence of missing 12 disruptions only. Almost 10 000 different predictors have been developed (they differ in their features) and after the chronological analysis of the 1237 discharges, the predictors recognize 94% of all disruptions with an average warning time (AWT) of 654 ms. This percentage corresponds to the sum of tardy detections (11%), valid alarms (76%) and premature alarms (7%). The false alarm rate is 4%. If only valid alarms are considered, the AWT is 244 ms and the standard deviation is 205 ms. The average probability interval about the reliability and accuracy of all the individual predictions is 0.811 ± 0.189.

  13. Improved charge-coupled device detectors for high-speed, charge exchange spectroscopy studies on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Burrell, K.H.; Gohil, P.; Groebner, R.J.; Kaplan, D.H.; Robinson, J.I.; Solomon, W.M.

    2004-01-01

    Charge exchange spectroscopy is one of the key ion diagnostics on the DIII-D tokamak. It allows determination of ion temperature, poloidal and toroidal velocity, impurity density, and radial electric field E r throughout the plasma. For the 2003 experimental campaign, we replaced the intensified photodiode array detectors on the central portion of the DIII-D charge exchange spectroscopy system with advanced charge-coupled device (CCD) detectors mounted on faster (f/4.7) Czerny-Turner spectrometers equipped with toroidal mirrors. The CCD detectors are improved versions of the ones installed on our edge system in 1999. The combination improved the photoelectron signal level by about a factor of 20 and the signal to noise by a factor of 2-8, depending on the absolute signal level. The new cameras also allow shorter minimum integration times while archiving to PC memory: 0.552 ms for the slower, lower-read noise (15 e) readout mode and 0.274 ms in the faster, higher-read noise (30 e) mode

  14. A measurement of deuterium neutral by the Balmer-series in the STP-2 high beta screw pinch tokamak

    International Nuclear Information System (INIS)

    Yamaguchi, S.; Hirano, K.

    1980-06-01

    The Balmer-alpha and beta are measured with a calibrated spectrograph in STP-2 screw pinch tokamak operated under the maximum toroidal field being 9.2 kG, peak plasma current 30 kA and filling pressure 5 mtorr. The electron temperature and density profiles are obtained by ruby laser Thomson scattering. It is shown that electron temperature is about 10 eV and density is of the order of 10 14 /cm 3 . A non-cylindrical symmetric Abel-inversion technique is used to deduce the emission coefficient profiles from that of the line intensity of the Balmer's. In the present parameter range the neutral deuterium density is almost equal to the population density of the ground state, so that it is obtainable from measured intensities of D sub(α) and D sub(β) which give the population densities of the upper levels i = 3 and 4. The Collisional Radiative (CR) model is applied to the rate equations to estimate the ground state population density. It is found that at 4 μsec from the start of the discharge the deuterium neutral density may be approximately 2 x 10 12 /cm 3 at the center of plasma and 2 x 10 14 /cm 3 at the periphery. These values may contain an error of about factor two. Time history of neutral deuterium density is consistent with the increase of plasma density. (author)

  15. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

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

  16. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  17. Integrated predictive modeling of high-mode tokamak plasmas using a combination of core and pedestal models

    International Nuclear Information System (INIS)

    Bateman, Glenn; Bandres, Miguel A.; Onjun, Thawatchai; Kritz, Arnold H.; Pankin, Alexei

    2003-01-01

    A new integrated modeling protocol is developed using a model for the temperature and density pedestal at the edge of high-mode (H-mode) plasmas [Onjun et al., Phys. Plasmas 9, 5018 (2002)] together with the Multi-Mode core transport model (MMM95) [Bateman et al., Phys. Plasmas 5, 1793 (1998)] in the BALDUR integrated modeling code to predict the temperature and density profiles of 33 H-mode discharges. The pedestal model is used to provide the boundary conditions in the simulations, once the heating power rises above the H-mode power threshold. Simulations are carried out for 20 discharges in the Joint European Torus and 13 discharges in the DIII-D tokamak. These discharges include systematic scans in normalized gyroradius, plasma pressure, collisionality, isotope mass, elongation, heating power, and plasma density. The average rms deviation between experimental data and the predicted profiles of temperature and density, normalized by central values, is found to be about 10%. It is found that the simulations tend to overpredict the temperature profiles in discharges with low heating power per plasma particle and to underpredict the temperature profiles in discharges with high heating power per particle. Variations of the pedestal model are used to test the sensitivity of the simulation results

  18. 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)

  19. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

  20. MAST: a Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

    1995-01-01

    The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)